ML19224C948
| ML19224C948 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear, Crane  |
| Issue date: | 05/30/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19224C937 | List: |
| References | |
| NUDOCS 7907100292 | |
| Download: ML19224C948 (14) | |
Text
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EVALUATION OF LICENSEE'S COMPLIANCE WITH THE NRC ORDER DATED MAY 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 Nuciear 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 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 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 agreed in a letter cated 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) of 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 Ah0-1 site.
An audit of the AND-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 licensee take the following action; 790710029L 2,59 154
" 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 indvpendent of offsite power, and one motor-driven EFW pump that must be manually transferred to a cital 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 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 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 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 driven EFW pump is permanently connected to vital power.
Since the time frame of five minutes is well within the 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 supr N.
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 operator at the necessary valves in communication with the control room during the surveillance mode 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 259 JE
Emergency Feedwater 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 determined that training of 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 fol bwing 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 precedures 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 tanner.
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 to specified values.
In addition, the NRC staff required that a test be conducted to demonstrate the canability to provide and control emergency flow to the steam generators.
The licensee has committed to perform a test at low pcwer operation (10-15%) during power 259 156
~..
The primary objective of the test will be to further verify the capability
" m on.
r ally control steam generator level independent of ICS.
A member of the NRC
' t tne ANO-1 site will witness the test and will verify acceptance prior to r.. d ng to full power operation.
Subject to the successful completion of this test, rw -onclude that the licensee has complied with this portion of the order.
h m clso ordered that; 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-t W of oparation.
Consistent with the cover letter for this evaluation, the NRC
...'11 receive from the licensee within seven days revised proposed Technical Spe:uichtions with regard to design and procedural changes.
It was also ordered that the licensee; "5
Review and revise, as necessary, the procedures and conduct training for providing altern2te sources of water to the suction of the EFW pumps."
s he riehns available to alert the operator to perform the manual transfer of 'EFW from
- t. e ca..dar.; ate storage tank (CST) to the service water system consists of an alarm in the control room which annunciates on low EFW pump suction pressure.
The licensee has
- M ion;l annunciation in the control room on low level in the condensate storage tm k This new feature allows direct control room annunciation that is radundant to the arittint low suction pressure switch annunciation.
The NRC staff reviewed procedure GP '.106.06 " Emergency FW Pump Operation" and requested revision of the guidance to the n -,r
- for providing alternate sources of water to the suction of the EFW pumps.
Tne -evision has been made to provide additional guidance to the operator for alternate meenr. of verifying low level in the condensate storage tank.
The NRC staff at the
c verified that the control room operators are properly trained to carry out t :s precedures.
We conclude that the licensee has complied with the requirements to rtview and revise procedures and has conducted operations personnel training for nrrvir:ing alternate sources of water.
I' was ah;o ordered that; "i
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 pump."
N li"?nsee has installed an automatic start of the motor-operated EFW pump on loss of eil RC pumps or loss of both main feedwater pumps.
Relay contacts associated with c"isting relays within the integrated control system cabinet, additional relays and t.c.it3 cts, and wiring are arranged in the final actuation control circuitry for the c riter-driven emergency feedwater pump such that, if offsite pcwer is available, the noter is provided a signal to start automatically.
Further, nanual capability to 9 c') o
) Q ^l L i
s a*
~
' Miate and/or override this automatic circuitry is included in the design.
In ado, ucc. annunciation within the control room has been provided whenever this pump is
.O L2a by the automatic circuitry.
Based on our review of this aspect of the design, we conclude that it is in accordance C th the order.
it was also ordered that; Procedures will be developed and implemented and training conducted to provice guidance for timely operator verification of any automatic init:a-tion of EFW."
'% 1%ee has revised procedure OP 1106.06 (Emergency Feedwater Pump Operation) to "rvide specific operator guidance as to the methods for confirming automatic initia-tien of EFW.
This includes:
Verification that pump discharge pressure is greater than OTSC pressure.
pri ication of feedwater flow (on the flow indicator installed pursuant to f
cr. 9, below).
Eervation of steam generator levels.
% rgr w crocedures for plant transients requiring initiation of emergency feedwater (sc-b is loss of normal feedwater or loss of reactor coolant flow) require the operator t.o verify the initiation of emergency feedwater.
Additionally, the operator is required to observe alternate instrumentation channels to provide further assurance. The NRC nat f hes confirmed that control room operators are properly trained to carry out triese procedures.
"e a'so ordered that; "8.
Verification that Technical Specification requirements for EFW capacity are 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 w ic'ent analyses.
Low power testing will substantiate the availability of at least this flow capacity by each EFW train (see Part 3).
