ML19224B856
| ML19224B856 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/14/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Bickwit L NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| Shared Package | |
| ML19224B857 | List: |
| References | |
| NUDOCS 7906270088 | |
| Download: ML19224B856 (8) | |
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IT.".J? 'J:7':1 FOR: Leonard Bickwit General Counsel sN FRO:':
Harold R. Centc'1, Director Office of ficclea.- Reactor Regulation 1
S'.T ECT:
PGP0 SED ORDER
.b Enclosed is a draf t copy of the proposed order t.hich v ill confirm the co;.mitnants of the Arkcnsas Pc'..er and Light Cornany to shut dc,;n and r.odify its facility and optrating crocedures. Also cnclosed is a copy of relatad corresponden'e from the licensee.
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7590-01 I
l UNITED.S7ATES OF AMERICA NUCLEAR REGULATORY COMMISSION
' f In the flatter of
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ARXANSAS POWER & LIGHT COMPANY
)
Docket No. 50-313
)
ARKANSAS NUCLEAR Ot;E, UNIT 1
)
ORDER I.
The Arkansas Power & Light Company (the licensee or AP&L) is the holder of Facility Operating License No. DPR-51 whicn 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 (PWR) 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 Island 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 ccnditions originating in the secondary system.
The features of the S&W design that contribute to this sensitivity are:
(1) design of the stear generators to operate with relatively small liquid volumes in the 2-v i]
c' n 7 G
7590-01
. t 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 ficw; (4) actuation befcre reactor trip of a pilot-operated relief valve on the prima *y 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.
I Because of these features, B&W designed 'eactors place more reliance on I
the reliability and performance characteristics of the auxiliary feed-water system, the integrated control system, and tha emergency core cool-ing system (ECCS) performance to reccver from frequent anticipated transients, such as loss of offsite power and loss of. normal feedwater, than do other PWR designs. This, '- turn, places a large burden on the plant operators in the event of off-normal system behavior during such anticipated transients.
j A s
- c. result of a preliminary review of the Three Mile Island Unit tio. 2 accident cnronology, the tGC staff initially identified several human errors that occurred during the accident and contributed significantly to its severity. All nciders of operating licenses were subsequently instructed to take a number of irmediate actions to avoid repetition of-these errors, in accordance with bulletins issued by the Commission's 255 2R4 e
i 7590-01 3-l
. l Office of Inspection and Enforcement (IE).
'n 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. This evaluation involved numerous meetings with B&W and certain of the affected licensees.
The evaluat1on identified design features as discussed above which indi-i cated that B&W designed re'ctors 3re unusually sensitive to certain off-normal transient conditions originating in the secondary system. As a i
~
result, an additional bulletin was issued by I "hich instructed holders of operating licenses for B&W designed reactor-take further actions, I
l including immediate changes to decrease the reat.or high pressure trip I
point and increase the pressurizer pilot-operated relief vdive 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 Commission of April 25, 1979.
Af ter a series of discussions between the NRC staff and the licensee concerning possible design modifications and changes in cperating pro-cedures, the licensee agreed in a letter dated May 11, 1979, to perform prcr.ptly the following actions:
255 295
. (a) Upgrade of the timeliness and reliability of the Emergency Feec* water (EFW) system by performing the items specified in Enclosure 1 of the licensee's Iiay 11, 1979, letter. Changes in design will be submitted to the NRC staff for review.
(b) Develo9 and implement operating procedures for initiating and controlling EFW independent of Integrated Control System (ICS) control.
(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feecwater and/or on turbine trip.
(d) Ccmplete analyses for potential small breaks and develop and implement operating instructions to define operator action.
(e) At least one Licensed Operator who has had Three Mile Island Unit No. 2 (TNT-2) training on the B&W simulator will be assigned to the control rova (one each shif t).
In its letter the licensee also stated that ANO-l was currently shut down and would remain shut down until (a) through (e) above are completed.
In addition to these modificaticns to be implemented prcmptly, the licensee has also proposed to carry out certain additional long-term modifications to further enhance the capability and reliability of the reactor to respond to various transient events.
These are:
1)
The items in Enclosure 2 of the licensee's letter of May 11, 1979, will be implemented during the next outage (following canpletion of the design change engineering) to cold shutacwn conditions which is of sufficient length to acccmmodate the change, but no later than the next refueling outage.
