ML19207B420
| ML19207B420 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/02/1979 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML19207B419 | List: |
| References | |
| NUDOCS 7908290139 | |
| Download: ML19207B420 (29) | |
Text
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EVALUATION OF LICENSEE'S C0FPLIANCE WITH THE NRC ORDER DATED PAY 16, 1979 Fl.0RIDA POWER CORPORA 1'M CRYSTAL RIVER UNIT N0. 3 NUCLEAR GENERATING STATION l
DOCKET NO. 50-302 1
\\
e Ju3y 2,1979
%0082001ff
INTRODUCTION By Order dated May 16,1979, (the Order) the Florida Power Corporation (licensee or FPC) was directed by the NRC to take certain actions with respect to Crystal River Unit No. 3 (CR-3). Prior to this Order, and as a result of a preliminary review of the Three Mile Island Unit No. 2 (TMI-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 instructed to take a nud:ber of irrnediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Comission's Office of Inspection and Enforcement (IE). Subsequently, an additional bulletin was issued by IE which instructed holders of operating licenses for Bab ock & Wilcox (B&W) designed reactors to take further actions, in-cluding immediate changes to decrease the reactor high pressure reactor 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) on page 1-7 of the
" Office of Nuclear Reactor Regulation Status Report to the Ccmmission" dated April 25, 1979. After a series of discussions between *.he NRC staff and the licensee concern-ing possible design modifications and changes in operating procedures, the licensee agreed, in a letter dated May 1,1979, to perform prcmptly certain actions.
The Comission found that cperatica of the plant should nct be resumed until actions i
I described in paragraphs (a) through (e) of paragraph (1) of Section IV of the Order.
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were satisfactorily completed.
41E Sulletins Nos. 79-05 (April 1,1979),79-05A (April 5,1979), and 79-05B l
(April 21,1979) apply to all B&W facilities.]
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.e 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 i
i May 16, and June 12, 15, 22 and 29, 1979, and numerous discussions with the licensee's staff.
Confirmation of design and procedural changes was made by members of the NRC staff at the Crystal River site. An audit of the Crystal River reactor operators was also performed by the NRC staff to assure that the design and procedural changes were understood and were being correctly implemented by the operators.
i EVALUATION I
Item (a)
It was ordered that the licensee take the follnwir.g action:
i "tJpgrade the timeliness and reliability of delivery from the i
l Emergency Feedwater System by carrying out actions as identified in Enclosure 1 of the licensee's letter of May 1,1979."
The Crystal River emergency feedwater (EFW) system design has one turbine-driven cump that.is automatically ac uated and controllec independent of offsite pcwer, and cne motor-driven EFW pump that is automatically started if offsite power is available, but must be manually started on a vital AC bus if offsite power is 1
lost.
By reference above to Enclosure (1) of the licensee's letter of May 1,1979, it was ordered that the licensee:
i 1.
" Review procedures, revise as necessary and conduct training to ensure timely and proper starting of motor-driven emergency b
857 26 1
3-feedwater (EFW) pump from engineered safeguards bus A upon loss of offsite power."
The licensee has revised EP-101 (" Unit Blackout") to provide the operators with a procedure for loading the motor-driven EFW pump on engineered safeguards bus 3A. This procedure will be used only if the following three conditions are met:
(1) loss of offsite power, (2) the turbine-driven pump is not function-ing, and (3) EFW is required. The procedure directs the operators to strip (remove) the following loads from bus 3A prior to loading the motor-driven EFW pump on the bus:
(1) the decay heat removal pump, (2) building spray. pump, (3) closed loop cooling pump, and (4) raw water pump.
The loads stripped from the bus are not required for shutdown cooling during the period that EFW is used for decay heat removal. However, a redundant decay heat removal train would be available on the other emergency bus.
When EFW is no longer needed, the motor-driven EFW pump would be removed from the bus and the other loads restored t'o the emergency bus.
The NRC staff performed an audit at the site and verified that t'.te operators were knowledgeable in the steps of this procedure.
The NRC staff concludes that the licensee has adequate procedures and the operators art properly trained to start the motor-driven EFW pump fro.a the diesel powered engineered safeguards bus 3A upon a loss of offsite ocwer, and therefore, meets-the requirements of this part of the Order.
