ML19289F278
| ML19289F278 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/16/1979 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | NRC COMMISSION (OCM) |
| References | |
| NUDOCS 7906070145 | |
| Download: ML19289F278 (10) | |
Text
{{#Wiki_filter:4 N A7?C ptgLIC DOCDLDIT R00M UNITED STA'IES OF AMERICA gnew p. umu -7 g y 7 jg79) 9 NUCLEAR LMTCRY CCMMISSION 7 g,;f fM IO s // In the Matter of ) )
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'n1E IDLEEC EDISCN CCMPAFt AND ) ' HIE CLEVM ELECTRIC ILWMINATDG ) Docket No. 50-346 CCMPANY ) ) Davis-Besse Nuclear Power Station, ) Unit No. 1 ) ORDER I. The Toledo Edison Company (TECO) and 'Ihe Cleveland Electric Illuminating Company (the licensees), are telders of Facility Operating License No. NPF-3 which authorizes the operation of the nuclear power reactor known as Davis-Besse Nuclear Power Station, Unit No.1 (the facility or Davis-Besse 1), at steady state power levels not in excess of 2772 megawatts thermal (rated power). The facility is a Babcock & Wilcox (B&W) designed pressurized water reactor (WR) located at the licensees' site in Ottawa County, Ohio. II. In the course of its evaluation to date of the accident at the Three Mile Island Unit No. 2 facility, which utilizes a B&W designed P@, the Nuclear Regulatory Comission staff has ascertained that B&W designed 2232 249 7006070345 u s
7530-01 J reactors appear to be unusually sensitive to certain off-normal transient conditions originating in the secondarv system. We features of the B&W design that contribute to this sensitivity are: (1) design of the steam generators to operate with relatively small liquid volumes in the second-ary side; (2) the lack of direct initiation of reactor trip upon the occur-rence of off-normal conditions in the feedwater system; (3) reliance on an integrated control system (ICS) to automatically regulate feedwater flow; (4) actuation before reactor trip of a pilot-operated relief valve on the primary system pressurizer (which, if the valve sticks open, can aggravate the event); and (5) a low steam generator elevation (relative to the reactor vessel) which provides a smaller driving head for natural circulation.* Because of these features, B&W designed reactors place more reliance on the reliability and performance characteristics of the auxiliary feed-water system, the ICS, ard the emergency core coolirg system (ECCS) per-formance to recover from frequent anticipated transients, such as loss of offsite power and loss of normal feedwater, than do other ER designs. 21s, in turn, places a large turden cn the plant operators in the event of off-normal system behavior durirg such anticipated transients.
- It is noted that although features numbers 3 and 5 do not apply to Cavis-Besse 1 to the same extent as they apply to other current'.y licens(d 4W designed reactors, the other features are fully a pli-cable. -
eP, 2232 250
7590-01 As a result of a preliminary review of the Three Mile Island Unit No. 2 accident chronology, the NRC staff initially identified several human errors that occurred during the accident and centributed significantly to its severity. All holders of operatig licenses were subsecuently instructed to take a nu:6er of i::raediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Cocmission's Office of Inspection and Enforcement (IE). In additien, the NRC staff began an immediate reevaluation of the desigr features of S&W reac*wrs to determine whether additional safety corrections or improvements were necessary with respect to these reactors. This evaluation involved numerous meetings with B&W and certain of the affected licensees. The evaluatien identified design features as discussed above which indicated that B&W designed reactors are unusually sensitive to certain off-normal transient conditions originating in the secondary system. As a result, an additional hulletin was issued by IE which instructed holders of operating licenses for B&W reactors to take further actions, including i::cediate changes to decrease ti e reactor high pressure trip point and increase the pressurizur pilot-operated relief valve setting. Also, as a result of this evaluation, the NRC staff identified certain other safety concerns that warranted additional short-term design and procedural changes at operating facilities having B&W designed reactors. 2232 251 se a-31 4
7590-01 These were identified as items (a) through (e) on page 1-7 of the Office of Nuclear Reactor Regulation Status Report to the Comission of April 25, 1979. After a series of discussions between the NRC staff and the licensees concerning possible design modifications and changes in operating pro-cedures, the licensees agreed in letters dated April 27 and May 4,1979, to implement promptly the following actions: (a) 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 cn the auxiliary feedpump turbine speed changer and provision of means for control room verification of the auxiliary feedwater flow 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 equipnent is available (esti:reted date is June 1, 1979). In addition, the licensees will review and verify the adequacy of the auxiliary feedwater system capacity. (b) Revise operating procedures as necessary to eliminate the option of using the Integrated Control System as a backup means for controlling auxiliary feedwater tiow. IYh 2232 252