Further, the licensee will provide a schedule for implementing any other nodifications identified as necessary as a result of the licensee's reviews shown on Encles tre 1 of the licensee's letter.
The design changes w1il be submi med to the NRC staf f for review.
2)
The f ailure modes and effects analysis (FUE") of the ICS is uncernay wi th high priority by 3&W and will be submitted as soon as practicable.
3)
The hard-wired teips addressed in Item (c) above will be upgraded to safety grace.
This design change will be submitted to the NRC 5taff for review.
26I c:4o-1r J
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7590-01 i.
4)
The licensee will continue operator training and drilling of response procedures as a part of an ongoing program to assure the high state of readiness and safe operation at ANO-1.
The Commission has concluded that the prompt actions set forth as (a) throuch (e) above are necessary to provide added reliability to the i
react?" system to respond safely to feedwater transients and should be confirmed by a Commission order.
The Commission finds that operation of ANO-1 should not be resumed until the actions cescribed in paragraphs (a) through (e) above have been satisfactorily completed.
For the foregoing reasons, the Commission has found that the public l
health, safety and interest require that t:iis Order be effective n
[
immediately.
III.
Copies of the following documents are available for inspection at the i
Commission's Public Document Room at 1717 H Street, N. W., Washington, D. C.
20555, and are being placed a Commission's local public document room at Arkansas Polytechnic College, Russellville, Arkansas:
(1) Office of Nuclear Reactor Regulation Status Report on Feecwater Transients in 23W Plants, April 25, 1979.
2' O s65?C7 L.
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7590-01 (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 HEREBI ORDERED THAT:
(1) The licensee shall take the following actions with respect to ANO-1:
(a)
Upgrade of the timeliness and reliability of the EFW system by performing the items specified in Enclosure 1 of the licensee's letter of May 11, 1979.
Provide changes 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 a feedwater and/or on turbine trip.
(d) Complete analyses for pctential small breaks and develop and implement operating instructions to define operator action.
(e) Assign -at least one Licensed Operator who has had TMI-2 training on the B&W simulator to the control room (one each shift).
(2) The licensee shall maintair ANO-l in a shutdown condition until items (a) through (e) in paragr aph (1) above are satisfactorily completed.
Satisfactory completien wil' require confirmation by the Director, Of fice of Nuclear Recctcr Regulation, that the actions specified have been taken, the specified analyses are acceptible, and the speci fied implementing procecures are appr priate.
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7590-01 7
(3) The 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 whos'e interest may be affected by this Orcer cay request a hearing with respect to this Order. Any such request shall not stay the immediate effectiveness of this Order.
FOR THE NUCLEAR REGULATORY COMF.ISS10!1 Samuel J. Chilk Secretary of the Commission Dated at Washington, D. C.
this day of May 1979.
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9 ARKANSAS POWER & LIGHT COMPANY POST CFRCE BCX 551 LITTLE ROCK. AAKANSAS 72203 (50t 371-v.22 May 11, 1979 WILLIAM CAVANAUGH ll1 V.ce President Genera on & Cons::ve:c, 1-059-16 Mr. Harold R. Denton Director, Nuclear Reactor Regulaticn Uni ted States Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 (File:
1510)
Dear Mr. Denton:
lhis letter provides additional information requested by your staff and supercedes our May 10, 1979 letter.
In response to the staff safety concerns identified as items a. through e.
on pages 1-7 on ONRR Status Report to the Commission of April 25, 1979, Arkansas Power and Light proposes the following actions:
(a) Upgrade of the timeliness and reliability of the Emergency Feedwater (EFW) system by performing the items specified in Enclosure 1.
Change.; in design will be submitted to the Nuclear Regulatory Commission for review.
(b) Develop and implement operating procedures for initiating and controlling EFW independent of Integrated Control System (ICS) control.
(c)
Implement a hard-wired control-grade reactor trip on loss of main feedwater and/or on turbine trip.
(d) Complete analyses for potential small breaks and develcp and implement cperatina instructions to define operator action.
See Enclosure DUPLICATE DOCUMENT (e) At least one Licensed inc on the B&W simulat Entire document previously entered room (one each shift).
into system under:
b ANO 7'95 J ']
no of vases:
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