ML19225A376
| ML19225A376 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19225A371 | List: |
| References | |
| NUDOCS 7907190105 | |
| Download: ML19225A376 (14) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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. E WASHINGTON. D. C. 20555 g' ' v
+..s' EVALUATION OF LICENSEE'S CCMPLIANCE WITH THE NRC ORDER DATED MAY 17, 1979 ARKANSAS POWER & LIGHT COMPANf ARKANSAS NUCLEAR ONE, UNIT l' 00CKET 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 iso. 2 accident, the NRC staff initially identified several human errors that 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 safety concerns that warranted additional short-term design and procedural changes at cperating 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 agreed 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 indefi' lite basis until actions described in paragraphs (a) through (e) of paragraph (1) of Sec+. ion 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 di',cussions 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 i It was oMered that the licensee take the following action; 7907190I05 34/-
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" 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."
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 transferred 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; "1.
Review procede es, 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.
Durir.g 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 the 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 reanual 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 accnrdingly. Members of the NRC staff on site have verified that the control room operators are properly trained to carry out this revised procedure.
The 1:censee has also agreed to have two operators stationed in the control room at all times until the electric driven EFW pump is permanently connected to vital power.
Since the time frame of five minutes is well within the allowable delay of 20 rinutes indicated by the generic S&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 surveillanct test mode, procedures will be implemented and training conducted to provide an operator at the necassary valves in communication with the control room during the surveillance mcde to carry out the valve alignment changes upon EFW cemand events."
The ANO-1 staff has revised 0F 1106.06 " Emergency Feedwater Pump Operation."
Supplements I and II provide procedures for conducting the Electric and Steam Driven 3 /j [
2'9
Emergency Feedsater Pump surveillance test, respectively.
The NRC staff has reviewed these procedures which require in part; " Operator shall remain in area for duration of test in communication with the control room to align system in the event of an EFW demand."
The NRC staff has also determiaed that training cf operators in use of this procedure has been conducted and is adequate.
Subject to confirmation by a member of the NRC staff that noise levels in this area during plant operation are conducive to communications with the control room, we conclude that the licensee has complied with the order.
It was also ordered that the licensee; "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 pracedures were reviewed by the NRC staff to assure that feedwater from both the motor-driven pump and the steam-driven pump would 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 opening the steam supply valves.
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 tc 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 pcwer b4[
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ascension.
The primary objective of the test will be to further verify the capability to manually control steam generator level independent of ICS.
A member of the NRC staff at the ANO-1 site will witness the test and will verify acceptanca prior to proceding to full power operation.
Subject to the successful completion of this test, we conclude that the licensee has complied with this portion of the order.
It was also ordered that; "4.
The EFW pumps will be verified operable in accordance with the ANO-1 Technical Specifications and Surveillance Procedures."
The ANO-1 Technical Specifications provide for EFW surveillance and limiting condi-tions of operation.
Consistent with the covar letter for this evaluation, the NRC staff will receive frcm the licensee within seven days revised proposed Techt ical Specifications with regard to design and procedural changes.
It was also ordered that the licensee; "S.
Review and revise, as necessary, the procedures and conduct training for providing alternate sources of water to the suction of the EFW pumps."
The means avnilable to ciert the operator to perform the manual transfer of EFW from the condensate storage tank (CST) to the service water system consists of an alarm in the control room which annunciates on Icw EFW pump suction pressure.
The licensee has an additional annunciation in the control room on low level in the condensate storage tank.
This new feature allows direct ccatrol room annuriciation that is redundant to the exis; ng low suction oressure switch annunciation. The NRC staff reviewed procedure OP 1106.01 " Emergency FW Pump Operation" and requested revision of the guidance to the operator for providing alts,rnate sources of water to the suction of the EFW pumps.
The revision has been made to provide additional guidance to the operator for alternate means of verifying low level in the condensate storage tank.
The NRC staff at the site has verified that the control room operators are properly trained to carry out these procedures. We conclude that the licensee has complied with the requirements to review and revise procedures and has conducted operations personnel training for providing alternate sources of water.
It was also ordered that; "6.
In the event emergency feedwater is necessary and offsite power is available, an auto start signal will be provided to the motor driven emergency feedwater p ump.
The 1 f censee has installed an automatic start of the motor-operated EFW pumo on loss of ali RC pumps or loss of both main feedwater pumps.
Relay contacts associated with existing relays within the integrated control system cabinet, additional relays and contacts, and wiring are arranged in the final actuation control circuitry for the motor-criven emergency feedwater pump such that, if offsite power is available, the motor is provided a signal to start automatically.
Further, manual capability to
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e initiate and/or override this automatic circuitry is included in the design.
