ML19224C554
| ML19224C554 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/21/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Cavanaugh W ARKANSAS POWER & LIGHT CO. |
| References | |
| FOIA-79-98 NUDOCS 7907020514 | |
| Download: ML19224C554 (20) | |
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Docket No.:
50-313 lN l
Mr. William Cavan: ugh, III Vice Presiden*,, Gereratier and Con?truct ~
Arkansas Power & L. p t Q s ny P.O. Box 551 Little Rock, M 72293 Dear Mr. Cavanaug'-
By Order of May l', 'Y79, the Commission confirmed your undertaking a series of actions, both immediate and long term, to increase the capability and reliability of Arkansas Nuclea.' One, Unit No. 1 (ANO-1) to respond to various transient events.
In addition, the Order confirmed that ANO-1 was shutdown and would not be restarted until the following actions had been accomplished:
(a) Upgrade of the timeliness and reliability of the Emergency Feedwater System (EFW) by performing the items specified in Enclosure 1 of tha licensee's letter of May 11, 1979.
Provide changes in design for NRC review.
(b) Develop and imp ement operating procedures for initiating and con-trolling EFW indeg ndent 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.
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(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 Three Mile Island, Unit No. 2 training on the Babcock & Wilcox simulator to the control rocm (one each shift).
By submittal of May 17, 1978, as supplemented by letters dated May 21, 22, 23 and 24, 1979, you have documented the actions taken in response to the May 17 Order.
I have reviewed this submittal, and am satisfied that, with respect to ANO-1, you have satisfactorily completed the actions prescribed in items (a) M.
through (e) of paragraph (1) of Section IV of the Order, the specified mofi(Q ica-tions and analyses are acceptable, and the specified implementing procedures are apprcpriate.
The bases for these conclusions are set forth in the
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enclosed Safety Evaluation.
272 225 7907020 80IS/HAAI n-W
4 Arkansas Power & Light Company '
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'- ~ ~ As-d ted on page 5 of Safety Evaluation you will be required to conduct a test during power operation to demonstrate operator capability to assume manual control of the EFW system independent of ICS.
In addition, we have discussed
',d the need for monitoring core exit temperature with ycur staff and they have_
provided 16 thermocouple indications of core exit temperature in the control room.
Also, your staff has agreed to provide an additional 16 thermocouple indications of core exit temperature in the control room by October 31, 1979.
Appropriate Technical Specifications-for Limiting Conditions for Operation and for surveillance requirements should be developed as soon as practicable and provided tu the staff within seven days with regard to the design and procedural changes which have been completed in compliance with the provisions of the May 17, 1979 Commission Orf '. The revised Technical Specification should cover:
(1) Changes to the EFW Systems; (2) Plant alignment changes made to ensure control of the EFW independent of the ICS; (3) Addition of the Anticipatory Reactor Trip; and (4) EFW capacity.
We note that by letter dated April 24, 1979, you have submitted proposed Technical Specifications for changes in setpoints for high pressure reactor trip and pilot operated relief valve actuation.
Also by letter dated May 16, 1979 you have submittsd proposed changes to the Technical Specifications which define limiting conditions of operation upon
?oss of EFW equipment.
Within 30 days of receipt of this letter, you should provide us with your
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schedule for completion of the long-term modifications described in Section ii of the May 17 Order.
My finding of satisfactory compliance with the requirements of items (a) through (e) of paragraph (1) of Section IV of the Order will permit resumption of operation in accordance with the terms of the Commission's Order; it in no way affects your duty to continue in effect all of the above provisions of the 272 2
Arkansas Power & Light Company '
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Orde'r pending your submission and approval by the Commission of the Technical Specification changes necessary for each of the required modifications.
Sincerely, Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
Notice of Authorization to Resume Operation cc w/ enclosure (s):
Phillip.4.
Lyon, Esq.
Director, Technical Assessment House, Holms & Jewell Division 1550 Tower Building Office of Radiation Programs Little Rock, AK 72201 (AW-459)
U.S. Environmental Protection Mr. David C. Trimble Agency Manager, Licensing Crystal Mall #2 Arkansas Power & Light Company Arlington, VA 20460 P.O. Box 551 Little Rock, AK 72203 U.S. Environmental Protection Agency Mr. James P. O'Hanlon Region VI Office General Manager ATTN:
EIS C0ORDINATOR Arkansas Nuclear One 1201 Elm Street P.O. Box 608 -
First International Building Russellville, AK 72801 Dalla, TX 75270 Mr. William Johnson Director, Bureau of Environmental U.S. Nuclear Regulatory Commission Health Services P.O. Box 2090 4815 West Markham Street Russellville, AK 72801 Little Rock, AK 72201 i.r. Robert B. Borsum Babcock & Wilcox declear Poer Generation Division Suit 420, 7735 Old Georgetown Road Bethesda, MD 20014 6
Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Pennsy1/ania Avenue, N.W.
