ML19249B115
| ML19249B115 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/02/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 7909040004 | |
| Download: ML19249B115 (12) | |
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s Cate ROUTIN6 AND ' TRANSMITTAL SUP 8/2/79 Initials Date TC: {Name. orrice symbol, room nv.110er, Dutiding. Agency / Post)
THOSE LISTED ON SERVICE LIST 3.
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Note and Return Fife Action Per Conversation For Ctearance Approval For Correction Prepare Reply As Requested For Your Information See Me Circulate Comment Investigate Signature Coordination Justify REMARKS It has been noted that pages 1-12 were incomplete.
Please substitute these new pages.
Thank you, DO NOT use this form as a RECORD of approvals, concurrences, dispossis, clearances, and sirtilar actions Room No.-D'hg.
FROM:(Name, org. symbol. Mency/ Post)
Phone No.
OPTIONAL FORM 41 (Rev. [ 76)
W 1-102 Prescribed by CSA FPMR (41 Cf el) 101-11.2o6 NU.S.Gro197m-o.261-ht?3354 7909040 @ fp;7 sayeso
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UNITEDhTATES i Yy, s, i
NUCLEAR REGULATORY COMMISSION 2g p.~
C W ASM NGTO N,0. C. 20555 s
July 2,1979 Docket No.:
50-346 MEMORANDUM FOR:
Chairman Hendrie Comissioner Gilinsky Conenissioner Kennedy Commissioner Bradford Commissioner Ahearne FRCM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation ExecutiveDirectorforOperations%h THRU :
SUBJECT:
DAVIS-BESSE 1 Enclosed is a revised Safety Evaluation for Davis-Besse 1, relative to the shutdown order of 16 May 1979.
We h3ve provided supplemental material at page 6, and following, to account for our evaluation of an LER, in the auxiliary feedwater system area, which we recently received. / iso enclosed for your information are two letters on this subject; cre is from NRR to Toledo Edison, dated June 29, 1979; the second letter is Toledo Edison's response.
/
~
Harold R. Denton, Director O'fice of Nuclear Reactor Regulation
Enclosures:
1 Revised SER 2.
Ltr dtd 6-29-79 Denton to Toledo Edison 3.
Ltr dtd 6-29-79 Toledo Edison to Denton cc w/ enclosures: See next page 8000?$$.
Toledo' Edison Company cc w/ enclosure (s):
Mr. Donald H. Hauser, Esq.
Director, Technical Assessment The Cleveland Electric Division Illuminating Company Office of Radiation Programs P. O. Box 5000 (AW-459)
Cleveland, Ohio 44101 U. S. Environmental Protection Agency Crystal Mall #2 Gerald Charnoff, Esq.
Arlington, Virginia 20460 Shaw, Pittman, Potts and Trowbridge U. S. Environmental Protection Agency 1800 M Street, N.W.
Federal Activities Branch Washington, D.C.
20036 Region V Office ATTN:
EIS COORDINATOR Leslie Henry, Esq.
230 South Dearborn Street Fuller, Seney, Henry and Hodge Chicago, Illinois 60604 300 Madison Avenue Toledo, Ohio 43604 Mr. Samuel J. Chilk, Secretary U. S. Nuclear Regulatory Commission Mr. Rooert B. Borsum Washington, D.C.
20555 Babcock & Wilcox Nuclear Powei Generation Division The Honorable Tim McCormack Suite 420, 7735 Old Georgetown Road Ohio Senate Bethesda, Maryland 20014 Statehouse Columbus, Ohio 43215 Ida Rupp Public Library 310 Madison Street The Honorable Tim McCormack Port Clinton,0hio 43452 170 E. 209th Street Euclid, Ohio 44123 President, Board of County Comissioners of Ottawa County Mr. Lcwell E. Roe Port Clinton, Ohio 43452 Vice President, Facilities Development Attorney General Toledo Edison Company Department of Attorney General Edison Plaza 30 EN Broad Street 300 Madison Avenue Columbus, Ohio 43215 Toledo, Ohio 43652 Harold Kahn, Staf f Scientist Bruce Churchill, Esq.