It was also ordered that:
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2.
"To assure that EFW will be aligned in a timely manner to inject on all EFV demand events when in the surveillance test mode, pro-cedures 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."
Surveillance Procedure SP-349 (" Emergency Feedwater System Operability Demon-stration") directs the operators to close the EFW pumps' discharge valves to perform the test and then provides direction to reopen the valves to their normal operating positions following the test.
These discharge valves are motor-operated valves that are controlled from the main control room; therefore, there is no.need to station an operator at the valve locations during the surveillance testing.
SP-349 requires that an operator determine that the EFW system valves are properly aligned and a second operator verifies the valve positions following the test.
In addition to the independent verification of valve lineups required following surveilla.;ce testing, independent verification is also required upon completion of maintenance on the system. A valve lineup check list for the EFW system is included in SP-300 ("0perating Daily Surveillance Log").
The NRC staff has conducted an audit to verify that the operators are aware of the valve lineup requirements.
The NRC staff concludes the licensee has developed adequate procedures and has preserly trained cperators to verify correct valve alignments in the EFW system, and, therefore, is in ccmpliance with this part of the Order.
857 26
It was ordered that:
3.
" Emergency feedwater bypass valves are normally.a Uc e.nen position.
Procedures have been develo?ed and implemented to require the opcrator to take control of these valves upon failure of the ICS steam generator level control.
If the ICS level control does net fa'11 the operator will close_ the bypass valves. Those valves in the EFW system not locked in position are verified to be in the proper position on.a daily basis. Training will be conducted on these revised procedures prior to June 1,1979."
Emergancy feedwa.2r flow is normally controlled by the' integrated control system (ICS) to maintain the required steam generator levels by actuation of the air-operated feedwater startup valves. An alternate path, independent of the ICS, is provided through the motor-operated EFW bypass valves.
The bypass valves are normally maintained in the open position; however, following EFW activation, the operators must close the bypass valves and monitor steam generator levels and EFW flow to determine if the ICS is functioning properly.
If the ICS fails to maintain the prns. steam generator levels, the operator is directed to centrol the level by throttling flow with the bypass valves.
The licensee has modified the following emergency precedures to provide this guidance to the cperators: EP-101 (" Unit Bl ackout"), EP-103 (" Loss of RC Flow /RC Pur.p Trip"), EP-106 ("Less of Reacter Ccolant or Reactor Coolant Pressure"), and EP-108 (" Loss of Steam Generator Feed").
857 26E
The licensee has installed an ultrasonic flowrate meter system to provice control room indication of emergency feedwater flowrate in gallons per mi;;ute (spm) to each steam generator.
Each system consists of the ultrasonic ficw transducer, mounted on the EFW piping, and the associated flow display computer,-
mounted locally.
Flowrate indicators are also located in the control room on the rsin control board.
In addition to the directions for operator control of EFV flow if required, the licensee has provided fc a daily verification of valva lineup in the EFW system in SP-300 (" Operating Daily Surveillance Log").
The NRC staff has conducted an audit at the site and verified that the-operators are trained in these procedures.
The NRC staff concludes that the licensee has provided adequate procedures and operator training to control the EFW system independent of the ICS, and is thus in compliance with this part of the Order.
It was also ordered that:
4.
"The EFW pumps will be verified operable in acccrdance with the CR #3 Technical Specifications and Surveillance Procedures."
The Technical Specifications for CR-3 require a m:,nthly test of the turbine-driven E Fn' pump to demonstrate its operability.
The surveillance procedure requires run-ning both of the EFW pumps with their discharge valves cl0 sed, with flow through the recirculation line, and measuring the discharge pressure of the pumps. We have reviewed the test procedure and find it acceptable.
Satisfactory results of gg 261
this monthly surveillance test is an acceptable basis for demonstrating the operability of the EFW pumps and, therefore, we conclude that the licensee is in compliance with this part of the Order.
The licensee was also ordered to:
5.
" Review and revise, as necessary, the procedures and training for providing alternate sources of water to the suction of the EFW pumps."