Washington, DC 20006 22[
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Arkansas Pcwer & Light Company,x 4
A.-kansas Polytechnic College Ressellville, AK 72801
'lono;able Ermil Grant r ting County Judge of Pope County e
hpa County Courthouse
ussellville, AK 72801
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EVALUATION OF LICENSEE'S COMPLIANCE WITH THE NRC ORDER DATED PAY 17, 1979 ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 INTRODUCTION By order dated May 17, 1979, (the order) the Arkansas Power & Light Company (AP&L or the licensee) was directed by the NRC to take certain actions with respect to Arkansas Nuclear One, Unit 1.
Prior to this order and as a result of a preliminary review of the Three Mile Island Unit No. 2 accident, the NRC staff initially identified several human errors tnat contributed significantly to the severity of the event.
All holders of operating licenses were subsequently instructed to take a number of immediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (IE).
Subsequently, an addi-tional 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 power-operated relief valve (PORV) setting.
The NRC staff identified certain other scfety concerns that warranted additional short-term design and procedural changes at operating facilities having B&W designed reactors.
Those were identified as items (a) through (e) in page 1-7 of the Office of Nuclear Reactor Regulation Status Report to the Commission on April 25, 1979.
After a series of discussions between the NRC staff and the licensee concerning possible design modifications and changes in operating procedures, the licensee agrecd in a letter dated May 11, 1979 to perform promptly certain actions.
The Commission found that operation of the plant should not be resumed or continued on an indefinite basis until actions described in paragraphs (a) through (e) c.f. paragraph (1) of Section IV. -
of. the order were satisfactorily completed.
Our evaluation of the licensee's compliance with items (a) through (e) of paragraph (1) of Section IV of the order is given below.
In performing this evaluation we have utilized additional information provided by the licensee on May 11, 16, 17, 21, 22, 23, 24, and 29, 1979 and numerous discussions with the licensee's staff.
Confirmation of design and procedure changes was made by members of the NRC staff at the ANO-1 site. An audit of the ANO-1 reactor operators was also performed by the NRC staff to assure that the design and procedure changes were understood and were being correctly implemented by the operators.
EVALUATION Item a It was ordered that the licensae take the following action; 272 229
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" Upgrade of the:timeliiless and reliability of the EN system bf rierforming the
,_ _ _ items'specified in Enclosure 1 of the liceme s letter of May 11, 1979."
The ANO-1 design has one turbine-driven emergency feedwater (EFW) pump that is automa-tically actuated and controlled independent of offsite power, and one motor-driven EFW pump that must be manually transfeired to a vital AC bus if offsite power is lost: By reference above to Enclosure (1) of the licensee's letter of May 11, 1979, it was ordered that the licensee;
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"1.
Review procedures, revise as necessary and conduct training to ensure timely and proper starting of motor driven emergency feedwater (EFW) pump from an engineered safeguards bus upon loss of offsite power.
Conduct a test of the manual startup of the motor driven EFW pump from a vital AC power supply."
Tests were conducted by the licensee and witnessed by a member of the NRC staff.
The test described in Item 1 above was conducted four times.
During the conduct of the first test to transfer to a vital AC power supply, a breakdown in communication be-tween the two operators performing the test resulted in a skipped step in the test procedure.
A second test was then successfully performed in less than five minutes.
However, the NRC staff subsequently required that the licensee repeat the test a third time, using the actual procedure available in the control room instead of ths test procedure.
This control room procedure was reviewed and modified at our request prior to the third test which was conducted subsequent to the addition of automatic start circuitry described in Part 6.
The results of this third test were incomplete due to a feature built into the new automatic start design of the motor-operated EFW pump which required an additional manual switching operation not previously included in the emergency procedure.
The procedure was again revised and the fourth test conducted satisfactorily within five minutes.
Subsequently, the design of the automatic start circuitry was modified so as to not require this additional manual switching operation, and the procedure was changed accordingly.
Members of the NRC staff on site have verified that the control room operators are properly trained to carry out this revised procedure.
The licensee has also agreed to have two operators stationed in the control room at all times until the electric dri' Ten EFW pump is permanently
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connected to vital power.
Since the time frame of five minutes is well within the. -
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allowable delay of 20 minutes indicated by the generic B&W analyses discussed in Item (d), we conclude that the licensee has complied with the requirement for demonstrating manual startup of the motor-driven EFW pump from a vital AC power supply.
It was also ordered that; "2.
To assure that EFW be aligned in a timely manner to inject on all EFW demand events when in the surveillance test mode, procedures will be implemented and training conducted to provide an operacor at the necessary valves in communication with the control room during the surveillance made to carry out the valve alignment changes upon EFW demand events."
The ANO-1 staff has revised OP 1106.06 " Emergency Feedwater Pump Operation."
Supplements I and II provide procedures for conducting the Electric and Steam Driven 272 230
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' Emergency Feedwater Pump'siTrveillance t'est, re'spectively. ' The NRC staf f his Yev%
.these procedures which require in part; " Operator shall remain in area for duration of test in c~ommunication with the control room to align system in the event of an EFW demand."
The NRC staff has also determ ned' that training of operators in use of this, j procedure has been conducted and 1.s adequate.
Subject to confirmation by a member of the NRC staff that noise levels in this area during plant operation are conducive to ccmmunications with the control room, we conclude that the licensee has complied with the order.