Consistent with the cover letter w this evaluation, we will require submittal of a Technical Specification change conc.uning EFW capacity.
This change will be a limiting condition of reactor opera-tien in the event the minimum allowable value assumed in the accident analysis is not ret. cnd will provide for periodic surveillance.
It was also ordered tha';
"9.
Modifications will be made to rrovide verification in the control room of EFW flow to each steam generator." qq
) q j.
J /
l ss 6
To ve.-ify that emergency feedwater is being pumped to the steam generators, the licensee is providing two orifice plates and differenti# pressure sensing equipment.
These ficw devices will be installed on each of the EFW injection flow paths downstream of the crossover line, so that flow to each steam 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 '.o assure performance of this design codifi-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 acceptable.
It was also ordered that the licenser; "10. Provide a means of notification to the control room that the EFW system has auto started.
This notification can be provide'd from a temporary modifica-tion or a dedicated operator."
As described in Part 7, above, the control roc.n operator can determine the initiaticn of emergency feed by observation of pump discharge pressure (as compared to steam generator pressure), emergency feed flow, and steam generator level.
In addition, caaunciation 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; "Cevelop and implement operating procedures for initiating and controlling EPr/
independent of Integrated Control System (ICS) control."
Several components in each EFW train are provided with an automatic initiation signal.
Four componants in one train are one steam-driven pump controller, 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 manually control these com-conents from the control room.
This manual provision provides overriding control of the autraatic 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 d rcuitry.
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 2b9 /(.
initiation failed 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 instrument: and whether the operator was given spacific 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 staf f 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 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 reactor trip (rather than delaying the trip until an operator took action or until a primary system parameter exceeded its trip setpoint),
the licenset 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) dated 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 circuitry 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-Hydraul.ic master trip circuitry.
The power for this circuitry is provided from a Class lE 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 Integrcted 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 29
!Ob
'y energized 24 volt direct current circuit containing two additional relays.
xi r.w i
% volt power supply is derived within the Integrated Control System cabinet.
To
- r. cf the breakers and trip the reactor, two associated direct current relays "o
our contact closures to energized two direct current shunt coils (two contact
'. r t.lcsures car shunt trip coil and one shunt trip coil for each of the two reactor trip r.lternating current circuit breakers).
Power is provided to the shunt trip coils from Clan !E 125 volt direct current buses.
"eodwater pump trip is sensed by two normally de-energized auxiliary r21ays W
i assaciat >d with the main fe.edwater pumps master trip circuitry (one relay associated W 'h eec" of the two main feedwater pumps).
The remaining circuitry associated with z.s R91s ideiltical to that described above for the turbine trip including power supcli23 with the exception that two corresponding relays and contacts are provided.
f.lio. the two associated contacts (these contacts are arranged in parallel) within the h m't direct current circuit are in series with the associated turbine trip contact.
Provisions have been included to automatically bypass and re-instate these additional trips at low power to allow a normal startup and shutdown.
Operator verification of the bypass removal is required procedurally during power escalation.
The NRC staff at
'N
'.t.'~.-l site audited a sample of ANO-1 operators and concluded that they were familar it;i Je functions of these trips and associated procedural requirements.
T 'e licensee has analyzed this additional circuitry with respect to its independence fic*r. the existing reactor trip system.
The licensee has stated that the shunt coil is cart of the existing AC reactor trip breaker.
However, it is separate and operates
- 4. N o. ently from the 120 volt alternating current undervoltage trip coil of the associatea breaker.
The reactor trip safety grade signal de-energizes the 120 volt Ctm 1:tirg :urrent undervoltage coil to produce a trip of the associated alternating
- tr cn* Sreaker.
based or. cur review of the implementation of the trip circuitry with respect to its Mc;er andence from the existing trip circuitry, we conclude that this addition will not
.50 existing reactor protection system design.
The licensee has installed and cunp ated checkout of the trip circuitry.
~5 1 h risee has committed to perform a monthly periodic test on the added circuitry to anoistrate its ability to open the AC circuit breakers (tripping the AC breakers via tt;a 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 hat 3Ntoown condition as the result of a normal outage or reactor trip (but not more T y.r.j 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 aoove, we conclude that this additional circuitry is in accordance with the
_&,: wuats of item (c) of the order.
. 259 161
Item d This item in the order requires the licensee to:
" Complete analyses for potential small breaks and implement operating instruc-tions to define operator action."
By letter from William Cavanaugh III (AP&L) to H. Denton (NRC) dated May 11, 1979, the licensee committed to 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 generic study, with the exception of emergency feedwater flow, conservatively bound the ANO-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 Acci' dent (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 generators, high pressure infection 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 hcci. 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 appropriate 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 design 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 v~ lve (PORV) from 2255 psi to a
2450 psi, change in the high pressure reactor trip setpoint from 2355 psi to 2300 psi and the installation of anticipat.ory reactor trips on turbine trip and on loss of feedwater.