Energency Procedure EP-108 (" Loss of Steam Generator Feed") provides adequate direction to the. operators for providing alternate sources of water to the suction of the EFW pumps. The primary source of water to the EFW pumps is the condensate storage tank (CST), which has a capacity of 150,000 gallons.
The operator is alerted by a level alarm on the CST when the level drops to 89,000. gallons.
The condenser hotwell, which has a capacity of 200,000 gallons, is the alternate source of water to the EFW pumps. The procedure directs the operators to open the motor-operated valves that will connect the hotwell to the suction of the EFW pumps and then to close the motor-cperated suction valves from the CST.
The procedure also contains instructions on valve operation to provide a third source of de-mineralized water from the water treatment system if needed.
The NRC staff, at the site, has verified tnat the control rcom cperators are properig trained to carry out these procedures.
The NRC staff concludes that the itcensee has complied with the req'irements to review and revise procedures and operatur training for providing alternate sources cf water for the EFW system, and, is thus in compliance with this portion of the Orcer.
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The licensee was ordered to:
6.
" Remove the interlock which prevents the turbine-driven emergency
The licensee has removed the interlock.
The turbine-driven.EFW pump will start, if required, regardless of the motor-driven EFW pump status.
Based on the above design modification, we conclude that the licensee has complied with this portion of the Order.
It was also ordered that:
7.
"In the event emergency feedwater is necessary and offsite power is available, an auto start signal will be provided to the motor-j driven emergency feedwater pump."
{
The licensee has installed circuitry to provide automa;ic starting of the motor-driven EFW pump if offsite pcwer is available.
The auto start signals include either of the following:
(1) coincident less of both main feedwater pumps, sensed by the loss of control oil pressure; cr (2) coincident low-low steam generator Itvel in both steam generators, detected by existing and new equipment in the ICS.
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Provisions have been included to manually bypass the loss of main feedwater pumps signal to allow for startup and/or shutdown. The bypass switch is keylocked, with annunciation and administrative control.
lhe steam generator low-low level signal is not bypassed.
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In addition, the licensee has modified the turbine-driven EFW punp start circuitry to include tfie same set of signals. Previously, this pump automatically started only on loss of both main feedwater punps.
Based on the above design modificatio'ns, we conclude that the licensee has complied with this portion of the Order.
It was also ordered that:
e 8.
" Design review and modification, as necessary, will be condacted to provide control room annunciation for auto start conditinns of the EFW system."
The licensee has provided control rcom annunciation to alert the operators that the EFW punps (motor-driven and/or turbine-driven) have started when required or f ailed to start when required.
Tne conditions which initiate the above alarms include the same signals as discussed in Part 7 above!
(1) loss of main feedwater; or (2 )
Icw-low level in both steam generators.
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These signals are combined with the,punp status (start or fail to start) to provide the annunciation. Based on the above modifications, we conclude that i
the licensee has complied with this portion of the Order.
l It was also ordered:
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9.
" Verification has been made that the air-operated level control valves (a) fail to the ~0% open position upon loss of power to the electrical / pressure converter, and (b) fail to the as is position upon loss of instrunent air.and electrical power to I
the air lock. At full load these valves are in the full (100%)
I open positions and at, low power levels (below 15%) they are partially open controlling flow.
If these valves were to fail r"
j closed, feedwater flow would be controlled using the EFb bypass valves as dest ribed in Item 3 above."
i The licensee has completed its verification tests of the failure made of the air-operated level control valves. The results show that one air-operated level control valve fails to a 54; open position and the other fails to a 47% cpen position upon loss of electrical powr to the electrical / pressure converter. These failure positions are within acceptable ta'.erance of the 50t open position specified in the Order.
On a test for loss of instrument air, both air-operated level control valves failed as is, i.e., remained at approximately the 5C% open position during the test.
The EFW bypass valves are motor-operated regulating valves which are operated independently from the ICS as discussed in part 3 above.
If the ai r-operated level control valves redain closed or ICS fails, EFW flow would be mar.ually 857 270
_ 11 _
controlled using the EFW bypass valves.
We conclude that the licensee has' satisfied this portion of the Order.'
C onc1usion:
Based upon our evaluation of Parts 1 through 9 above, we conclufe that the licensee has upgraded the timeliness and reliability of delivery from the EFW system by carrying out.the actions icentified in Enclosure 1 of the licensee's letter of Pay 1,1979, and is, therefore, in compliance with Item (a) of the Order.
