ML19270G255
| ML19270G255 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/07/1979 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 7906050195 | |
| Download: ML19270G255 (8) | |
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7590-01 UNITED STATES OF AMERICA NUCLEAR REGUIATORY CCMMISSION M PUBLIC$0CMfENT ROOM In the Matter of
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SACRAMDITO MUNICIPAL LTTILITY DISTRICT)
Ebcket No. 50-312
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Rancho Seco Nuclear Generating Station )
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N The Sacramento Municipal Utility District (the licensee or SMUD) is the holder of Facility Operatirq License No. DPR-54 which authorizes the operation of the nuclear power reactor known as the Rancho Seco Nucl sar Generatig Station (the facility or Rancho Seco), at steady state r.ower levels not in excess of 2772 megawatts thermal (rated power).
The facility is a Babcock & Wilcox (B&W) designed pressurized water reactor (PWR) located at the licensee's site in Sacramento County, California.
II.
In ce course of its evaluation to date of the accident au the Three Mile Island Unit No. 2 facility, which utilizes a B&W designed PAR, the Nuclear Regulatory Commission staff has ascertained that B&W designed reactors appear to be unusually sensitive to certain off-normal transient conditions originating in the secondary syst:m. The features of the B&W design that contribute to this sensitivity are:
(1) design of the steam generators to operate with relatively small licuid voltraes in the secondary
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' side; (2) the lack of direct initiation of reactor trip upon the occurrence of off-normal conditions in the feedwater system; (3) reliance on an integrated control system (ICS) to automatically regulate feedwater for; (4) actuation before reactor trip of a pilot-operated relief valve on the primary systen 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 snaller driving head for natural circu-lation.
Because of these features, B&W designed reactors place more reliance on the reliability and perfonnance characteristics of the auxiliary feedwater system, the integrated control system, and the energency core cooling system (ECCS) performance to recover from frequent anticipated transients, such as loss of offsite power and loss of normal feedwater, than do other NR designs.
Bis, in turn, places a large burden on the plant operators in the event of off-normal system behavior during such anticipated transients.
As a result of a preliminary review of the tree Mile Island Unit No. 2 ac-cident chronology, the tac staff initially identified several human errors that occurred during the accident and contributed significantly to its severity.
All holders of operating licenses were subsequently instructed to take a ntrber of imediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (II).
In addition, the !RC staff began an imediate reevaluation cf the design fea-W
t 7590-01 4 tures of B&W reactors 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 licenseas.
Se evaluation identified design features as discussed above which indicated tnat B&W designed reactors are unusually sensitive to certain off-normal transient conditions originating in the secondary system. As a result, an additional bulletin'was issued by IE which instructed holders of operating licenses for B&W designed reactors to take further actions, including i=ediate changes to decraase the reactor high pressure t. ip point and increase the pressurizer pilot-operated relief valve setting. Also, as a result of this evaluation, the NRC staff identified certain oteqr ss. ety concerns that warranted additional short-tem design and procedural changes at operating facilities having B&W designed reactors. These were identified as items (a) through (e) on page 1-7 of the Office of
- aclear Reactor Regulation Status Report to the Commission of April 25, 1979.
After a series of discussions between r.he NRC staff and the licensee concerning possible design modifications and changes in operating procedures, the licensee agreed in a letter dated April 27, 1979, to perfom promptly the following actions:
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7590-01 (a) Upgrade the timeliness and reliability of delivery from the Auxiliary Feedwater System by carrying out actions as identified in Enclosure 1 of the licensee's letter of April 27, 1979.
(b) Develop and implement operating procedures for initiating and controlling auxiliary feedwater independent of Integrated Control System contrel.
(c)
Implenent a hard-wired control-grade reactor trip that would be actuated on loss oi main feedwater and/or turbine trip.
(d)
Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.
4 (e)
Provide for cne Senior Licensed Operator assigned to the control room who has had ihree Mile Island Unit No. 2 (T*4I-2) training on the B&W simulator.