It was also ordered that the licensee; y
"3.
Write and implement procedures for the manual initiation and control of the EFW System following failure of the Integrated Control System."
The licensee has revised OP 1106.06 (Emergency Feedwater Pump Operation) and this procedure has been reviewed by the NRC staff.
This procedure provides operator guid-ance concerning manual initiation and control of the EFW System following failure of the Integrated Control System.
The procedures were reviewed by the NRC staff t, assure that feedwater from both the motor-driven pump and the steam-driven pump woulo be available in a timely manner.
The procedures provide for verification of pump start, either automatic or manual.
If offsite power is not available to the motor-driven pump, EP 1202.05 (Degraded Power) provides operator guidance to provide diesel generator power for this pump.
If manual intervention to control cooldown rate is required, procedures provide for initiation and control of emergency feedwater flow through the bypass valves.
These procedures would be implemented by the operator in the event of failure of the Integrated Control System.
Specific procedural steps provide for:
Startup of the electric driven EFW pump (including procedures to provide power supply from the diesel generator, if normal offsite power is not available).
Startup of the steam driven EFW pump by cpening the steam supply valves.
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Closing the ICS-controlled EFW valves (using the control room handswitch).
Opening, and modulating as necessary, the emergency feedwater bypass valves to control EFW to the steam generator (using their control room handswitches).
Verifying system operation by observation of EFW flow, EFW pump discharge pressure, steam generator pressure, and steam generator level.
We have reviewed these revised procedures for manual initiation and control of the EFW system and conclude that there is sufficient guidance to the operator to perform these actions to control and maintain level in the steam generators to specified values.
In addition, the NRC staff required that a test be conducted to demonstrate the capability to provide and control emergency flow to the steam generators.
The licensee has committed to perform a test at low power operation (10-15%) during power 272 231 n rsion.
The primary objective of the _ test will be to further verify thA capability
' o mnuzily control steam generator level li1 dependent of ICS.
A member of the NRC
TT n tne ANO-1 site will witness the test and will verify acceptance prior to
,iroceding to full power operation.
Subject.to the successful completion of this test, va conclude that the licensee has complied with this portion of the order.
't was also ordered that;
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The EFW pumps will be verified operable in accordance with the ANO-1 Technical Specifications and Surveillance Procedures."
rh3 gn-1 Technical Specifications provide for EFW surveillance and limiting condi-
~itns cf uperation.
Consistent with the cover letter for this evaluation, the NRC
.. u r v.iii raceive from the licensee within seven days revised proposed Technical 7,scifications with regard to design and procedural changes.
It was also ordered that the licensee; "R
Feview and revise, as necessary, the procedures and conduct training for providing alternate sources of water to the suction of the EFW pumps.'
The means available to alert the operator to perform the manual transfer of EFW from A condaosate storage tank (CST) to the service water system consists of an alarm in t.ha centrol room which annunciates on low EFW pump suction pressure.
The licensee has
- cJc '. ml annunciation in the control room on icw level in the condensate storage This new feature allows direct control room annunciation that is redundant to
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the enisting low suction pressure switch annunciation.
The NRC staff reviewed procedure DP 1106.05 " Emergency FW Pump Operation" and requested revision of the guidance to the 7>r.',or for providing alternate sources of water to the suction of the EFW pumps.
The revision has been made to provide additional guidance to the operator for alternate reens cf verifying low level in the condensate storage tank.
The NRC staff at the
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', verified that the control room operators are properly trained to carry out
~c procedures.
We conclude that the licensee has-cgmplied with the requirements.to-itview and revise procedures and has conducted operations personnel training for nr" W ng alternate sources of water.
It was also ordered that; "S.
In the event emergency feedwater is necessary and offsite power is available, an auto star signal will be providad to the motor driven emergency feedwater pump."
% li ensee has installed an automatic start of the motor-operated EFW pump on loss of all RC pumps or loss of both main fLdwater pumps.
Relay contacts associated with
isting relays within the integrated control system cabinet, additional relays and contacts, and wiring are arranged in the fina? Tctuation control circuitry for the notor-driven emergency feedwater pump such tho-, if offsite power is available, the m tor is provided a signal to start automatically.
Further, manual capability to 4-272 232
_ initiate and/or override this automatic circuitry is included in the design.
In addition, annunciation within the control room has been provided whenev~er this pump is
.s_ta.rted by the automatic circuitry.
Based on our review of this aspect of the design, we conclude that it is in accordance with the order.
It was also ordered that;
" 7.
Procedures will be developed and implemented and training conducted to provide guidance for timely operator verification of any automatic initia-tion of EFW."
The licensee has revised procedure OP 1106 3 (Emergency Feedwater Puap Operation) to provide specific operator guidance as to the methods for confirming automatic initia-tion of EFW.
This includes:
Verification that pump discharge pressure is greater than OTSG pressure.
Verification of feedwater flow (on the flow indicator installed pursuant to Part 9, below).
Observation of steam generator levels.
Emergency procedures for plant transients requiring initiation of emergency faedwater (such as loss of normal feedwater or loss of reactor coolant flow) require the operator to verify the initiation of emergency feedwater.