Item (b)
The licensee was ordered to:
" Develop and implement operating procedures for initiating and control-ling emergency feedwater independent of Integrated Control System Control."
The NRC staff has reviewed the revised procedures for the EFW system to assure that there is sufficient guidance for the operators to actuate the system if. automatic initiation f ails and to control steam generator levels at the required values. The,.
The NRC staff review of the procedures and the operator trai'iing focused on whether-the operators were directed to observe the proper instruments and whether operators wre given specific values of parameters, such as steam generator level, to main-
, tain by operating the control valves. The review also determined that the validity of the instrument readings of certain key parameters, such as steam generator level, would be confirmed. The modifications to the procedures to satisfy these deter-minations were verified by the NRC staff.
(See further discussicn of these procedures in part 3 of It em ( a) ).
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857 271
We will require the licensee to perform a test during power ascensicn (less than 15% rated power) to demonstrate the' capability to provide and control EFW flow to both steam generators.
The primary objective is to verify that the operators can initiate EFW and control steam generator levels independent of the ICS.
A metber of the NRC staff will witness the test and verify acceptability prior to authorizing the licensee to proceed to full power operation.
The NRC staff audited a sample of Crystal River operators to determine if they were familiar with the revised procedures and could implement them correctly.
Based on the NRC audit, we concl ude that the revised procedures and operator training are satisfactory, and, therefore, the licensee is in compliance with item (b) of the Order.
Item (c)
~The Order required that the licensee:
"knplenent 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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The CR-3 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 cotain 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
.icensee committed to install a hard-wired, control-grade reacter tri.p on the loss of all main feedwater and/or on turbine trip (letter from B. L. Griffin (FPC) to H. Denton (NRC) dated May 1,1979). The purpose of th~is anticipatory trip is 9
857 273s
to minimize the potential for opening of the PORY and/or the safety valves on the pressurizer.
The licensee has added control 'rade circuitry which is designed to provide an g
automatic reactor t' rip when either the main turbine trips or all main feedwater is lost.
The main turbine / generator trip is sensed by an existing, pressure switch in the turbine electro-hydraulic control system.
On a turbine trip, the pressure switch energizes a normally deenergized relay in the ICS. A contact from this relay is arranged in a normally energized circuit containing two parallel reactor trip actuation relays. Deenergizing both of these relays provides an output to energize the 125 DC volt shunt trip coils of the two reactor trip breakers.
Ene r-gizing both reactor trip breakers trips the reactor.
The loss of main feedwater is sensed by either of two signals: loss of the main feeawater pumes or low-low level in both steam generators (the same signals which start EFW).
The signals are generated separately for each feedwater path.
Any one of these signals will energize a relay in the ICS (cne relay f.or each feedwater,'
pat h).
The centacts from these relays are arranged in the same circuity as the react]r trip actuatien re16ys such that any coincidence of signals frem the two feedwater patns will deenercize the relays, in the same manner as the turbine trip, causing a reactor trip.
Provisiens have been inciided to automatically bypass and reinstate the loss.of main feedwater pum; and turbine trip signals at less than 10% power to allow A7 g]
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for normal startup and shutdown of equipnent without tripping the reactor.
Operator verification of the bypass removal is required by procedure during
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power ascention.
The licensee has analyzed this additional circuitry with respect to its independ, -
ence from the 9-isting reactor trip system. They have stated that the shunt coil ~
is part of the existing AC reactor trip breaker. Each shunt coil is powered by a separate Class IE 125 volt supply and operates independently from the 120. volt AC undervoltage trip coil of the same AC reactor trip breaker, which receives a safety-grade reactor protection system trip signal..
An IRC inspector has confirmed that the checkout tests for the circuitry were completed successfully.
In addition, the licensee has cormiitted to perform a monthly test on the added circuitry in order to demonstrate its ability to open the AC reactor trip circuit breakers.
Based on our raiew 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.
Based on the licensee's design modifications and comitment to perform a monthly test on the new circuitry, we conclude that there is reascnable assurance that the system will perform its intended function.