In its letter the licensee also stated that Rancho Seco would be shut down on April 28, 1979 and wuld remain shut down until (a) through (e) above are completed (The facility was shut down on April 28, 1979 as stated).
In addition
- these modifications to be implenented promptly, the licensee is also proposed to carry out certain additional lorg-term modifications to further enhance the capability and reliability of the reactor *w respord to various transient events. These are:
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I 7590-01 The licensee will provide to the NRC staff a proposed schedule for implementation of identified design modifications whict specifically relate to items 1 throtgh 9 of Enclosure 1 to the licensee's letter of April 27, 1979, and would significantly improve safety.
- The licensee will subnit a failure mode and effects analysis of the Integrated Control System to tha NRC ctaff as soon as practicable.
The licensee 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 to this Order will thereafter be upgraded so that the ecmponents are safety grade. The licensee will subnit this design to the NRC staff for review.
- The licensee will continue operator training and have a minimum of two licensed operators per shift with TMI-2 simulator training at B&W by June 1,1979. Thereafter, at least one licensed coerator with iMI-2 simulator training at B&W wi'.1 be assigned to the control room. All training of licensed personnel will be completed by June 28, 1979.
~he Cor:nission has concluded that the prompt actions set forth as (a) through (e) above are necessary to provide added reliability to the reactor system to respond safely to feedwater transients and should be confi:=ed by a Commission order.
7590-01 The Co::: mission finds that operation of Rancho Seco should rot be resumed until the actions described in paragraphs (a) through (e) above have been satisfactorily completed.
For the foregoing reasons, the Commission has found that the public health, safety and interest regaire that this Order ce effective
' i:rediately.
III.
Copies of the following documents are available for inspection at the Co:=nission's Public Document Room at 1717 H Street, N.W., Washington, D.C.
20555, and are being placed in the Commission's local public document room in the Business and Municipal Department, Sacramento City - County Library, 828 I Street, Sacramento, California 95814:
(1) Office of Nuclear Reactor Regulation Status Report on Feedwater Transients in B&W Plants, April 25, 1979.
(2) Le ter from J. J. Actimoe (SMUD) to Harold Der.*.on (NRR) dated April 27, 1979.
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7590,01 Accordinglv, pursuant to the Atomic Energy Act of 1954, as amended, and the Cor ission's Rules and Regulations in 10 CFR Parts 2 and 50, I.' IS HEREBY CRDERED THAT:
(1) The licensee shall take the following actions with respect to Rancho Seco:
(a) Upgrade the timeliness and reliability of delivery from the Auxiliary Feedwater System by carrying out actions as identified in Enclosure 1 of the licensee's letter of April 27, 1979.
(b) Develop and implement operating procedures for initiating and controlling auxiliary feedwater independent of Integrated Control System control.
(c) Implement a hard-wired cc.T:rol-grade reactor trip that would be actuated on loss of main feedwater and/or turbine trip.
(d) Cc=plete analyses for patential small breaks and develop and implement operating instructions to define operator actio.t.
(e) Provide for one Senior Licensed Operator assigned to the control room who has had Diree Mile Island Unit No. 2 (DtI-2) training on the B&W simulator.
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- (2) The licensee shall maintain Rancho Seco in a shutdown condition (the facility was shut down on April 28, 1979) until itens (a) through (e) in paragraph (1) above are satisfactorily ccepleted.
Satisfactory ccepletion will require confirmation by the Director, Office of Nuclear Reactor Regulation, that the actions specified have been taken, the specified analyses are acceptable, and the specified implenenting procedures are apropriate.
(3) The licensee shall as promptly as practicable 41so accomplish the long-term modifications set forth in Section II of this Order.
V.
Within twenty (20) days of the date of this Order, the licensee or any 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 i.:nediate effectiveness of this Order.
FC" E NUC UIATORY CCMMISSION
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h"a Secretary of mission
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Oated a; Washington, D.C.
this J Q day of May 1979.
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