Power fiting Commission Shaw, Pittman, Potts & Trowbridge 361 East Broad Street 1800 M Street, N.W.
Columbus, Ohio 43216 Washington, D.C.
20036 Ohio Department of Health Atomic Safety & Licensing Board Panel ATTN:
Director of Health U. S. Nuclear Regulatory Commission 450 East Town Street Washingi.on, D. C.
20555 Columbus, Ohio 43216 Atomic Safety and Licensing Appeal Panel Docketing and Service Secticn U. S Nuclear Regulator.v Commission
,,ashington, D.C.
205cc Office of the Secretary n
U. S. Nuclear Regulatory Ccmmission Washington, D.C.
20555 S00 82 0
Toledo Edison Company cc w/ enclosure (s):
Ivan W. Smith, Esq.
Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Cadet H. Hand, Jr.
Director, Bodega Marine Laboratory University of California P. O. Box 247 Bodega Bay, California 94923 Dr. Walter H. Jordan 881 W. Outer Drive Oak Ridge, Tennessee 37830 Ms. Jean DeJuljak 381 East 272 Euclid, Ohio 44117 Ohio Department of Heaith ATTN: Director of Health 450 East Town Street Columbus, Ohio 43216 S*0nGisa
EVALUATION OF LICENSEE'S COMPLIANCE WITH THE NRC ORDER DATED MAY 16, 1979 TOLEDO EDISON COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING CCMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT No. 1 DOCKET NO. 50-346 INTRODUCTION By Order dated May 16, 1979, (the Order) the Toledo Edison Company and the Cleveland Electric Illuminating Company (TECO or the licensee) were directed by the NRC to take certain actions with respect to Davis-Besse Nuclear Power Station, Unit 1 (D8-1).
Prior to this Order and as a result cf a preliminary review of the Three Mile Island, Unit No. 2 (TMI-2) accident, the NRC staff initially identified several human errors that contributed significantly to the severity of the event.
All holders of operating licenses were subsequently instructed to take a number of immediate actions to avoia repetition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (IE).
Subsequently, an additional bulletin was issued by IE which instre ted holders of operating licenses for Babcock &
Wilcox (S&W) designed reactors to take further actions, including immediate changes to decrease the reactor high pressure trip point and increase the pressurizer power-operated relief valve (PORV) setting.*
3:etins Nos. 79-05 (April 1, 1979),79-05A (April 5, 1979), and 79-05B (April 21, 1979) apply to all B&W acilities.]
903056 The NRC staff identified certain other safety concerns that warranted addi-tional short-term design and procedural changes at operating facilities having B&W designed reactors.
Those were identified as items (a) through (e) on page 1-7 of the " Office of Nuclear Reactor Regulation Status Report to the Commission" dated April 25, 1979.
After a series of discussions between the i,RC staf f and the licensee concerning possible design modifications and changes in operating procedures, the licensee agreed, in letters dated April 27,1979 and May 4, 1979, to perform promptly certain actions.
The Commission found that operation of the plant should not be resumed until the actions described in Items (a) through (g) of paragraph (1) of Section IV of the Order are satisfactorily completed.
Our evaluation of the licensee's compliance with items (a) through (g) of paragraph (1) of Section IV of the Order is given below.
In performing this evaluation we have utilized additicnal information provided by the licensee in letters dated May 11, 18, 19, 22 (2), 23 (2), 26 (2), 29 and June 15 (2), 18, 21, 23 and 25, 1979 and numerous discussions with the licensee's staff.
Confirmation of design and procedural changes was made by members of the NRC staff at the 08-1 site.
An audit of the training and performance of the DB-1 reactor operators was also performed by the NRC staff to assure that the design and procedural changes were understood and were being correctly implemented by the operators.
3Of)0EG EVALUATION Item (a)
It was ordered that the licensee take the following action:
" Review all aspects of the safety grade auxiliary feedwater system to further upgrade components for added reliability and performance.
Present modifications will include the addition of dynamic braking on the auxiliary feedpump turbine speed changer and provision of means for control room verification of the auxiliary feedwater flew to the steam generators.
This means of verification will be provided for one steam generator prior to startup from the present maintenance outage and for the other steam generator as soon as vendor-supplied equipment is available (estimated date is June 1, 1979).