Additionally, the operator is required to observe alternate instrumentation channels to provide further assurance.
The NRC staff has confirined that control room operators are properly trained to carry out these procedures.
It was also ordered that;
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VerificatidnthatTechnicalSpecificationrequirementsforEFWcapacityare[
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in accordance with the accident analysis will be conducted."
The licensee has stated that a minimum flow of 550 gpm is required to support the.
accident analyses.
Low power testing will substantiate the availability of at least this flow capacity by em:h EFW train (see Part 3).
Consistent with the cover letter to this evaluation, we vill require submittal of a Technical Specification change concerning EP4 capacity.
This change will be a limiting condition of reactor opara-tion in the event the minimum allowable value assumed in the accident analysis is not met, and will provide for periodic surveillance.
It was also ordered that; "9.
Modifications will be made to provide verification in the control room of EFW flow to each steam generator." 2/2 233
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'To verify that emergency ~ feedwater is being putnped to the " steam ger, ors', the licensee
.is providing two orifice plates and differential pressure sensing equipment.
These flow devices will be installed on each of the EFW injection flow paths downstream of the crossover line, so that flow to each ste"am generator will be measured.
The output of the differential pressure transmitter will be displayed in the control room, indicated in gallons per-minute.
A verification test will be performed to assure performance of this design modifi-
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cation.
This will be performed as part of the test described in Part (3) in this report.
The test procedure has been reviewed by the NRC staff and verified as acceptabie.
It was also ordered that the licensee; "10. Provide a means of notification to the control room tnt the EFW system has auto started.
This notification can be provided frer, a temporary modifica-tion or a dedicated operator."
As described in Part 7, above, the control room operator can determine the initiation of emergency feed by observation of pump discharge pressure (as compared to steam generator pressure), emergency feed flow, and steam generator level.
In addition, annunciation has been provided in the control room whenever either pump is automa-tically started. Based on our review of this design, we conclude that it is in acco rdance with the order.
Item b It was ordered that the licensee;
" Develop and implement operating procedures for initiating aid controlling EFW independent of Integrated Control System (ICS) control."
Several components in each EFW train are provided with ari automatic initiation signal..
Fo~ur cornpanents in one train are one steam-driven pump centroller, one motor operated valve located at the discharge of this pump, and two motor-operated valves associated with the steam supply for this turbine-driven pump.
Two components for the other EFW train are the motor-driven pump and one motor-operated valve at the pump discharge.
Although the automatic actuation signal is provided by common circuitry within the integrated control system cabinet, provisions exist to aanually control these com-ponents from the control room.
This manual provision provides overriding control of the automatic signal (from the Integrated Control System cabinet).
We conclude that manual means exist in the design whereby the operator can initiate and control emer gency feedwater following failure of the Integrated Control System automatic initia-tion circuitry.
We have reviewed the revised procedures for the emergency feedwater system to assure that there is sufficient guidance to the operator to actuate the system if the automatic 272 234
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' nitiation fai. led and~ to'c6iitrol the steam generator level'to s;iecif'ie'd varues' ';The i
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.rev.iew.ofd he procedures focused on whether the operator was directed to observe the proper instruments and whether the operator was given specific values of parameters, such as steam generator level, to maintain b~y operating controls.
The review also determined that the operator should confirm the validity of the instrument readings of certain key parameters such as steam generator level.
The necessary modifications ~ to the procedures to satisfy thece determinations were preserted to tne licensee, and the NRC staff has verified that the modifications have been incorporated in the procedures.
(See further discussion of these procedures and test requirements in Part 3 of Item a).
The NRC staff at the ANO-1 site walked through the emergency feedwater procedures with ANO-1 operators to evaluate whether the procedures were functionally adequate.
In addition, the NRC staff audited a sample of ANO-1 operators to determine if they were familiar with the revised procedures and could implement them correctly.
Based on the NRC staf f audit, we conclude that the revised procedures and operator training are satisf actory.
Item c The order requires that the licensee;
" Implement a hard-wired control ~ grade reactor trip that would be actuated on loss of main feedwater and/or on turbine trip."
The Arkansas Nuclear One Unit 1 original design did not have a direct reactor trip frcm a malfunction in the secondary system (loss of main feedwater and/or turbine trip).
To obtain an earlier reactor trip (rather than delaying the trip until an operator i.cok action or until a primary system parameter exceeded its trip setpoint),
the licensee committed to install a hardwired control grade reactor trip on the loss of all main feedwater and/or on turbine trip (letter from William Cavanaugh III (AP&L) to H. Denton (NRC) da,ted May 11, 1979).
The purpose of this anticipatory trip is to minimize the potential for opening of the power-operated relief valve (PORV) and/or the safety valves on the pressurizer.
The licensee has indicated that this new tircuitry meets this objective by providing a reactor trip during the incipient stage of the related transients (turbine trip and/or loss of main feedwater).
AP&L has added control grade circuitry to ANO-1 which is designed to provide an automa-tic reactor trip when either the main turbine trips or both of the two main feedwater pumps trip.