In addition, annunciation wit':n the control room has been provided whenever this pump is started 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 t1mely operator verification of any automatic initia-tion of EFW."
The licensee has revised procedure OP 1106.06 (Emergency Feedwater Pump Operation) to prnvide scecific operator guidance as to the methods for confirming automatic initia-tion of EFW.
This includes:
Verif. ation 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 feedwater (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 instrementation channels to provide further assurance.
The NRC staff has confirmed that control room operators are properly trained to carry out these procedures.
It was also ordpred that; "8.
Verification that Technical Specification requirements 'or EFW capacity are in acccrdance with the accident analysis will be conduc.ed."
The licensee has stated that a minimtm flow of 550 gpm is required to support the accident analyses.
Low power testing will substantiate the ava'! ability af at least this flow capacicy by each EPl train (see Part 3).
Consistent with the cover letter to this evaluation, we will require submittal of a Technical Specification change concerning EFW capacity.
Th.s change will be a limiting condition of reactor opera-tion in the event the minimtm allowable value assumed in the accident analysis is not met, and will provide for p sciodic 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.".i n.
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To verify that emerge..cy feedwater is being pumped to the steam generators, the licensee is providing two orifice plates and differential pressure sensing equipment.
These flow devices will be installed on each of the EPd injection flow paths downstream of the crossover line, so that ficw to each steam generator sill be measured.
The output of the differential pressure transmitter will be displayed in the control room, indicated in gallons per-mir.ute.
A verification test will be performed to anur e performance of this desicr. modifi-cation.
This will be performed as part of tne test described in Part (3) in this report.
The test procedure has been reviewed by the NRC staff and verified as acceptable.
It was also ordered that the licensee; "10. Provide a means of notification to the control room that the EFW system has auto started.
This notification can be provided from 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 ccmpared 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 accordance with the order.
Item b It was ordered that the licensee;
" Develop and implement operating procedures for initiating and controlling EFW independent of Integrated Control System (ICS) control."
Several components in each EFW train are provided with an automatic initiation signal.
Four components in one train are one ste Hriven pump controller, one motor-eperated valve located at the discharge of this pt.
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 manually control these com-ponents frcm 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 K/
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initiation f ailed and to control the steam generator level to specified values.
The review of the 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 by 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 these determinations were presented to the 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 emergercy 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 th? revised procedures and could implement them correctly.
Based on the NRC staff audit, we conclude that the revised procedures and operator training are satisfactory.
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 from a malfunction in the secondary system (loss of main feedwater and/or turbine trip).
To obtain an earlier "eactor trip (rather than delaying the trip until an operator took action or until a primary system parameter exceeded its trip setpoint),
the licensee committed tu install a hardwired control grade reactor trip on the loss cf all main feedwater and/or on turbine trip (letter from William Cavanaugh III (AP&L) to H. Denton (NRC) dated May 11, 1979).
The purpose of this anticipatory trip is to minimize the t
_1 for opening of the power-operated relief valve (PORV) and/or the safety valves e the pressurizer.
The licensee has indicated that this new circuitry meets this objective by providing a reactor trip during the incipient stage of the related transients (turbine trip anu/cr loss of Main feecwater).
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 3 normally de-energized auxiliary relay associated with the main turbine Electro-Hydraulic master trip circuitry.
The power for this circuitrv i 'irovided from a Class 1E 125 volt dirai. 'irrent bus by way of a 125 volt dist
' panel. A contact from this auxiliary relay is arranged into a 118 volt alternaw
.Jrrent Circuit Containing a normally de-energized relay.
This alternating current relay is physically located within the Integrated Cor.crol System cabinet and is proviced power from the associated Integrated Control Sys+em power supply.
A contact from +,his alternating current relay is arrar.ged into a if 7 ogj
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normally energized 24 volt direct current circuit containing two additional relays.
This 24 volt power supply is derived within the Integrated Control System cabinet.
To open each of the breakers and trip the reactor, two associated direct current relays provide four contact closures to energize two direct current shunt coils (two contact closures per shunt trip coil and one shunt trip coil for each of the two reactor trip alternating cn**ent. circuit breakers).
Power is provided to the shunt trip coils frcm Class 15 125 s t direct current buses.
The main feedwater pump trip is sensed by two normally de-energized auxiliary relays associated with the main feedwater pumps master trip circuitry (one relay associated with each of the two main feedwater pumps).
The remaining circuitry associated with this trip is identical to that described above for the turbine trip including power supplies, with the exception that two corresponding relays and contacts are provided.
Also, the two associated contacts (these contacts are arranged in parallel) within the 24 volt direct currant circuit are in series with the associated turbine trip contact.
Provisions have been included to automatically bypass and re-instate these additional trips at icw power to allow a normal startup and shutdown.