Based on the above evaluation, we conclude that the licensee is in compliance with the requirements of Item (c) of the Order.
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.15 -
Itam (dk This item in the Order required the licensee to:
" Complete analyses for potential small breaks and develop and implement operating instructions to define operator action."
By letter dated May 1,1979, the licensee ccmitted to prociding the analyses and operating procedures of this requirement.
MW, the reactor vendor for the Crystal River plant, st.bmitted analyses entitled,
" Evaluation of Transient Behavior and Small Reactor Cociant System Breaks in the 177 Fuel Assembly Plant" and supplements to these analyscs (References 1 through 6).
Tne major parameters used in this generic study bound the Crystal River plant.
i The staff evaluation of the MW generic study has been completed and the results of the evaluation will be issued as a IUREG report in July 1979.
A principal finding of our generic review is a reconfirmation that loss-of-coolant accicent (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 and the high pressure injection (HPI) system combined with cperator action ensure adequate core cooling.
The EFW system, used to remove heat through the steam generators, has been modified to enhance its reliability as discussed in Item (a).
The HPI system is capable of providing emergency core cooling up to the safety valve pressure setpoint. The ability to remove heat via the steem generators has always been recognized to be an important consideration when analyzing very small breaks. Separate sensitivity analyses ( for breaks
/
D 857 275 Smaller than 0.01 sq. ft.) were performed assuning permanent loss of all' feed-water (with operator initiation of the HPI system at 20 minutes), and loss of Uncovering of the feedwater for only the first 20 minutes of the accident.
reactor core was not predicted for these events. The calculated peak cladding 0
temperature'was less than 800*F, well below the 10 CFR 50.46 requirement of 2200 F.?
These results are applicable to Crystal River considering the ability to manually start the redundant EFW punps and HPI punps from the control room, assuning failure of automatic EFW actuation.
Another aspect of the study was the assessment of recent design changes on the lif t frequency of the pressurizer PORY and safety valves.
The design changes in-ci tded :
(1) a change in the setpoint of the PORY from 2255 psig to 2450 psig; (2) change in the high pressure reactor trip setpoint from 2355 psig to 2300 psig; and (3) the installation of an anticipatory reactor trip on turbine trip and/or
' loss of all main feedwater.
In the past, during the turbine trip or loss of feed-water transients, the PORY lifted. With the design changes, the initial pressure increase of these transients do not result in lifting of this valve.
However, the consecuent depressurization could initiate HPI which could repressurize the system and lift the PORV valve.
It is expected that the operator would terminate HPI before the PCRV or safety valves lift, since the 50*F subcooling criteria would be satisfied at pressures below the PORV setpoint.* Also, lifting of both the PORV and safety valves might occur in the cases of control rod withdrawal or in-advertent boron dilution transients, using the normally conservative assumptions found in the Final Safety Analysis Report Chapter 15 safety analyses. The abcve Une 50'F subcooling criteria is discussed en page 20 of this evaluatien.)
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-re design changes do not effect the lift frequency of the valves for these Chapter 15 safety analyses.
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Based on our review of the small break analyses presented by B&W, thd staff has i'
determined that a loss of all main feedwater with (1) an isolated PORI [ but safety valves opening and closing as designed, or (2) a stuck open PORY does not result in uncovering the reactor core, provided either EFW or HPI (2 punps) is initiated within 20 minutes. Based on the consequences calculated for small
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break LOCAs and loss of all main feedwater events, and taking into account ex-pected reliability of the EFW and FPI systems, we conclude that the licensee has complied with the analyses portion of Item (d) of the Order.
To support long-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 analyses 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 cf the recent tEC " Staff Report of the Generic Assessment of Feedwater Trans~ients in Pressuri:ed Water Reactors Designed by the Sabcock and Wilcox Company" (NUREG 0560). Further details on these analyses and their applicability to other PWRs and SWRs will be specified by the staff in the near future.
In addition, to assist the staff in developing more detailea guidance on design requirements of PCRV and safety valve reliability during anticipated transients, as discussed in Section 8. 4. 6 cf the MJREG 0560, the licensee will be required to provide analyses of the lift frequeray and mechanical reliability of the pressurizer PORY and safety valves of the Crystal River facility.