The main turbine trip is sensed by a normally de energized auxiliary relay associated with the main turbine Electro-Hydraulic master trip circuitry.
The power for this circuitry is provided from a Class 1E 125 volt direct current bus by way of a 125 volt distribution panel.
A contact from this auxiliary relay is arranged into a 118 volt alternating current circuit containing a normally de-energized relay.
This alternating current relay is physically located within the Integrated Control System cabinet and is provided power from the associated Integrated Control System power supply.
A contact from this alternating current relay is arranged into a 272 235
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-% energi.ted 24 volt direct current ci cuit containing two additional re.l c:
24 volt power su'pply is derived withi the Integrated Control Syst'em i^abine To
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- n rb'bf the breakers and trip the re tor, two associated direct current relays rcvide four contact closures to energized two direct current shunt coils (two contact r
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closures per shunt trip coil and one shunt trip coil for each of the two reactor trip al ternating current circuit breakers).
Power is provided to the shunt trip coils _from (1 css 1E 325 volt direct current buses.
ire maia.'eedwater pump trip is sensed by two normally de-energized auxiliary relays
'ss,ci:ted with the main feedwater pumps master trip circuitry (one relay associated
..,. e d) cf the two main feedwater pumps).
The remaining circuitry associated with this trip is identical to that described above for the turbine trip including power
'u')lin, with the exception that two corresponding relays and contacts are provided.
/1 30, ttv: twu associated contacts (these contacts are arranged in parallel) within the ci volt direct current circuit are in series with the associated turbine trip contact.
Provisions have been included to automatically bypass and re-instate these additional trip at Icw power to allow a normal startup and shutdown.
Operatar verification of
'N Syr.a removal is required procedurally during power escalation.
The NRC staff at 9 ?. W i site audited a sample of ANO-1 operators and concluded that they were familar
,f M '.mctions of these trips and associated procedural requirements.
Ite i'censee has analyzed this additional circuitry with respect to its independence from the existing reactor trip system.
The licensee has stated that the shunt coil is gr c; U.e existing AC reactor trip breaker.
However, it is separate and operates inaeaencently from the 120 volt alternating current undervoltage trip coil of the
- ssociated bceaker.
The reactor trip safety grade signal de-energizes the 120 VM t alterndog current undervoltage coil to produce a trip of the associated alternating c.. wr.t :ceaker.
Casad on our review of the implementation of the trip circuitry with respect to its
-'- 1'ce from the existing trip circuitry, we conclude that this addition will not
- w. m a the existing. reactor protection system desigo.,The licensee has installed a.nd-
- cg.pleted checkout of the trip circuitry.
ine licensee has ccmmitted to perform a monthly periodic test on the added circuitry to demonstrate its ability to open the AC circuit brt.akers (tripping the AC breakers
. 'a tne shunt trip circuit).
Additionally, the licensee has committed to perform a r..cce cc...plete test of this additional circuitry whenever the reactor is brought to a
':' ;huMewn conoition as the result of a normal outage or reactor trip (but not more frequently than once per 31 days). We conclude that there is reasonable assurance that the additional circuitry will perform its function.
Accordingly, on the basis of
'": h a, we conclude that this additional circuitry is ir. accordance with the requirements of item (c) of the order.
_g.
272 236
. Item d
.This_ item 9n the order requires the licensee to:
" Complete analyses for potential small treaks and implemen; operating instruc-tions to define operator action."
By letter from William Cavanaugh III (AP&L) to H. Denton (NRC) dated May 11, 1979, the license committed tn providing the analyses and operating procedures of this requir ement.
Babcock and Wilcox,' the reactor vendor for the ANO-1 plant, submitted an analysis entitled, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" and supplements to this analyses (References 1 through 6).
The major parameters used in this gene ic study, with the exception of emergency feedwater flow, conservatively bound the AN-1 plant.
An additional analysis assuming a bounding value for emergency feedwater flow was subsequently submitted (Reference 6).
In a letter dated May 16, and 22, 1979, AP&L has referenced these analyses as appropriate for ANO-1.
The staff evaluation of the B&W generic study has been com-pleted and the results of the evaluation will be issued as a NUREG report in June 1979.
A principal finding of our generic review is a reconfirmation that loss-of-Coolant Accident (LOCA) analyses of breaks at the lower end of the small break spectrum (smaller than 0.04 sq. ft!) demonstrate that a combination of heat removal by the steam generator 3, high pressure injection system and operator action ensure adequate core cooling.
The emergency feedwater system used to remove heat through the steam generators has been modified to enhance its reliability as discussed in item (a).
The high pressure injection system is capable of providing emergency core cooling even at the safety valve pressure setpoint.
Reactor cara u'covery is not predicted for these events.
The cJ culated peak cladding temperature was less than 800 F, well below the 10 CFR 50.46 requirement of 2200 F.
The ability to remove heat via the steam gener-ators has always been recognized to be an important consideration when analyzing very small breaks.
Sensii.ivity analyses were performed wjth acceptable results assuming. -
m permanent loss of all feedwater (with operator initiatioii of the high pressure in-Jection system at 20 minutes) and loss of feedwater for only the first 20 minutes of the accident.