Operator verification of the bypass removal is required procedurally during power escalation.
The NRC staff at the ANO-l site audited a sample of ANO-l operators and cancluded that they were familar with the functions of these trips and associated procedural requirements.
Tha licensee has analyzed thir additional circuitry with respect to its independence from the existing reactor trip system.
The licensee has stated that the shunt coil is part of the existing AC reactcr trip breaker. However, it is separate and operates independently from the 120 volt alternating current undervoltage trip coil of the associated breaker.
The reactor trip safety grade signal de-energizes the 120 volt alternating current undervoltage coil to produce a trip of the associated alternating current breaker.
Based on our review of the implementation of the trip circuitry with respect to its independence from the existing trip circuitry, we conclude that this addition will not degrade the existing reactor protection system design. The licensee has installed and completed checkout of the trip circuitry.
The licensee has committed to perform a monthly periodic test on the added circuitry to demonstrate its ability to open the AC circuit breakers (tripping the AC breakers via the shunt trip circuit). Additionally, the licensee has committed to perform a more complete test of this additional circuitry whenever the reactor is brought to a hot shutdown condition 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 the above, we conclude that this additional circuitry is in accordance with the requirements of item (c) of the order. 34[
796 n,
Item d This item in tne order requires the licensee to:
" Complete analyses for potential small breaks and implement operating instruc-tions to define operator action."
By letter from William Casanaugh III (AP&L) o H. Denton (NRC) dated May 11, 1979, the licensee committed to providing the scalyses and operating procedures of this requirement.
Babcock and Wilcox, the reactor vendoi for the ANO-1 plant, submitted an analysis entitled, " Evaluation of Transient
.havior and Smai, Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" and sylements to these analyses (References 1 through 6).
The major parameters used in this ge,e M study, wi'h the exception of emergency feedwater flow, conserva'ively bound the d0 1 p ant an additional analysis assuming a bounding value for emergency f + 4ter 'lw was sOsequently submitted (Reference 6).
In a letter dated May 16, ac
, ' l, aP&L has referenced f e analyses as appropriate for ANO-1.
The sta+1 evalu/ Ma of tbc C&W generic st_dy has been com-pleted and the results of the e @ at w wil' be issued as a NUREG report in June 1979.
A principal finding of our ge.ieri.
1ew is a reconfirmation that Loss-of-Coolant Accident (LOCA) analyses of breaks et the lower end of the small break spectrum (smaller than 0.04 sq. ft.) demorotrate that a combination of heat removal by the steam generators, 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 core uncovery is not predicted for these events.
The calculated 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.
Sensitivity analyses were performed with acceptable results assuming permanent loss of all feedwater (with operator initiation 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 appropriat-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 recent de,ign changes on the lift frequency of pressurizer safety and relief valves.
The design changes included change in the setpoint 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 transients the PORV was lifted. With the new design these transients do not result in lifting of this valve. However, lifting of coth PORV and safety valves might occur in case of rod witharawal and inadvertant boron dilution transients, using the normally conservative assumotions 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. 34[
22b
Based on our review of the small break analyses presented by B&W, the staff has deter-mined that a loss of all main feedwater with (a) an isolated PORV, but safety valves opening and closing as designed, or (b) a stuck open PORV consequentially does not result in core uncovery, provided either EFW oe. 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.
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. More 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 spccified by the staff in the near 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 NUREG repc
, 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.
Immediate 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 issued 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.'3 (Steam Generator Tube Rupture), and EP 1202.05 (Degraded Power).
These emergency procedures cefine 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 failures.
The procedure dealing with loss-of-reactor coolant (EP 1202.06) is divided into three sections. The first deals with a rupture well in excess of the capability of the high pressure injection pumps (a large break in which the system depressurizes to the point of low pressure injection).
An automatic reactor trip is assumed.
The second section of this procedure assumes the small break is within the cacacity 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 makeuo pump and no automatic reactor trip.
A separate procedure (EP 1202.23) provides guidance to the /
operator in the event of a steam generator tube rupture.
In all cases dealing with a small break, the operator actions are aimed at achieving a safe cold shutdown in accordance with the normal cooldown procedure.
As indicated above, other procedures provide guidance to the operators for dealing with small breaks in the event of a degraded condition (such as a loss-of-feedwater and/or loss of reactor coolant pumps).
These procedures are EP 1202.05, EP 1202.14, and EP1202.26.
If all feedwater is lost, a heat removal path is established from the high pressure injection system through the break and the pressurizer power-operated relief valve or the safety valves. Once feedwater is reestablished, the steam genera-tors can be used as a heat sink.
If the reactor coolant pumps are not available, the operator is directed to establish and verify natural circulation.
Additional guidance is provided if natural circulation is not immediately achieved.