7 857 27 1 The B&W analyses show that seme operator actions, both immediate and follow-up, are required under certain circunsta'nces for a small break accident. Immediate
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operator actions are defined as those actions, comitted to memory by the operators, which must be ' carried out as soon as the prr,h. is diagnosed. Follow-up actions require operators to consult and follow the steps in written and approved pro-ced ures. These procedures must always' be readily available in the control room for the operators' use.
Guicelines were developed by S&W to assist the operating E&W facilities in t,he develognent of emergency procedures for the small break accident. " Operating Guidelines for Small Breaks" were issued by B&W on May 5,1979, and reviewed by the NRC staff.
These guidelines were revised on May 15, 1979, to include
~ revisions recoc nended by the staff (Reference 7). In response to these guidelines, the licensee made substantial revisions to EP-106 (" Loss of Reactor Coolant /RC System Pressure"), and EP-103 (" Loss of RC Flow /RC Pump Trip"). These emergency procedures define required operator actions in response to a spectrum of break sizes for a LOCA in conjunction with various equignent availability and failures.
EP-106 ("Less of Reactor Coolant /RC System Pressure") is divided into three sections.
The first section deals with a leak or rupture that is within the capability of one makeup punp.* In this case, the operators proceed with an orderly plant shutdom, if the leak is in excess of the Technical Scecification limits.
The stcond section of EP-106 defines required operator actions for a snall break tnat is within the capability of the HPI system to maintain RCS pressure and
'L4 CR-3 Ine HOI pumps are used for makeup pumps.]
g 2n pressurizer level.
This assunes that the initial break was of a size sufficient to cause a depressurization with a resulting reactor trip and FPI actuation.
This part of the procedure provides the operators with the guidance necessary to achieve a safe hot shutdown condition for a variety of degraded conditions. 'I f '
all feedwater is lost, a heat removal path is established by the HPI system through the break and 'the pressurizer PORV or the safety valves.
Once feedwater is reestablished, the steam generators can be used as a heat sink.
If the reactor -
coolant pumps are not available, the operator is directed to EP-103 (" Loss. of Reector Flow /RCP Trip") which defines the actions necessary to establish natural ci rc ul ation. Additional guidance is provided in EP-103 if natural circulation is not imediately achiesed.
This includes "bunping" reactor coolant punps or if they cannot be operated, using the PORV to control RCS pressure ur.til either forced flow ur natural circulation can be achieved.
If natural circulation has been established and plant conditions are stable, the operator is directed to AP-ll3
("Reacter Cooldown by Natural Circulation").
If forced circulation is established, the normal plant cocidow1 procedure (OP-209, " Plant Cooldown") is used in con-junction with EP-106.
The third section of EP-106 deals with a larce pipe rupture in which the system depressurizes to the point of low pressure injection (LPI).
Tne system response is not de;:endent upcn the availablity of reactor coolant pumps or feedwater and, therefore, no other procedures need be referenced.
For all cases in which HPI is manually or autcmatically initiated, the operators are specifically instructed to maintain maximun HPI flow unless one of the folicw-ing criteria is met:
79 857 2M
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(1) the LPI system is in operation and providing cooling at a rate in excess of 1000 gU*and the situation has been stable for 20 minutes; or
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(2 ) all hot and cold leg temperatures are at least 50 degrees below the iaturation temperature for the existing RCS pressure.
If the 50 aegrees subcooling cannot be maintained af ter hPI cIkoff, hPI shall be reactuated.
A requirement to determine and maintain 50*F subcooling has been incorporated in n
all other procedures in which HPI has been manually or automatically initiated.
These procedures include, EP-105 (" Steam Supply System Rupture"), EP-108 (" Loss of Steam Generato,r Feed"), EP-101 (" Unit Blackout"), and EP-103 (" Loss of RC Fl ow/ RC Pum p Tri p"). Each of these procedures, in addition to EP-106 (" Loss of Reactor Coolant /RC System Pressure") procedure, provide additional instructions to the operators in the event of faulty or misleading indications. A subset,sent action statement directs the operators to check alternate instrunentation i
channels to confirm key parameter readings. The Crystal River staff has made revisiens to all of tneir emergency procedures to ' include this confirmation.