7590-01 . (c) I= clement a hard-wired control-grade reactor trip that would be actuated on loss of =ain feedwater and/or turbine trip. (d) Ceeplete analyses for pstential small breaks and develop and i=plement operatirg instructions to define operator action. (e) All licensed reactor operators and senior reactor operators will have cocpleted the Bree Mile Island Unit No. 2 simulator trainire at B&W. (f) Subnit a reevaluation of the ECO analysis of the need for automatic or administrative centrol of steam generator level setpsints during auxiliary feedwater system operation, previously submitted by TECO letter of Dece.ber 22, 1978, in light of the tree Mile Island Unit Ma. 2 incident. (g) Suhnit a review of the previous TECO evaluation of the September 24, 1977 event involvirg equipnent problens and depress-urization of the primary systen at Davis-Besse 1 in light of the tree Mile Island Unit No. 2 incident. In its letters the licensees also stated that the actions listed in (a) through (g) above would, except as reted in item (a), be completed prior to startup from the current maintenance outage. 2232 253 ~'
7590-01
- In addition to these :nodifications to be implemented prcmptly, the licensees have also proposed to carry out certain additional long-term ::cdifications to further enhance the capability and reliability of the reactor to re-spend to various transient events. Dese are:
- 2e licensees will continue to review performance of the auxiliary feed-water system for assurance of reliability and performance. - te licensees will submit a failure mode and effects analysis of the ICS to the NRC staff as soon as practicable. S e licensees stated that this analysis is now underway with high priority by B&W. - The reactor trip following loss of main feedwater and/or trip of the turbine to be installed promptly pursuant r_o this Order will thereafter be upgraded so that the components are safety grade. Se licensees will submit this design to the NRC staff for review. - Continued attention will be given to transient analysis and procedures for managecent of small breaks. - Se licensees will continue reactor operator training and drilling of response procedures to assure a high state of preparedness. ?M 2232 254
7590-01 . The Comission has concluded that the prompt actions set forth as (a) through (g) above are necessary to provide added relir.bility to the reactor system to respond safely to feedwater transients and should be confirmed by a Cocnission order. The Comission finds that operation of Davis-Bese 1 should rot be re-sumed until the actions described in paragraphs (a) through (g) above, with the exception as noted in item (a), have been satisfactorily completed. For the foregoing reasons, the Comission has found that the p2blic health, safety and interest require that this Order be effective immedi-ately. III. Copies of the follovirs documents are available for inspection at the Cocnission's Public Document _ Room at 1717 H Street, N.W., Washington, D.C. 20555, and are being placed in the Comission's local p1blic document room in the Ida Rupp Public Library, 310 Madison Street, Port Clinton, Ohio 43452: (1) Office of Nuclear Reactor Regulation Status Report en Feedwater Transients in B&W Plants, April 25, 1979. 2232 255 b .t
~ ~ ~ ~- ~~ 7590-01 ~ (2) Letters from Lowell E. Roe (TECO) to Harold Centon (NRR) dated April 27 and May 4, 1979. IV. Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and the Cocnission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS FELEBY CRDERED TFAT: (1) The licensees shall take the following actions with respect to Davis-Besse 1: (a) 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 cn the auxiliary feedpump turbine speed changer and provision of means for control room veri-fication of the auxiliary feedwater flow to the steam generators. 'Ihis 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 vender-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 y ') Revisej operating procedures as necessary to eliminate the option of using the Integrated Control System as a backup means for controlling the auxiliary feedwater sys*a. 2232 256
7590-01 _g. (c) I=plement a hard-wired control-grade reactor trip that would be actuated cn loss of main feedvater and/or turbine trip. (d) Coc:clete analyses for Intential small breaks and develop and i=plement operating instructions to define operator action. (e) All licensed reactor operators and senior reactor operators wil.1 have completed the tree Mile Island Unit No. 2 simulator trainire at B&W. (f) Submit a reevaluation of the TECO analysis of the need for automatic or administrative control of steam generator level setpoints durire auxiliary feedwater system operation previously sdnitted by TECO letter dated December 22, 1978, in light of the tree Mile Island No. 2 incident. (g) Sutait a review of the previous TECO evaluation of the September 24, 1977 event involving equipnent problems and depressurizatien of the primary system at Davis-Besse 1 in light of the tiree Mile Islard Unit No. 2 incident. (2) The licensees shall maintain Davis-Besse 1 in a shutdcwn condition until items (a) throtgh (g) in paragraph (1), except as noted in item (a), above are satisfactorily cocpleted. Satisfactory ccmpletion will recuire confir-mation by the Director, Office of Nuclear Reactor Regulation, that the 2232 257
7590-01 ~ actions specified have been taken, the specified analyses are acceptable, ard the specified implement 1rg procedures are appropriate. (3) Die licensees shall as prceptly as practicable also accernplish the long-term modifications set forth in Section II of this Order V. Within twenty (20) days of the date of this Order, the licensees or a y n person whose interest may be affected by this Order may request a hearing with respect to this Order. Any such request shall not stay the immediate effectiveness of this Order. M WC REGtRNIORY CCNMISSIQi ~ (m. - q,. .... v LSamuel J. Ik Secretary o the Comission DatedagbdayofMay1979 Washirgton, D.C., thisf6,,, 2232 258 I A g b =b}}