In the past, during turbine trip and loss of feedwa er transients the PORV was lifted.
With the new design these transients do not result in lifting of this valve.
Haever, lifting of both PORV snd safety valves might occur in case of rod withdrawal and inadvertant baron dilution transients, using the normally conservative assumptions found in the Chapter 15 safety analysis.
The above design changes did not effect the lift fr.auency of the valves for these Chapter 15 safety analyses. i <'
9qq L
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 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.
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 recmt NRC Staff Repurt 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 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 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 reouired under certain circumstances for a small break accident.
Immediate operator action is defined as those actions cocmitted 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 tne opera-tors usa.
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 1202.23 (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 failures.
The procedure dealing with loss-of-reactor coolant (EP 1202.06) is divided into threc 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 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 autcmatic reactor trip.
A separate procedure (EP 1202.23) provides guidance to the 259 10s 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 fro::: 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 reacto= coolant pumps are not available, the operator is directed to establish and verify natural circulation. Additicnal guidance is provided if natural circulation is not immediately achieved.
If normal power to the aotor-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 or 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 1F 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 providing 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.
Comments generated as a result of this review were incorporated in a further revision to the procedure.
A cember of the NRC staff 2
,,.;d on the foregoing evaluation,13 - aclude that the licensee has complied with the "eun rencats of item (d) of Paragraph 0) of the order.
yae The order required that; "At least one Licensed Operator who has had THI-2 training on the B&W simulator
','ll De assigned to the control room (one each shift)."
-rwe has confirmed that all reactor operators and senior operators have com-This training pleted the TMI-2 simulator training at B&W as required by the Order.
- onsisted of a class discussion cf the TMI-2 event and a demonstration of the event on The class dis-c..alstor as it occurred and how it should have been controlled.
- Lsio:i was about one hour long and the remainder of the four hour session was a
conduct.ed on the simulator.
The THI-2 event, including operational errors, was The event was again initiated and the operators were demonstrated to each operator.
gioen " hands-on" experience in successfully regaining control of the plant 'ay several Other transients which resulted in depressurization and sat" aticn condi-utbods.
- 'c v 'ere presented to the operators in which they maneuvered the p' y t to a stable,
'.c b o condition.
CLM.LL':ICM Us co c".-@ that the actions described above fulfill the requirements of sur Order of May 17, 1079 in regard to Paragraph (1) of Section IV. The licensee having met the re;cirements of Paragraph (1) may restart ANO-1 as provided by Paragraph 2.
ParagrW. 3 of Section IV of the Order remains in force until the long term modifica-set torth in Section II of the Order are completed and approved by the NRC.
t '. 259 165
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 conclude 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 operators and ser.ior operators was ccnducted 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.
TN ANO-1 staff has conducted special training sessions for the operators on the concept of and use of emergency procedure 1202.06.
The operator. were found to 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 trade at ANO-1.
The discussions covered the initiating events of the incident, the response of the plant to the simultaneous loss of feedsater ana small break LOCA (PORV stuck cpen), and the operational actions that were taken during 1.he course of the incident. We found their level of understanding sufficient to be a:e to respond to a similar situation if it happened at ANO-1.
We also cor.cluded that they have adequate knowledge of subcooling and saturated conditions and are able to recognize each condition ir. the primary coolant system by various methods.
The emergency feedwater system was also discussed during the audit to determine the opera-tcrs' sbility to assure proper starting and operation of the system during normal conditions, as well as during adverse conditions such as loss of offsite power or loss of ncemal feedwater.
The long term operation of the system was examined to evaluate the opt ators' 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, small break LOCA recognition, design modifications, and procedure changes.
To determine the effectiveness of this training program a written exam wus administered to all licensed personnel by the licensee.
Individuals scoring less than 85 percent on the exam will receive additional training and will not assume licensed duties until a score of at least 85 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 ANO-1 staff.
The ANO-1 staff will use the results of these audits and any generic weaki. esses 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 requalif: cation program.
We conclude that there is adequate assurance that the operators at ANO-1 have and will continue to receive a suf ficient level of training concerning the TMI-2 accident. 259 166
Based on the foregoing evaluation, we conc.lude 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 TM1-2 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.
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 resrltc" ':. depressurization and saturation condi-tions were presented to the operators in which they manauvered the plant to a stable, subcooled condition.
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 inng term modifica-tions set forth in Section II of the Order are completed and approved by the NRC. 259 1 9