In addition, the licensees will review and verify the adequacy of the auxiliary feedwater system capacity."
The auxiliary feedwater (AFW) system at DB-1 consists of two safety grade AFW pumps capable of beino actuated and controlled by safety grade signals that ensure the availability of feedwater to at least one steam generator, under the assumed conditions of a single failure.
In addition, the capability to manually actuate and control AFW is available in the control rocm.
The sources of water include two condensate storage tanks (CST), the service water system
'N
' ire orotection system.
The CSTs provide the normal supply (non-safety-grade) and the service water system is used as a backup safety grade supply.
OOf?CE.'i3 A low level in either CST is alarmed to the operator and a continuous level is displayed inside the control room.
Low pressure switches on the AFW pump suction provide safety grade signals to automatically shift suction for the pump from the CSTs to the backup service water supply.
Additionally, the operator could also manually transfer the AFW suction to the fire water storage tank (FWST) in the fire protection system.
Both steam-driven auxiliary feedwater pump turbines at DB-1 are provided with a governor used for variable pump speed control.
The governor is equipped with a small DC rotor which changes the speed setpoint on the turbine control valve, thereby controlling steam flow which regulates the turbine and pump speed.
This DC motor receives " raise-and-lower" pulses from the safety grade steam generator level control system or the manual control switches (located in the control room), which change the turbine speed as required.
Pulse length is automatically increased the further steam generator level deviates from its setpoint.
These changes in pamp speed alter the AFW flow and thus control the water level in the steam generators.
A " dynamic brake" feature has been added, which consists of a resistor and electrical contacts in parallel with the windings of the DC motor.
When the control pulse is terminated, the braking resistor is placed in parallel with the motor windings, causing rapid dissipation of the energy associated with the motor mcmentum (thus reducing the amount of motor coast).
This, in turn, w ces the amount of pump speed overshoot, thereby allowing fewer speed cnanges to match the AfW flow rate to the steaming rate of the steam generators.
20r >0E-?
The licensee has also added flow rate indication for both steam generator AFW inlet lines.
Each inlet line has a pipe-mounted ultrasonic flow transducer and signal conditioner.
These are located in the auxiliary building and are accessible during normal plant operations.
The signal conditioners provide outputs both locally and in the control rocm on the AFW pump section of the main control console.
Each device is designed to provide flow rate indication to each steam generator from 0 to 1000 gpm.
The systems are powered from 120 VAC, 60 Hz buses which are fed by redundant non-Class IE station inverters.
Functional testing of the installed auxiliary feedwater flow rate indication is to be conducted in conjunction with the functional testing of the dynamic braking modification of AFW pump turbine controls.
The staff concludes that the dynamic bra u and AFW flow rate indication modifications are acceptable contingent upon successful testing prior to restart.
We have reviewed the piping and instrumentation diagrams and have determined that no active failure of a mechanical component, such as a pump or valve, would preclude obtaining the required AFW flow rate.
The licensee has pre-viously performed tests of the manual and automatic level control system.
The test results showed that the control system functioned as designed to control steam generator level.
Verificati n of acceptable flow capacity for each of the two AFW pumps was based upon recorded steam generator level changes following a previous reactor trip.
These data showed that each pump exceeded the design flow rate of 800 gpm at a steam generator pressure of 1050 psig.
(N ^00 gm is the flow rate delivered to the steam generators and does not include tne aoproximately 250 gpm recirculation flow rate.)
SOoOE8 Additional information submitted by the licensee (letter from Lowell E. Roe (TECO) to Mr. Robert W. Reid (NRC) dated May 23,1979) shows that a total minimum flow, to one or both steam generators, of 550 gpm is required to support the accident analyses.
Based on these data and analyses, and the agreement by the licensee to perform checkout testing of the dynamic braking and flow rate indication modifications prior to restart, we conclude that adequate assurance exists that the AFW system will deliver the required flow rate upon demand.
By 'etter (Lowell E. Roe (TECO) to Mr. Robert W. Reid (NRC) dated May 23, 1975), the licensee provided results of a review of the operating history of the AFW system at 08-1.