These results are apprcpriate for ANO-1 considering the ability to manually start the EFW pumps within 20 minutes as discussed under item (a) and (b) of this evaluation, assuming failure of automatic EFW actuation.
Another aspect of the studies was the assessment of rec.ent design changes on the lift frequency of pressurizer safety and relief valves.
The design changes included change in the setooint of the pressurizer power-operated relief valve (PORV) from 2255 psi to 2450 psi, change in the high pressure reactor trip setpoint from 2355 psi to 2300 psi and the installation of anticipatory reactor trips on turbine trip and on loss of feedwater.
In the past, during turbine trip and loss of feedwater cransients the PORV was lifted. With the new design these transients do not result in lifting of this valve.
However, lifting of both PORV and safety valves might occur in case of rod withdrawal and inadvertant boroa dilution transients, using the normally conservative assumptions found in the Chapter 15 safety analysis.
The above design changes did not effect the lift frequency of the valves for these Chapter 15 safety analyses. 272 237
' Based on our review of the;small break analyses presented by B&W, the staff has deter-mined that a Toss of all main feedwater with (a) an isolated PORV, but safety valves
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' ope ~ning and closing as designed, or (b) a stuck open PORV consequentially does not result in core uncovery, provided either EFW.or 2 HPI pumps are initiated within 20 minutes.
Based on the acceptable consequences calculated for small break LOCAs and loss of all main feedwater events and the expected reliability of the EFW and high pressure injection systems, we conclude that the licensee has complied with the ana lysis portion of paragraph (1)(d) of the Order.
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To support longer term operation of the facility, requirements will be developed for additional and more detailed analyses of loss of feedwater and other anticipated transients.
Mare detailed analysis of small break LOCA events are also needed for this purpose.
Accordingly, the licensee will be required to provide the analyses discussed in Sections 8.4.1 and 8.4.2 of the recent NRC Staff Report of the Generic Assessment of Feedwater Transients in Pressurized Water Reactors Designed by the Babcock and Wilcox Company (NUREG 0560).
Further details on these analyses and their applicability to other PWRs and BWRs will be specified by the staff in the n>ar fu-ture.
In addition, to assist the staff in developing more detailed guidance on design requirements of relief and safety valve reliability during anticipated transients, as discussed in Section 8.4.6 of the hJREG report, the licensee will be required to provide analyses of the mechanical reliability of the pressurizer relief and' safety valves of the ANO-1 facility.
The B&W analyses show that some operator action, both immediate and followup, is required under certain circumstances for a small break accident.
Immed.iate operator action is defined as those actions committed to memory by the operators which are necessary to take as soon as the problem is diagnosed.
To perform followup actions, operators must consult and follow instructions in written and approved procedures.
These procedures must always be readily available in the control room for the opera-tors use.
Guidelines were developed by B&W to assist the operating B&W facilities to develop emergency procedures for the small break accident.
The Operating Guidelines for Small Breaks were issuelf by B&W on May 5,1979.and rev [
iewed by the NRC staff.
Revisions recommended by the staff were incorporated in the guidelines.
In response to these guidelines, the licensee made substantial revisions
.to EP 1202.06 (Loss of Reactor Coolant /RC Pressure), EP1202.14 (Loss of Reactor Coolant Flow-RCP Trip), EP 1202.26 (Loss of Steam Generator Feed), EP 1202.23 (Steam Generator Tube Rupture), and EP 1202.05 (Degraded Power).
These emergency procedures define the required operator action in response to a spectrum of break sizes for a loss-of-coolant accident in conjunction with various equipment availability and f.ilures.
The procedure dealing with loss-of-reactor coolant (EP 1202.06) is divided into three sections.
The first ?-
- a rupture well in excess of the capability of the high pressure injection pt
.ge break in which the system depressurizes to the point of low pressure injec a).
An automatic reactor trip is assumed.
The second section of this procedure ast _mes the small break is within the capacity of the high pressure injection system and the reactor may not automatically trip.
The third section assumes reactor coolant system leakage within the capacity of a single makeup pump and no automatic reactor trip.
A separate procedure (EP 1202.23) provides guidance to the 272 238
9 ic
- ac in the. event of a' steam generator tube rupture.
In all cases ~ dealing-wi.th a
-' '
- T A, the operator actions are aimed at achieving a safe cold shutdown in' n with the normal cooldown procedure.
A; r.AaWi above, other procedures provide guidance to the operators for dealing wit... s.ndl breaks in the event of & degraded condition (such as a loss-of-feedwater and/.c inss of reactor coolant pumps).
These procedures are EP 1202.05, EP 1202.14, arnt 7. Pit M'. 26.
If all feedwater is lost, a heat removal path is established from the high ne Meure injection system through the braak and the pressurizer power-operated re h f valve or the safety valves.
Once feedwater is reestablished, the steam genera-cors an be used as a heat sink.
If the reactor coolant pumps are not available, the is dir.cted to establish and verify natural circulation.
Additional guidance
.,r is provided if natural circulation is not immediately achieved.