The CR-3 incere thermoccuples will be hard-wired to a dedicated monitoring system which is programmed to alarm at high temperature.
In addition, the operators will be able to check all input readings and/or get a printout of the status of each thermocouple with this system. A process computer in the control room is also i
I available to provide this indication.
I If feedwater is not initially available following a transient or accident, core i
l cooling is maintained by flow from te HPI pumps and relief through the PORV, 857 280
t,
which is opened by the cperator.
B&W has performed studies that show density differences between the downcomer arEd reactor core will cause recirculation flow between the core exit'and downcomer via the vent valves.
Mixing of the hot l
I core exit water with the cold }PI water will provide sufficiently warm vessel l
tenperatures to preclude any signifi[ ant thermal shock effects to the vessel.
l Under these' conditions, with no circulation of water' from the steam generators, the cold leg resistance temperature detectors (RTD) may not provide a satisfactory indication of the vessel temperature. B&W has recomended using the core exit thermocouples as a measure of vessel temperature, based on B&W analyses that l
r conservatively show that the vent valves will open at temperature differences between the core exit and downcomer of less than 150*F.
They have also proposed a more appropriate pressure-temperature limit curve for the vessel that reflects allowable stresses under these faulted conditions (no feedwater).
i The IRC staff has reviewed these guidelines and finds them acceptabla based on the expected recirculation through the vent valves and the vessel stress limits used. The licensee has incorporated these revised guidelines into Emergency Pro-cedure EP-108 (" Loss of Steam Generator Feed").
Subsequent restoration of EFW would depressurire the reactor coolant system to below 600 psig where pressure vessel integrity is assured for any reasonable thermal transients that might sub-sequently occur. We concluce that further reliability analyses are needed as part of the long-term requirements of the Order to ccnfirm that EFW can be restored (if lost) in a reasonable period of time. B&W has agreed to provide a cetailed thermal-mechanical report on the behavior of vessel materials for these extreme conditions, to be applicable generically to the Ccenee class of plants, which incl udes Crystal River.
i 857 2R lhe Crystal River. Unit 3 main control board has an annunciator which alarms when the PORV scienoid is energized (to open the valve). In addition, there are 3 indicating lights which are actuated by 2 selector switches of the valve control circuitry. The green light is lit when the " AUT0-OPEN" selector switch is.in the "AUT0" position.
In this position, the pressure signal will provide the open and close control of this valve. The red light is lit when the same switch is in the "0 PEN" position.
In this position, the selector switch will control the valve (to the open position). The amber light is lit when the ',-
"NCR.M0" selector switch is in the "LO" position.
In this position, the low pressure protection circuit is operable and can cren the valve for this mode of operation.
EP-106 (" Loss of Reactor Coolant /RC System Pressure") was reviewd by the tRC 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 member of the NRC staff walked thrnugh this emergency procedure in the Crystal.
River centrol recm. The procedure was judged to provide adequate guidance to the cperators to cope with a small break LOCA.
The instrumentation ",necessary to -
diagnose the break, the indications and controls required by the action state-ments, and the administrative controls which prevent acceptable limits frem 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 shutdom con-cition in the event of a small break accident.
An audit of 8 of the 28 licensed operatcrs assigned tc shift duty was conducted by the tRC staff to determine the cperators' understanding of the small break k
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.857 accident, including how they are required to diagnose and respond to it.
The Crystal River staff has conducted special training sessions for the operators ori the concept and use of EP-105 and other emergency procedures related to the small break ac:ident. The audit revealed several deficiencies in the knowledge of the small break phenomenon and the requirements of the procedure.
Mditionally, there were deficiencies in the knowledge of the details of the recent design modifications made to the Crystal River plant.
These deficiencies were primarily the result of design modifications and procedure revisions not finalized at.that tirre'.
As a result of the audit, each licensed individuai received additional training by the plant training organization and by the General Physics Corporation (G PC). This additional training has been completed and verified by the NRC staff.
A subsequent reaudit of 10 licensed individuals by the NRC revealed satisfactory res ul ts.
The audit of the operators also included questiornng about the TMI-2 incident and the resulting impact on the Crystal River plant.