The largest number of failures" occurred during the initial operating and debugging phase of the facility.
Fourteen (14) of the seventeen (17) reported failures occurred prior to January, 1978.
Subsequent to implementing system design changes as a resuit of several of these failures, the systems failure rate has been reduced and its reliability enhanced.
There have been 3 failures of AFW system components from January 1378 to date.
(There were a total of 65 actuations of the AFW system in this time period.)
Three different components in the AFW system were involved in these chree failures:
(1) the speed control circuit for #1 AFV pump turbine, (2) a faulty limit switch on an AFW discharge valve, and (3) two sticky AFW pump turbine steam supply valves.
In each case, the licensee performed corrective actions and failures in these components have not reoccurred.
A more rectnt letter
^;For the purpose of demonstrating improvement in the performance of the AFW system, the licensee has defined a failure of the AFW system to be any event for which at least one train of the AFW system is not delivering design flow to a steam generator.]
S00;C63 (Lowell E. Roe (TECO) to Mr. Robert W. Reid (NRC) dated June 29,1979) addressed a series of pressure switch failures which wt.re discovered on May 21, 1979, and which affected both AFW trains.
An evaluation of these failures by the licensee concluded that both trains would have automatically actuated if required, but that one train would not have shifted automatically to the service water supply.
The NRC staff has discussed these failures with TECO and has requested that an improved surveillance program for tnese pressure switches De initiated to determine the cause of the failures and the optimum calibration interval.
The licensee has agreed to increased frequency of switch calibration.
In addition, the licensee has made procedural changes, requested by the staff, to instruct the operator to manually shift to the alternate supply of water for the AFW pumps, when the CST level drops to three feet (if automatic switchover has not occurred).
This procedure provides greater assurance that, even with failures of this nature, the AFW system is available during the longer term.
The staff concludes, that the licensee has increased the reliability of the AFW system by implementing appropriate corrective actions and design modifications.
In addition, the licensee has revised the administrative procedure pertaining to valve alignment and control.
These revisions to AD 1839.02 (" Operation and Centrol of Locked Valves") provide further assurance that mispositioning of AFW system valves would be detected.
~, sed on the above evaluation, the NRC staff concludes that the licensee has complied with the requirement of Item (a) of the Order.
80f)C30 Item (b' It was also ordered that the licensee:
" Revise ops.ating procedures as necessary to eliminate the option of using the Integrated Control System as a backup means for controlling auxiliary feedwater flow."
As indicated in Item (a), the 08-1 AFW system has been designed as a safety grade system and, as such, is separate from the integrated control system (ICS); however, the licensee has indicated that the AFW system is capable of being switched to the ICS mode for a backup means of control.
As currently designed, the AFW system has three operational modes of controlling flow:
"ICS control", " auto-essential" and " manual." We requested that the licensee consider a more positive means to assure the continued separability of the ICS centrol position of the code selector switches.
The licensee agreed (letter from Lowell E. Roe (TECO) to Mr. Robert W. Reid (NRC) dated June 15, 1979) to install a mechanical stop on these switches to further deter use of the ICS control position.
The IE site inspector has verified the installation of this mechanical stop.
The licensee has revised SP 1106.06 (" Auxiliary Feedwater System"), which describes procedures for AFW system cperation.
This procedure specifically prdibits the use of the ICS control position on the mode selector switches.
Procedural steps for placing tne AFW system in service for plant startup SOr>0S1 require the operator to place the AFW mode selector switches in the auto-essential position.
We have reviewed the revised procedure for AFW switch operation and conclude there is sufficient guidance to prevent use of the AFW system in the ICS mode of control.
Other plant procedures that made reference to the ICS control mode of AFW have been revised by the 1-;censee to no longer authorize that made of control..The staff has reviewed those procedures and concludes that those revisions are equate.
In addition, the NRC staff audit confirmed that the control room operators are aware that ICS control of AFW is prohibited.
t Based on the above evaluation, we conclude that the licensee has complied with the requirements of Item (b) of the Order.
Item (c)
The Order requires that the licensee:
" Implement a hard-wired control grade reactor trip that would be actuated on loss of main feedwater and/or turbine trip."