If normal power to tae m;ter-oriven emergency feed pump is lost, guidance is provided to the operator to pump from the diesel generator.
- p... c r u..>
For all cases in which high pressure injection is manually or automatically initiated, the operators are specifically instructed to maintain maximum HPI flow unless two criteria are met.
These criteria are:
' DI ' as been operating for <..reater than 20 minutes with flow ratas in excess
- 2M O gpm per train, or 3 eater than 3100 gpm with one train operating, o/'
Z.
Ali bot. and cold leg temper 3'.ures are at least 50 degrees below the saturation temperature for the existing RCS pressure.
If the 50 degrees subcooling canrot
-..ii.dained,after HPI cutoff, the HPI shall be reactuated.
Tre meu'rement to determine and maintain 50 F subcooling has been incorporated in all other orx edures in which HPI has been manually or sui.omatically initiated.
These
- s include, Steam Supply System Rupture, St.eam Generator Tube Rupture, Loss of nr -
.hartor Coolant Flow and Loss of Steam Generator Feedwater.
Each of these procedures, in addition to the loss of Reactor Coolant procedure, provide additional instructions _
rators in the event of faulty or misleading indications.
A subsequent
~
x.io.' scatement directs the operators to check alternate instrumentation channels.to_.
coafira the key parameter readings.
The ANO-1 staff have made revisions to all of
?.h p <; q <v'Argency procedures to include this T quirement.
Also, the licensee has pro-Su;g
- s. m <oc computer readout of 16 thermocouple indications of core exit temperatures i
s available to the operator in the control room.
The licensee further committed to
.-[y f r.stalisticn of an additional 16 thermocouples to be available before October 31, 1979.
The staff has reviewed the additional information to be gained with regard to n ~ vidinc additional verification of reactor coolant system temperature and finds the, modifications acceptable.
I I
N L 4: of Reactor Coolant procedure was reviewed by the NRC staff to determine its confercsoce with the B&W guidelines.
Comments generated as a result of this review m ircorporated in a further revision to the procedure.
A member of the NRC staff 272 239
nei th oogh this. emerge ~nty procedure in the ANO-1 control room.
The procecui
~
T @ d to Provide adequate guidance to the operators to cope with a small break loss A at accident.
The instrumentation necessary to diagnose the break, the indica-u ns ed controls required by the action statements, and the administrative controls 9 hich prevent unacceptat,le limits from being exceeded are readily available to the t p e ratort.. We conclude that the operators should be able to use this procedure to tring the olant to a safe shutdown condition in the event of a small break accident.
o, anit of nine af the 27 licensed operators and senior operators was conducted by ire N. sLaff to determine the operators' understanding of the small break accident,
. H %w they are required to diagnose and respond to it.
The ANO-1 staff has
~
conducted special training sessions for the operators on the concept of and use of snugacy procedure 1202.06.
The operators were found to h&ve sufficient knowledge of m ;' becak phenomeno
_d the general requirements of the emergency procedure.
.ach siceased individual i
Ialso receive additional training on the approved pro-cedure prior to power oper ion.
g
Tiie auriit of the operators also included questioning about the TN-2 incident and the resultire design changes made at ANO-1.
The fiscussions covered the initiating events
'f tu incident, the response of the plant to the simultaneous loss of feedwater and
' ' '.. A LOCA (PORV stuck open), and the operational actions that were taken during tne course of the incident. We found their level of understanding sufficient to be e t,e to rcscond to a similar situation if it happened at ANO-1.
We also concluded that they have adequate knowledge of subcooling and saturated conditions and are able r c:; he each condition in the primary coolant system by various methods.
The cac: pcy fecdwater system was also discussed during the audit to determine the opera-tors' ability to aistre proper starting and operation of the system during normal corditiors, as well as during adverse conditions such as loss of offsite power or loss or *rt.al feedwater.
The long term operation of the system was examined to evaluate the operators' ability to use available manual controls and water supplies.
The level of undentanding was found to be sufficient to assure proper short and long term
- v y fecdwater flow to the steam generators.
1N licensed operators and senior operators have received training concerning the, -
TMT-? accident, small break LOCA recognition, design modifications, and procedure
..c.ges.
Tc, determine the effectiveness of this training program a written exam was administared to all licensed personnel by the licensee.
Individuals scoring less than
" parcent on the exam will receive additional training and will not assume licensed cuties until a score of at least 85 percent is attained on an equivalent, but dif-fe"ent exam.
A' kansas Power and Light also contracted with B&W and NUS Corporation to concact audits to determine the effectiveness of the training program.
The NRC staff also conducteo audits which were judged satisfactory with some deficiencies noted to t w 'C -1 staff.
The ANO-1 staff will use the results of these audits and any gene:*:
weaknesses discovered on the written exams in their development of future training W roTulification programs.
The NRC staff will review all results and records as pa of the normal inspection function of the ANO-1 requalification program.
We concl e that ther e is adequate assurance that the operators at ANO-1 have and will continue to receive a sufficient level of training concerning the TMI-2 accident. 272 240
~ Based on the for.egoing evaluation, we conclude that the. licensee has compljed wit,h the
,_ requirements of item (d) of Paragraph (1) of the order.
