ML19208D600

Summary of 790830 Meeting W/Utils in Bethesda,Md Re Preliminary Results of Auxiliary Feedwater Sys Reliability Study

ML19208D600
Person / Time
Site:	Rancho Seco
Issue date:	09/20/1979
From:	Capra R Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
TASK-2.E.1.1, TASK-TM TAC-44673, NUDOCS 7909280696
Download: ML19208D600 (28)

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NUCLEAR REGULATORY COMMISSION -- i -E WASHINGTON, D. C. 20555 g.Ng') j o ~g y September 20, 1979 Docket No. 50-312 FACILITY: Rancho Seco Nuclear Generating Station LICENSEE: Sacramento Municipal Utility District

SUMMARY

OF MEETING HELD ON AUGUST 30, 1979 TO DISCUSS THE PRELIMINARY RESULTS OF THE AUXILIARY FEEDWATER SYSTEM RELIABILITY STUDY FOR RANCHO SECO On August 30, 1979, members of the NRC staff met in Bethesda, Maryland with representatives of the Sacramento Municipal Utility District (SMUD) and the Babcock & Wilcox Company (B&W) to discuss the preliminary results of the auxiliary feedwater (AFW) system reliability study for the Rancho Seco Nuclear Generating Station. A list of attendees is provided as Enclosure 1. BACKGROUND As part of the long-term requirements of the Commission Orders issuai to each B&W operating plant licensee in May 1979, the licensees are committed to further review and upgrade their AFW/EFW systems. In order for each licensee to assess which areas of its AFW/EFW system are in need of improvement, the staff directed that each licensee perfonn a reliability study of its AFW/EFW system. The study is to be of a similar scope as that done by the NRC staff for each of the Westinghouse (W) and Combustion Engineering (CE) operating plants. On August 9, 1979, the NRC staff met with the B&W Owners' Group to discuss the scope and schedule for such a program. The Owners' Group picked Ranche Seco 6e lead plant for tne study. It was agreed to at that meeting that the NRC > sould meet with SMUD and B&W on August 30, 1979 to discuss the preliminary ,s of that study. DISCUSSION The meeting was divided into two parts: (1) a review of the Rancho Seco AFW system and (2) a discussion of the plant specific reliability analysis. Part 1 (Rancho Seco AFW system discussion) B&W presented a detailed suninary of the Rancho Seco AFW system. Basically, the Rancho Seco AFW system is comprised of two separate trains. Each train has a separate suction header connected to the seismic Category I condensate storage tank (CST). Backup water supplies can be obtained from the Folsum South Canal or the plant reservoir. Neither of these alternate water sources are classified 1037 7909280(, $

, as seismic Category I. One train utilizes a turbine / motor-driven AFW pump (P-318) and the other train uses a straight motor-driven AFW pump (P 119). Each of the pumps can supply a total of 840 gpm, with 780 gpm fed to the steam generators (S/G) and 60 gpm recirculated to the high pressure and low pressure condensers. The discharge of each pump is cross-connected through a header containing two nomally open, motor-operated valves. Thus, either pump can be used to supply water to either S/G. Flow control to each S/G is controlled through two parallel paths. The nomal path is through a flow control valve which receives its actuatior. signal from the integrated control system (ICS). The flow control valve is electro / pneumatically actuated and will fail to the full open position upon loss of air pressure and to the 50% open position upon loss of electrical power. The alternate path is through the safety-grade, AFW bypass valves. These valves will fully open on a safety features actuation signal (SFAS). The AFW bypass valves are independent of ICS control. Automatic starting of both AFW pumps occurs upon loss of all four reactor coolant pumps (RCPs) or low discharge pressure (850 psig) on both main feedwater (MFW) pumps. In addition, auto-start of the turbine / motor-driven pump will occur upon receipt of an ESFAS signal. All motor-operated valves can be controlled from the control room. Indication of AFW flow to erch S/G and level in the CST are also provided in the control room. All conditions of auto-start of the AFW system are annunciated in the control room. Upon initiation of AFW, either manually or automatically, no repositioning of valves is necessary to get water from the CST to the S/G. Functional tests of the system are performed monthly and quarterly a full flow test is performed on the AFW system. shows the basic mechanical and electrical leyout of the Rancho Seco AFW system. Part 2 (AFW Reliability Analysis) In perfonning the reliability ana' 'is, SMUD defined mission success as being able to provide AFW flow from at lease..ne pump to at least one S/G within either 5,15, or 30 minutes. These times were obtained from the NRC studies conducted on the W and CE plants. The staff suggested that proper justification of these times for~ B&W plants should be made in the draft report which will be supplied to the NRC staff for review. SMUD also used NRC values for unreliability data for hardware, human factors, and preventive maintenance. The study looked at the reliability of the AFW system under three conditions of power availability: (1) loss of main feedwater (LMFW), with both AC and DC power available with a p'robability of 1.0, (2) loss of offsite power (LOOP) with the most limiting DG unavailable with a probability of 10-4 (the other DG is available with a probability of 1.0, and (3) loss of all AC power (LOAC) with only DC and battery-backed AC power available with a probability of 1.0. In?7 .) I aJ