The 08-1 original design did not have a direct reactor trip from a malfunction in the secondary system (loss of main feedwater and/or turbine trip).
To cbtain an earlier reactor trip (rather than delaying the trip until an cperator took action or until a primary system parameter exceeded its trip setpoint),
S000'"2 the licensee committed to install a hard-wired, control grade reactor trip on the loss of all main feedwater and/or on turbine trip (letter from Lowell E. Roe (TECO) to H. Denton (NRC) dated April 27,1979).
The purpose of this antici-patory trip is to minimize the potential for opening of the power-operated relief valve (PORV) and/or the safety valves on the pressurizer.
This new circuitry meets this objective by providing a reactor trip during the incipient stage of the related transients (turbine trip and/or loss of main feedwater).
TECO has added control grade circuitry to 08-1 which is designed to provide an automatic reactor trip when either the main turbine trips or there is a reverse differential pressure of 177 psid across both of the two main feedwater check valves (one check valve is located in the main feedwater discharge piping ecsociated with each steam generator).
The main turbine trip is sensed by a normally deenergized auxiliary relay associated with the main turbine generator master trip bus.
The power for this bus is provided from a 24 volt DC source, which in turn is provided power (through rectifier circuitry) from a non-Class lE inverter supplied 120 volt AC distribution panel.
A contact from the above auxiliary relay is arranged into a 120 volt AC circuit containing four normally deenergized relays.
Power for this 120 volt circuit is provided from a Class lE inverter supplied distribution panel.
The design for these four relays and appropriate associated circuitry conform to Class lE requirements, including physical independence and provisions for testing.
Each of these four relays provide one contact which is arranged in series with one of the four Class lE undervoltage coils associated with one of the four AC reactor trip circuit r:<, c; v. c o a Le breakers (one undervoltage coil associated with each AC reactor trip circuit breaker).
When these relays are energized, power to the associated Class lE undervoltage coils is interrupted so as to produce the desired reactor trip.
As indicated above, differential pressure switches a:ross check valves, located in the main feedwater pump discharge piping, actuate upon sensing a reverse differential pressure across these check valves.
Two contacts from these differential pressure switches are arranged into a 125 volt DC circuit, which is provided power from a Class lE 125 volt distribution panel.
This circuit contains two associated DC relays.
Two contscts (one contact per relay) associated with these relays are arranged in series.
This teries contact arrangement is provided in parallel with the contact associated with the main turbine generator master trip bus.
The remaining circuitry associated with this trip is
.entical and common (shared) to that described above for the turbine trip (including power supply identification).
rovisions have been included in the design to manually bypass and to reinstate o
the reactor trip feature associated with the main turbine generator trip.
To supplement this feature, the design includes an annunciator which actuates whenever this reactor trip is bypassed and the reactor power level is above 15 percent.
Access to this bypass switch will require a key which is under suitable administrative control.
Operator verification of the bypass removal is required by procedure during power escalation.
The NRC staff has reviewed 15cce - ocedures and concludes that sufficient administrative control exists.
No bycass features are included in the cesign for the reactor trip feature 80'?CS4 associated with the loss of main feedwater circuitry.
During nor.Tal startup or shutdown, an electric auxiliary. pump is used when the steam driven main feedwater pumps are not available.
The licensee has analyzed this additional circuitry with respect to its independence from the existing reactor trip system and to assure that the design and operation of this additional circuitry will neither degrade the reliability of the existing reactor protection system nor create any new adverse safety system interactions.
Based on our review of the implementation of the added trip circuitry, with respect to its independence from the existing trip circuitry, we conclude that this addition will not degrade the existing reactor pr,tection system design.
In addition, the licensee has satisfactorily completed testing of this trip circuitry.
The iicensee has committed to perform a monthly periodic test of the added circuitry to demonetrate its ability to open the AC reactor trip circuit areakers (tripping of the AC reactor trip circuit breakers via the under-voltage trip circuit).
We conclude that there is reasonable assurance that the additional circuitry will perform its intended function.
Based on the above evaluation, we conclude that the licensee has complied with the requirements of Item (c) of the Order.
I',-fd' This Item in the Order requires the licensee to:
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