ML19289F109
| ML19289F109 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/16/1979 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | |
| References | |
| NUDOCS 7906020060 | |
| Download: ML19289F109 (8) | |
Text
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" #GLIC DOCU. LENT ROOM 7590-01 t.NITED STATES CF AMERICA NUCLEAR REGUUCDRY CCrMISSICN D)
N In the Matter of
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bmo FI4RIDA POWER CCRPORATION, ET AL )
Docket No. 50-302 p.
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Crystal River Unit No. 3 E
- .n.a+1. d YO Nuclear Generating Plant
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ORDER j
I.
Florida Power Corporation (FPC or the licensee) and eleven other co-owners are the holders of Facility Cperating License No. CPR-72 which authorizes the operation of the nuclear power reactor known af Crystal River Unit No. 3 Nuclear Generating Plant (the facility or Crystal River Unit 3), at steady state power levels not in excess of 2452 megawatts thermal (rated power). The facility is a Babcock & Wilcox (B&W) designed pressurized water reactor (PWR) located at the licensees' site in Citrus County, Florida.
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 PWR, the Nuclear Regulatory Comission staff has ascertained that B&W designed reactors appear to be unusually sensitive to certain off-normal transient conditions originating in the secondary system. The 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 secondarf side; (2) the lack of direct initiation of reactor trip upon the 2236 232 Y906020ggy
. occurrence of off-normal conditions in the feedwater sys+m; (3) reliance on an integrated control system (ICS) to automatically regulate feedwater flow; (4) actuation before reactor trip of a pilot-operated relief valve en the primary system pressuris.er (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 performance characteristics of the auxiliary feedwater system, the integrated centrol system, and the energency core cooliry system (ECCS) performance to recover from frequent anticipated transients, such as loss of offsite power and loss of normal feedwater, than do other IHR designs.
'Ihis, 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 Bree Mile Island Unit No. 2 accident chronology, the NRC staff initially identified several htrnan errors that occurred during the accident and contributed significantly to its severity. All holders of operati:q licenses were subsequently instructed to take a ntrber of immediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (IE).
In additien, the NRC staff began an iscediate reevaluation of the design features of B&W 2236 233
7590-01 reactors to deter:nine whether additional safety corrections or i=provements were necessary with respect to these reactors. This evaluation involved numerous meetings with B&W aM certain of the affected licensees.
The evaluation 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 tulletin was issued by II which instructed holders of operating licenses for B&W designed reactors to take further actions, including immediate changes to decrease the reactor high pressure trip point and increase the pressuri::er 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 operatire facilities having B&W designed reactors. These were identified as items (a) through (e) on page 1-7 of the Office of Nuclear Reactor Regulation Status Report to the Coranission of April 25, 1979.
After a series of discussions between the NRC staff and the licensee concerning possible design redifications a-d cha x;es in operating procedures, the licensee agreed in a letter dated.v y 1,1979, to perform prcmptly the following actions:
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7590-01
, (a) Upgrade the timeliness and reliability of delivery frcct the Emergency Feedwater System by carryirzg out actions as identified in Enclosure 1 of the licensee's letter of May 1,1979.
(b) Develop and implement operating procedures for initiating and controlling emergency feedwater independent of Inte-grated Control System control.
(c)
Implenent a hard-wired control-grade reactor trip that would be actuated cn loss of main feedwater and/or turbine trip.
(d)
Complete analyses for ptential small breaks and develop and i=plement operatirg instructions to define operator action.
(e) A1.1 licensed reactor operators and senior reactor operators will have completed the Three Mile Island Unit No. 2 (TMI-2) simulator training at MW.
In its letter the licensee also stated that the facility is shut down and would remain shut down until (a) through (e) ateve are completed.
In addition to these nodifications to be implemented premptly, the licersee has also proposed to carry out certain additional long-term modifications *w further enhance the capability and reliability of the reactor to respond *w various transient events. "bese are:
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7590-01 The licensee will make modifications to provide verification in the control room of energency feedvater flow to each steam generator.
The licensee will submit a failure mode and effects analysis of the Integrated Control System to the NRC staff as soon as prac-ticable. 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 compnents are safety grade.
We licensee will submit this design to the NRC staff for review.
The licensee will continue reactor operator training and drilling of response procedures to assure a high state of preparedness.
The Connission 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 confirmed by a Commission order.
We Comission finds that operation of the facility should ret be resumed until the actions described in paragraphs (a) though (e) above have been satisfac*wrily completed.
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7590-01 For the foregoing reasons, the Comission has found that the public health, safety and interest require that this Order be effective imediately.
I"I.
Copies of the followirg documents are available for inspection at the Comission's Public Docunent Room at 1717 H Street, N.W., Washington, D.C.
20555, and are being placed in the Comission's local public document room in the Crystal River Public Library, Crystal River, Florida, 32629:
(1)
Office of Nuclear Reactor Regulation Status Report on Feedwater Transients in B&W Plants, April 25, 1979.
(2) Letter frem B. L. Griffin (FFC) to Harold Denton (NRR) dated May 1, 1979.
IV.
Accordingly, pursuant to the Atccic Energy Act of 1954, as amended, and the Comission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS HERE3Y CRDERED THAT:
(1) The licensee shall take.the following actions with respect to Crystal River Unit 3:
(a) Upgrade the timeliness and reliability of deliver'f from the Emergency Feedwater System by carryiry out actions as identified in Enclos'ure 1 of the licensee's letter of.%ay 1, 1979.
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7590-01 (b)
Develop and implement operating procedures for initiating and controlling emergency feedvater independent of.~.nte-grated Control System control.
(c)
Implement a hard-wired control grade reactor trip that would be actuated on loss of main feedwater and/or turbine trip.
(d)
Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.
(e)
All licensed reactor operators and senior reactor operators will have completed the TMI-2 simulator trainirg at B&W.
(2) te licensee shall maintain Crystal River Unit 3 in a shutdown condition (the facility was shut down on April 23, 1979) until items (a) through (e) in paragraph (1) above are satisfactorily completed.
Satisfactory completion will require confirmation by the Director, Office of Nuclear Reactor Regulation, that the actions specified have been taken, the specified aralyses are acceptable, and the specified i=plementing procedures are appropriate.
(3)
De licensee shall as promptly as practicable also acecmplish the long-term modifications set forth in Section II of this Order.
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7590-01 V.
Within twenty (20) days of the date of this Order, the licensees or any persen ese interest may be affected by this Order may request a hearing with respect to this Order.
Any such request shall not stay the Lraediate effectiveness of this Order.
FCR nlE NUCLEAR REGUIAERY CCMMISSION
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/, _ wm.
muel J.
1 Secretary o the Commission Dated at Washington, D.C.
this /d4 day of May 1979.
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