, Other assumptions which were made in the study included: (1) Lines smaller than 1 inch were not considered as possible diverted flow paths. (2) Where valves of identical function were required to be opened or closed manually, both were considered to either be actuated together or not actuated at all. (3) Degraded failures were not considered (components were either fully operable or considered failed). (4) The probability of failure of the CST was assumed 5 X 10-6 (5) The probability of failure of a single train of AFW due to a malfunction of ICS control and initiation was assumed / X 10-3 The major contributors, identified by B&W during the study, which contribute to the unavailability of the AFW system are: (1) a manually operated full flow recirculation valve (FWS-055), (2) the necessity to manually load the motor-drives for the AFW pumps upon loss of offsite power and (3) equipment outages due to preventive maintenance. FWS-055 is a nomally closed, manually-operated valve, located between the discharge of the AFW pumps and the H.P. and L.P. condensers. This valve is manually opened (locally) during the quarterly surveillance testing of the AFW pumps. This valve, when opened, allows full AFW pump discharge flow to be pumped to the H.P. and L.P. condensers. This bypass allows testing of the pumps at rated flow without injecting water into the S/Gs. This valve is not operable from the control room; however, surveillance procedures require that an operator be stationed at the valve whenever the valve is open. This operator must be in continuous communication with the control room, such that if AFW flow is needed while FWS-055 is open, the local operator can manually shut the valve in a timely manner. In the event of a loss of offsite power, the motor-driven AFW pump must be manually loaded on a vital bus (nuclear service bus "4A"). Also, in the event the turbine-drive on the dual-drive AFW pump is inoperable, the motor for this pump must be manually loaded on a vital bus (nuclear service bus "4B"). While procedures have been developed and the operators trained on perfoming this evolution, the reliability study shows that this could be a major contributor to AFW unavailability during the first 15 minutes of a LOOP transient. Preventive maintenance is the major contributor to AFW unavailability in t of loss of all AC power. The assumption in this case, is preventive maintena, being perfonned on the turbine-drive of the dual-drive pump at the time a total LOAC power is experienced. 1037

. In all three cases, the probability of mission success is lower for the first 5 minutes than for the 15 and 30 minute cases. This is due to credit being taken in the latter two cases for operator action to correct for malfunctioning components. Three areas which will be discussed in detail in the report will be: (1) auto-matic actuation of AFW, (2) reliability of back-up water supplies and (3) AFW system dependence on AC power. These areas were found to be weak areas on certain non-B&W designed plants. contains a copy of the reliability analysis handouts used at the meeting. An outline for the report is included as Enclosure 4. CONCLUSIONS The overall methodology for performing the study appears to be consistent with the techniques used to perform the same type of study for W and CE plants. The NRC staff requested that the draft report,which would be submitted to the NRC for review,should incorporate the following items which were not covered in the meeting: (1) The time segments used in the definition of mission success (i.e., 5,15, and 30 minutes) need more definition. That is, the time segments chosen should be justified as being applicable and meaningful for B&W plants. (2) The system description should include failure modes for each active component. (3) The presentation made at the meeting contained no recommendations on improve-ments which should be made to the AFW system based on *.he study. It was pointed out that the purpose of the study was to define or point out dominant faults and/or major contributors to FW unavailability. Therefore, prior to submitting its report, SMUD should incorporate its recommendations on upgrading the timeliness and reliability of the AFW system based on this report. Consistent with the comitments made at the August 9,1979 meeting, the draft report will be sent to the NRC for review by September 17, 1979. ~~2 d. CW R. A. Capra, B&W Project Manager Project Management Group Bulletins & Orders Task Force