Item e The order required that; "At least one Licensed Operator who has had THI-2 training on the B&W simulator will be assigned to the control room (one each shift)."
The licensee has confirmed that all reactor operators and senior operators have com-pleted the TMI-2 simulator training at B&W as required by the Order.
This training consisted of a class discussion of the TMI-2 event and a demonstration of the event on the simulator as it occurred and how it should have been controlled.
The class dis-cussion was about one hour long and the remainder of the four hour session was conducted on the simulator.
The TMI-2 event, including operational errors, was demonstrated to each operator.
fhe event was again initiated and the operators were given " hands-on" experience in successfully regaining control of the plant by several methods. Other trar,sients which resulted in depressurization and saturation condi-tions were presented to the operators in which they maneuvered the plant to a stable, subcooled condition.
CONCLUSION We conclude that the actions described above fulfill the requiret.ents of our Order of May 17, 1979 in regard to Paragraph (1) of Section IV.
The licensee having met the requirements of Paragraph (1) may restart ANO-1 as provided by Paragraph 2.
Paragraph 3 of Section IV of the Order remains in force until the long term modifica-tions set forth in Section II of the Order are completed and approved by the NRC.
3-272 241
v, e
REFERENCES
-^ '
x 1.
Letter from J. H. Taylce (B&W) to R. J. Mattson (NRC) transmitting report entitled,
" Evaluation of Transient dehavior and Small Reactor Coolant System Breats in the 177 Fuel Assembly Plant," dated May 7, 1979.
2.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting revised
'~
Appendix 1, " Natural Ciculation in B&W Operating Plants (Revision 1)," dated May 8, 1979.
3.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting additional information regarding Appendix 2, " Steam Generator Tube Thermal Stress Evalua-tion," to report identified in Item 2 above, dated May 10, 1979.
4.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC), providing an analysis for "Small Break in the Pressurizer (PORV) with no Auxiliary Feedwater and Single Failure of the ECCS," identified as Supplements 1 and 2 to Section 6.0 of report in Item 2, dated May 12, 1979.
5.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC), providing an anal.ysis for "Small Creak in the Pressurizer (PROV) with no Auxiliary Feedwater and Single Failure of the ECCS" identified as Supplements 1 and 2 to Section 6.0 of report in Item 2, dated May 12, 1979.
6.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC), providing Supplement 3 to Section 6 of report in Item 2, dated May 24, 1979.
~
272 242
_ 1, _
~
-UNrTED STATES NUCLEAR -REGULATOR COMMISSION'
r~3 r
' - ~ - ' '&
ARKANSAS POWER & LIGHT COMPANY DOCKET NO.
50-313 NOTICE OF AUTHORIZATION TO RESUME OPERATION The United States Nuclear Regulatory Commission issued an Order on May 17, 1979 (44 FR 29997, May 23, 1979), to Arkansas Power & Light Company (the licensee), holder of Facility Operating Licensee No. DPR-51 for Arkansas Nuclear One, Unit No. 1 (ANO-1), confirming that the licensee e:complished &
seried of actions, both immediate and long term, to increase the capability and reliability of ANO-1 to respond to various transient events.
In addition, the Order confirmed that the licensee would maintain ANO-1 in a shutdown condition until the following actions had been satisfactorily completed:
(a) Upgrade of the timeliness and reliability of the Emergency Feedwater (EFW) system by performing the items specified in Enclosure 1 of the licensee's letter of May 11, 1970.
Provide changes in design for NRC review.
(b) Develop and implement operating procedures for initiating and centrol-ling 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 Three Mile Island Unit No. 2 training on the Babcock & Wilcox similar to the control room (one each shift).
~ ~ ~
By sumittal of May 17, 1979, as supplemented by letters dated May 21, and 22, 1979 the licensee has documented the actions taken in response to the May 17, Order.
Notice is herebygiven that the Director of Nuclear Reactor Regulation (the Director) has reviewed this submittal and has concluded that the licensee has satisfactorily completed the actions prescribed in items (a) through (e) of paragraph (1) of Section IV of the Order, that the specified modifications and analyses are acceptable and the specified implementing procedures are appropriate.
Accordingly, by letter dated the Director has authorized the licensee to resume operation of ANO-1.
The bases for the Director's conclusions are more fully set forth in a Safety Evaluation dated 272 243 m.-m.==u m.=mau===-
a===-.
4 Copies of (1) the licensee's letters dated May 17, 21 and 22, 1979, (2) the Director's letter dated
, and (3) the Safety Evaluation dated are available for inspection at the Cc:.: mission's Public Document -Roo:: at 1717 H Street, N.W., Washington, D.C.
%555, and are being placed in the Commission's local public document room at the Arkansas Polytechnic College, Russellville, Arkansas.
A copy of items (2) end (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory
'x:aission, Washington, D.C.
20555, Attention:
Director, Division of Operating
<. i : '. o rs.
FOR THE NUCLEAR REGULATORY COMMISSION Rcbert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors Dated at Bethes d, Maryland t.iiis _ day of May 1979.
272 244