ML19269E038
| ML19269E038 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/17/1979 |
| From: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| To: | |
| References | |
| NUDOCS 7906230129 | |
| Download: ML19269E038 (7) | |
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UNITED STATES OF AMERICA NUCLEAR REGUIAIORY COMMISSION
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6 In the Matter of y
s ARKANSAS POWER & LIGHT COMPANY )
Docket No. 50-313 4
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.D ARKANSAS NUCLEAR CNE, UNIT 1
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ORDER I.
The Arkansas Power & Light Company (the licensee or AP&L) is the holder of Facility Operating License No. DPR-51 which authorizes the operation of the nuclear power reactor known as the Arkansas Nuclear One, Unit 1 (the facility or ANO-1), at steady state power levels not in excess of 2568 megawatts thermal (rated power). The facility is a Babcock &
Wilcox (B&W) designed pressurized water reactor (PdR) located at the licensee's site in Pope County, Arkansas.
II.
In the course of its evaluation to date of the accident at the Three Mile :.sland Unit No. 2 facility, which utilizes a B&W designed PWR, 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 system. 'Ihe 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 4'
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Page 2 7590-01 secondary side; (2) the lack of direct initiation of reactor trip upon the occurrence of off-normal conditions in the feedwater system; (3) re-liance 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 snaller 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 integrated control system, and the emergency core cool-ing system (E CS) performance to recover from frequent anticipated transients, such as loss of offsite power and loss of normal feedwater, than do other PWR designs. This, 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 'Ihree Mile Island Unit Ib. 2 accident chronology, the NRC 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 number of imediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Commission's 2250 122
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7590-01 Office of Inspection and Enforcement (IE).
In addition, the NRC staff began an immediate reevaluation of the design features of B&W reactors to determine whether additional safety corrections or improvements were necessary with respect to these reactors. 'Ihis evaluation involved numerous meetings with B&W and certain of the affected licensees.
The evaluation identified design features as discussed above which indi-cated that 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 immediate changes to decrease the reactor high pressure trip point and increase the pressurizer 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.
These were identified as items (a) through (e) on page 1-7 of the Office of Nuclear Reactor Regulation Status Report to the Conmission of April 25, 1979.
After a series of discussions between the NRC staff and the licensee concerning possible design. modifications and changes in operating pro-cedures, the licensee agreed in a letter dated May 11, 1979, to perform promptly the following actions:
(a)
Upgrade of the timeliness and reliability of the Emergency Feedwater (EEW) system by performing the items specified in Enclosure 1 of the licensee's May 11, 1979, letter. Changes in design will be submitted to the NRC staff for review.
(b)
Develop and implement operating procedures for initiating and controlling EEW irdependent of Integrated Control System (ICS) control.
(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feedwater and/or on turbine trip.
(d) Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.
(e) At least one Licensed Operator who has had tree Mile Island Unit No. 2 (TMI-2) training on the B&W simulator will be assigned to the control room (one each shift).
In its letter the licensee also stated that ANO-1 was currently shut down and would remain shut down until (a) through (e) above are completed.
In addition to these modifications to be implemented promptly, the licensee has also proposed to carry out certain additional long-term nodifications to further enhance the capability and reliability of the reactor to respond to various transient events. Rese are:
1)
The items in Enclosure 2 of the licensee's letter of May 11, 1979, will be implemented during the next outage (following completion of the design change engineering) to cold shutdown conditions which is of sufficient length to accommodate the change, but no later than the next refueling outage. Further, the licensee will provide a schedule for implementing any other modifications identified as necessary as e result of the licensee's reviews shown on Enclosure 1 of the licensee's letter. The design changes will be subnitted to the NRC staff for review.
2)
The failure modes and effects analysis (EMEA) of the ICS is underway with high priority by B&W and will be subnitted as soon
~
as practicable.
3)
The hard-wired trips addressed in Item (c) above will be upgraded to safety grade. 21s design change will be submitted to the NRC staff for review.
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4)
The licensee will continue operator training and drillirg of response procedures as a part of an ongoing program to assure the hign state of readiness and safe operation at ANO-1.
The Comission 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 Comission order.
The Comission finds that operation of ANO-1 should not be resumed until the actions described in paragraphs (a) through (e) above have been satisfactorily completed.
For the foregoirg reasons, the Comission has found that the public health, safety and interest require that this Order be effective imediately.
III.
Copies of the following documents are available for inspection at the Comission's Public Document Room at 1717 H Street, N. W., Washington, D. C.
20555, and are being placed in the Comission's local public document room at Arkansas Polytechnic College, Russellville, Arkansas:
(1) Office of Nuclear Reactor Regulation Status Report on Feedwater Transients in B&W Plants, April 25, 1979.
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e
. (2)
Letter from William Cavanaugh III (AP&L) to Harold Denton (NRR) dated May 11, 1979.
IV.
Accordingly, pursuant to the Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS HEREBY ORDERED 'nlAT:
(1)
The licensee shall take the following actions with respect to ANO-1:
(a)
Upgrade of the timeliness and reliability of the EEW system by performing the items specified in Enclosure 1 of the licensee's letter of May 11, 1979.
Frovide charges in design for NRC review.
(b)
Develop and implement operating procedures for initiating and controlling EFW independent of Integrated Control System control.
(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feedwater and/or on turbine trip.
(d)
Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.
(e)
Assign at least one Licensed Operator who has had 'IMI-2 training on the B&W simulator to the control room (one each shift).
(2)
'Ihe licensee shall maintain ANO-1 in a shutdown condition until items (a) through (e) in paragraph (1) above are satisfactorily completed.
Satisfactory completion will require confirmation bv the Director, Office of Nuclear Reactor Regulation, that the actions specified have been taken, the specified analyses are acceptable, and the specified implementing procedures are appropriate.
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. (3) 'Ihe licensee shall as promptly as practicable also 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 immediate effectiveness of this Order.
'nlE NUC REGULNIORY COM'4ISSION s..D R
Samuel J.
1 Secretary q'f the Commission Dated at Washington, D. C.
this gEday of May 1979.
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