Enclosures:

1. List of Attendees 2. Rancho Seco AFW System 3. Reliability Evaluation 4. Proposed Outline for SMUD Report cc w/ enclosures: }[][ j Eee attached

= BABCOCX & WIL;0X OPERATING PLANTS Mr. William 0. Parker Jr. Vice President - Steam Production Duke Power Company P.O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 Mr. William Cavanaugh, III Vice President, Generation and Construction Arkansas Power & Light Company Little Rock, Arkansas 72203 Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municipal Utility District 6201 S Street P.O. Box 15830 Sacramento, California 95813 Mr. Lowell E. Roe Vice President, Facilities Develapment Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, Ohio 43652 Mr. W. P. Stewart Manager, Nuclear Operations Florida Power Corporation P.O. Box 14042, Mail Stop C-4 St. Petersburg, Florida 33733 Mr. R. C. Arnold ~ Senior Vice President Metropolitan Edison Company 260 Cherry Hill Road Parsippany, New Jersey 07054 Mr. James H. Taylor Manager, Licensing Babcock & Wilcox Company Power Generation Group P.O. Box 1260 Lynchburg, Virginia 24505 w-1LU/ .J

'?acram' ento Municipal Utility Page 1 o f 2 District Christopher Ellison, Esq. David S. Kaplan, Secretary and Dian Grueuich, Esc. California Energy' Commission General Counsel 1111 Howe Avenue 6201 S Street Sacramento, California 95825 P. O. Box 15830 Sacramento, California 95813 Ms. Eleanor Sci 1 wart: California State Office Sacranento County 600 Pennsylvania Avenue, S.E., Rm. 201 Board of' Supervisors 327 7th Street, Room 424 Washington, D.C. 20003 Sacramento, California 95814 Docketing and Service Section Office of the Secretary U. S. Nuclear Regula ory Commission Washington, D.C. 20555 Michael L. Glaser, Esq. 115017th Street, N.W. Director, Technical A,sessment Washington, D.C. 20036 Divisicn Office of Radiation Programs Dr. Richard F. Cole (AW-459) Atomic Safety anc Licensing Board U. S. Environmental Protection Agency Panel Crystal Mal? #2 U. S. Nuclear Regulatory Commission Arlington, Virginia 20460 Washington, D.C. 20555 U. S. Environmental Protection Agency Mr. Frederick J. Shen Region IX Office Atomic Safety and Licensing Board ATTN: EIS COORDINATOR Panel 215 Fremont Street U. S. Nuclear Regulatory Commission San Francisco, California 94111 Washington, D.C. 20555 Mr. Robert B. Borsum Timothy V. A. Dillon, Esq. Sabcock & Wilcox Suite 380 Nuclear Power Generation Divisicn 1850 K Street, N.W. Suite 420, 7735 Old Georgetown Road Washington, D.C. 20006 tethesda, Maryland 20014 James S. Reed, Esq. Michael H. Remy, Esq. Reed, Samuel & Remy 717 K Street, Suite 405 Sacramento, California 95814 9,,, l, i G.) s i

Page 2 of 2 o ~ Sacramento Municipal Utility District Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission !!ashington, D.C. 20555 Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D.C. 20555 fir. Richard D. Castro 2231 K Street Sacramento, California 95814 Mr. Gary Hursh, Esq. 520 Capital Mall Suite 700 Sacramento, California 95814 California Department of Health ATTN: Chief Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 e A M U

ENCLOSURE 1 LIST OF ATTENDEES RANCHO SECO AFW RELIABILITY STUDY MEETING AUGUST 30, 1979 NAME ORGANIZATION S. I. Anderson SMUD (Nuclear Engineer) R. J. Finnin B&W (Licensing) W. W. Weaver B&W (Tech. Staff) R. W. Dorman B&W (Plan' Integration) B. J. Short B&W (Customer Service) T. M. Novak NRC (Deputy Dir. B&O Task Force) P. R. Matthews NRC (Section Leader, Systems Group, B&O Task Force) C. Y. Liang NRC (Systems Group, B&O Task Force) W. T. LeFave NRC (Systems Group, B&O Task Force) M. A. Taylor NRC (Probabilistic Analysis Branch) R. A. Capra NRC (B&W Project Manager, B&O Task Force) / 1U;7 os