If normal power to tne motor-driven emergency feed pump is lost, guidance is provided to' the operator to power this pump from the diesel generator.
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:
1.
LPI has been operating for greater than 20 minutes with flow rates in excess of 2650 gpm per train, or greater than 3100 gpm with one train operating.
2.
All hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure.
If the 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactuated.
The requirement to determine and maintain 50 F subcooling has been incorporated in all other procedures in which HPI has been manually or automatically initiated.
These procedures include, Steam Supply System Rupture, Steam Generator Tube Rupture, Loss of Reactor Coolant Flow and Loss of Steam Generator Feedwater.
Each of these procedures, in addition to the Loss of Reactor Coolant procedure, provide additional instructions to the operators in the event of faulty oc misleading indications.
A subsequent action statement directs the operators to check alternate instrumentation channels to confirm the key parameter readings. The ANO-1 staff have made revisions to all of these emergency procedures to include this requirement. Also, the licensee has pro-vided for computer readout of 16 thermocouple indications of core exit temperatures available to the operator in the control room.
The licensee further committed to installation 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 previding additional verification of reactor coolant system temperature and finds the modifications acceptable.
The Loss of Reactor Coolant procedure was reviewed by the NRC staff to determine its conformance with the B&W guidelines. Ccmments generated as a result of this review were ir.corporatM in a further revision to the procecure.
A member of the NRC staff 347 m
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walked through this emergency procedure in the ANO-1 control room.
The procedure was judged to provide adequate guidance to the operators to cope with a small break loss of coolant accident. The instrumentation necessary to diagnose the break, the indica-tions and controls required by the action statements, and the administrative controls which prevent unacceptable limits from being exceeded are readily available to the operators.
We concl'ude that the operators should be able to use this procedure to bring the plant to a safe shutdown condition in the event of a small break accident.
An audit of nine of the 27 licensed oper; tors and senior operators was conducted by the NRC staff to determine the operators' understanding of the small break accident, including how 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 emergency procedure 1202.06.
The operators were found
+.o have sufficient knowledge of the small break phenomenon and the general requirements of the emergency procedure.
Each licensed individual will also receive additional training on the approved pro-cedure prior to power operation.
The audit of the operators also included questioning about the TMI-2 incident and the resulting design changes made at ANO-1.
The discussions covered the initiating events of the incident, the response of the plant to the simultaneous loss of feedwater and small break LOCA (PORV stuck open), and the operational actions that were taken during the course of the incident. We found their level of understanding sufficient to be able to respond to a similar situation if it happened at ANO-1.
We also coicluded that they have adequate knowledge of subcooling and saturated conditions and are able to recognize each condition in the primary coolant system by various methods.
The emergency feedwater system was also discussed during the audit to determine the opera-tors' ability to assure proper starting and operation of the system during normal conditions, as well es during adverse conditions such as loss of offsite power or loss of normal 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 understanding was found to be sufficient to assure proper short and long term emergency feedwater flow to the steam generators.
The licensed operators and senior operators have received training concerning the TMI-2 accident, smail break LOCA recognition, design modifications, and procedure changes.
To determine the effectiveness of this training program a written exam was administered to all licensed personnel by the licensee.
Individuals scoring less than 90 percent on the exam will receive additional training and will not assume licensed duties until a score of at least 91 percent is attained on an equivalent, but dif-ferent exam. Arkansas Power and Light also contracted with B&W and NUS Corporation to conduct audits to determine the effectiveness of the training program.
The NRC staff also conducted audits which were judged satisfactory with some deficiencies noted to the ANU-l staff. The ANO-1 staff will use the results of these audits and any generic weaknesses discovered on the written exams in their development of future training and requalification programs.
The NRC staff will review all results and records as part of the normal inspection function of the ANO-1 requalification program.
We conclude that there 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.
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Based on the foregoing evaluation, we conclude that the licensee has complied with 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 TMI-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 THI-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 rt..lainder of the four hour session was conducted on the simulator.
The TMI-2 event, including operational errors, was demonstrated to each operator.
The event was again initiated and the operators were given " hands-on" experience in successfully regaining control of the plant by several methods.
Other transients which resolted in depressurization and saturation condi-tions were presented to the operators in which they maneuvered the plant to a stable, subcooled cotidition.
CONCLUSION We conclude that the actions described above fulfill the requirements 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.
Dated: May 31,1979 347 250
REFERENCES 1.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting report entitled,
" Evaluation of Transient Behavior and Small Reactor Coolant System Breaks 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 Fiants (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 is 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 analysis for "Small Break in the Pressurizer (PROV) with no Auxiliary Feedwater and Single Failure of the ECCS" identified as Supplernents 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. n L, \\
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