The discussions covered the initiating events of the incident, the response of the plant to the simultaneous loss of feedwater and small break LOCA (PCRV stuck open), and the operational actions that were taken during the course of the incident. We identified a,.
deficiency in interpreting the initial sequence cf the TMI-2 incident on the part' of several of the operators. Additional training has been conducted in this area by the plant staff and their consultant GPC. and his been verified by the NRC staff.
In summary, we found their level cf understanding sufficient to be able to respond to a similar situation if it happened at Crystal River.
We als: ccncl ude they 3
857 28L have adequate knowledge of subcooling and saturated conditions and are able to recognize each in the primary coolant system by several methods. The EFW system was also discussed during the audit to determine the operators', ability to assure proper sf.arting and operation of the system during normal conditions, as well as,during adverse conditicns such as loss of offsite power or loss of normal feedwater. The long-term operation of the system.was exam.ined to evaluate ~
the operators' ability to use available manual controls and water supplies.
The level of understa'nding was found to be sufficient to assure proper short-and long-tenn EFW flow to the steam generators.
In addition to the oral audit conducted by the PRC, the licensee administered a written exanination to all licensed personnel.
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 90 percent is attained on an equivalent, but different exam. The written exam and the grading were audited by the NRC staff and j udged to be acceptable.' The staff will also review all subsequent results and records as part of the normal inspection function of the Crystal River re-qualification program.
We conclude that there is adequate assurance that the operators at Crystal River have and will continue to receive a high level of training concerning the TMI-2 accident and the consequent impact on their unit.
Sased on the foregoing evaluation, we conclude that the licensee has complied with the requirements of item (d) of the Order.
857 281
Item (e)
The Order required that:
"All licensed reactor operators and senior reactor operators will have completed the TMI-2 simulator training at B&W."
The licensee has confirmed that all reactor operators and senior reactor operators have. completed 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 demon-stration of the event on the simulator as it occurred and how it should have been cont roll ed. The class discussion was about one hour long and the remainder of the four hour session was conducted on the simulator.
The TMI-2 event, incl uding 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 resulted in depressurization and saturation conditions were presented to the operators in which they maneu/ered the plant to a stable, subcooled condition.
Based on the foregoing evaluation, we conclude that the licensee has complied with the requirements of Item (e) of the Order.
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CONCLUSION We conclude that the actions descrilbed above fulfill the requirements of our Order of May 16, 1979, in regard to Paragraph (1) of Secticn IV.
The licensee having met the requirements of Paragraph (1) may restart Crystal River as pro-viced by Paragraph (2).
Paragraph (3) of Secticn IV cf the Crder remains in
,8s 857
26 -
force until the long-term actions set forth in Section II of the Order are completed and approved by the PRC. '
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857 28%
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REFERENCES 1.
Letter from J. H. Taylor (MW) to R. J. i%ttson (NRC) transmitting report entitled, " Evaluation of Transient Behavior and Small Peactor Coolant System Breaks in the 177 Fuel Assembly Plant," dated May 7, 1979.
l l
2.
Letter from J. H. Taylor (MW) to R. J. Mattson (NRC) transmitting I
revised Appendix 1, " Natural Circulation in MW Cperating Plants l
l (Revision 1)," dated May 8,1979.
3.
Letter from J. H. Taylor (MW) to R. J. Mattson (NRC) transraitting additional information regarding Appendix 2, " Steam Generator Tube Tnermal Stress Evaluation," to report identified in Item 1 above, i
f~
dated Pay 10, 1979.
4.
Letter from J. H. Taylor (MW) to T. M. Novak (NRC) providing background f
I information on react 6r coolant punp operation, dated May 10, 1979.
i 5.
Letter from J. H. Taylor (MW) to R. J. Mattson (NRC), providing an analysis for "Small Ereak 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 1, dated May 12, 1979.
6.
Letter f rom J. H. Taylor (MW) to R. J. Mattson (NRC), providing
[
Suppl ement 3 to Section 6 of report in Item 1, dated Vay 24, 1979.
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.857 288
v 7.
Letter from J. H. Taylor (B&W) to 2. R. Rosztoczy (NRC) transmitting revised " Operating Guidelines for Small Breaks," dated May 16, 1979.
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