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ENCLOSURE 2 page 5 FILTERS g, AIR S g DRYER 480V AC BUS "3C2" - C 900A 480V AC BUS "3C)" C 9008 ^ FV -20527 W X 480V AC BUS "302" ~ C-9000 FV-20528 M X I C0' PRESSORS, COOLERS, J EDISTURE SEPARATORS, RECEIVERS , AIR SUPPLY TO AFWS CONTROL VALVES-Sh00 1j l u,m ]/ 4

ENCLOSURE 3 page 1 ASSUMPTIONS / CRITERIA FOR SMUD AFWS RELIABILITY STUDY 1. Definition of Mission Success - Flow from at least one pump to at least one steam generator within 5, 15 and 30 minutes. 2. Assumed validity of NRC - supplied unreliability data including hardware, human factors, preventive maintenance. 3. Fower Availability - LMFW - All AC and DC available with probability of 1.0. LOOP - Most limiting DG is unavailable with a probability of 10-2 n . The other ger.erator (typically DG "B") is available with a probability of 1.0 LOAC - Only DC and bat'ary-backed AL is available with a probability of 1.0. 4. Lines of i 1" were ignored as possible dive. ad flow paths. 5. Assumed coupled manual initiation of valves with identical function. i.e., valves were assumed both opened manually or both not opened. 6. Degraded failures were not considered i.e., components were either 100% okay or were considered failed. (Exception was loss of power to E/P converters re-considered not failed closed). sultirg in 50% valve position Condensate Storagd Tank failure probability = 5x10-6 7. 8. Assumed a single ain for ICS control and initiation with a failure pro-bability of 7x10-037

~ MAJOR CONTRIBUTORS TO UNAVAILABILITY e FWS-055 e MANUAL LOADING OF MOTOR DRIVES FOR PUMPS e PREVENTIVE MAINTENANCE 9 0 0 tiler ITEMS m u a e AUTOMATIC ACTUATION ~- o BACKUP WATER SUPPLY e AC DEPENDENCIES -g ~ D

CASE I : LMF %/ C AsE 2. LOO P C ASE 3: LOAC m.__ Low MEb HIGH low MEb 416 8 low MGb 4tG R s A A A I SMoD is a n s i 30 0 0 S .i i {. = n i 5 ALL N i 0 I C h '^ rv1 PLAMTS i m u m l I A-Missioo success Wirgic s S.co res O* 5- $issico Suctnss diTuio IS Miuu res 4-G ~ C' W Dt.A u v s ~ $~ M iSstou SUCCESS bd Tuio '50 AiuuTE s Co mma,sou SMOD AF W S Rcuxw u ry ou iu era NRC. Resu a s Fog d Ptsats

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ENCLOSURE 4 PROPOSED OUTLINE FOR SMUD REPORT ON AUXILIARY FEEDWATER SYSTEtt RELIARILITY 1.0 Introduction 1.1 Background - NRC study for W and CE 1.2 Objectives - Comparative reliability assessment, identi-fication of dominant failure contributors 1.3 Scope - 3 cases: LMFW, LOOP and LOAC; 3 times: 5, 15 and 30 minutes; Baseline confiauration: 1 Auqust 1979 1.4 Analysis Technique - Fault tree analysis leadino to reliability insights 1.5 Assusptions and Criteria 1) Power availability for each case 2) Criteria for mission success 3) NRC failure data 4) etc. 2.0 System Descriotion 2.1 Overall AFWS Desian 2.2 Supportino Systems - Including backup water sources 2.3 Power Sources 2.4 Instrumentation and Control - Initiation and Control 2.5 Human Factors - Initiation, backup actions, indications and contols available 2.6 Maintenance / Testing - Frequency, extent, duration 2.7 Tech Spec - Limitations imposed 3.0 Reliability Evaluation 3.1 Comparative Reliability - Table for comparison with W and CE 3.2 Dominant Failure Contributors 3.2.1 LMFW 3.2.2 LOOP 3.2.3 LOAC J

Distribut;on: Docket 50-312 NRC PDR LPDR ORB 4 Reading NRR Reading B&O T/F Reading HDenton ECase VStello RVollmer WRussell BGrimes TCarter DEisenhut RReid DGarner Slewis LBrenner RIngram OI&E (3) RFraley, ACRS (16) anan Dross JHeltemes SIsrael GMazetis ZRosztoczy PNorian HSilver TNovak PMatthews CLiang WLeFave MTaylor RCapra l ], -0}}