ML19242C007
| ML19242C007 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 06/19/1979 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML19242C005 | List: |
| References | |
| NUDOCS 7908090649 | |
| Download: ML19242C007 (29) | |
Text
I
~
EVALUATION OF LICENSEE'S COMPLIANCE i
WITH THE NRC ORDER DATED MAY 7, 1979
^
SACRAMENTO MUNICIPAL UTILIT( DISTRICT
.s' RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 5
y
\\)
790 8090 hy};f r
s
l l-l INTRODUCTION
~
By Order dated May 7, 1979, (the Order) the Sacramento Municipal Utility _
Oistrict (SMUD or licensee) was directed by the NRC to take certain actions l
with respect to Rancho Seco Nuclear Generating Station.
Prior to this Order f
{
and as a result of a preliminary review of the Three Mile Island Unit No. 2 h-l (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 subsequently instructec to take a number of immediate actions to I
avoid repetition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (IE).* Subsequent'ly, an additional bulle+in was issued by IE which instructed holders of operating licenses for B&W designed reactors to take further actions, including immediate L
changes to decrease the reactor high pressure trip point and incrtase 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 S&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 Commission" dated April 25, 1979.
After a series of discussions between the NRC staff and the licensee concerning possible design modifications and changes in operating procecuras, the licensee agreed, in a letter datec April 27, 1979, to perform promptly certain actions.
The Commission found that operation of the plant
- [IE Bulletins Nos. 79-05 (April 1,1979),79-05A (April 5,1979), and 79-058 (April 21, 1979) apply to all B&W facilities.]
iC,
should not be resumed 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 f
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 14, 22, 24, 29, 30, and June 6, 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 Rancho Seco site.
An audit of the Rancho Seco reactor operators was also performed by r.e NRC staff to assure that the design and procedure changes werc understood a'd were being correctly implemented by the operators.
EVALUATION Item a It was ordered that the licensee taka the following action:
%fcgrade the timeliness and reliability of delivery from theatnetMy baw;;e2-Ibasteg by carrying out actions as identified in Enclosure 1 of the licensee's letter of April 27, 1979."
The Rancho Seco auxiliary feedwater (AFW) design has one turbine / motor tancem drive pump (P-318) that is automatically actuated and controlled independent
_g.
cf offsite power, and one motor-driven AFW pump (P-319) that is autcmatically started, but must be manually transferred to a vital AC bus if offsite pcwer is lost.
Th'e turbine / motor driven pump will be manually started, according to procedure, from a vital AC bus if the turbine drive fails.
By reference above to Enclosure (1) of the licensee's letter of April 27, 1979, it was ordered that the licensee:
"1.
Review procedures, revise as necessary and conduct training to ensure timely and proper Ear:tisgat.gorgr.ArJy.Wim r icm grt&TAFW)~ptraisyL--b Cus I upon loss o f of fsi te pcwe'. "
The licensee has alevdeoed:Secturdd;nfdperatinghw+-MMb %d (gedwater Systen") to provide specific direction for the operator on the steps required to load motor driven pump P-319 on nuclear service bus 4A and to secure the steam to the turbine on the dual-drive pump P-318, in the event of inoperability of the steam drive, and load the motor drive on nuclear service bus AB.
typass keys are required to complete the connection of the auxiliary feedwater pumo motors to the d:esel pcwered buses (nuclear service buses 4A and 43); these keys are available in the office adjacent to the control rocm.
Emergency Procedure 0.1 (" Load Rejection") directs the operator to use Cperating Procedure A.51 if main feed pump operation cannot be maintained.
The NRC staff verified that the cperators are kncwledgeable in the procedure for loading the AFW pumps en the vital AC buses.
The NRC staff concludes that the licensee has adequate precedures and the operators are trained to start tne AFW system '" m diesel pcwered buses upcn less of offsite power or load rejection and therv ;.
is in compliance with this part of the Order.
~
,iv
-3
h It was also ordered that:
f
~
t
{
"2.
To assure that APd 'will be aligned in a timely manner to inject _on 1
all AFd demand events when in the surveillance test mode, procedures f
will be implemented and training conducted to provide an ---
-- e t
l Wweessa&valvesaar-1.2--.icz n a n w w.a. w e G, h m j
during.the surveillance mode to carry out the valve alignment changes upon APd demand events."
Lettreillance-Precedures-5P ekholAw.u 4F.J.10.Olft are used for the quarterly surveillance and inservice testing of auxiliary feed pumps P-318 and P-319, r 2s pectively.
These procedures have been revised to includ.e the follcwing g
statement; " Station an operator at FWS-055, auxiliary feedwater system full I
ficw recirculation valve in continuous communication with the control room until Pd5-055 is secured closed at the completion of this test."
In addition to the above procedure revisions, the licensee has added FWS-492 (bypass valve for F45-055) to the " Locked Valve List" (SP 214.03).
The licensee has also incorporated independent verification of valve lineups folicwing surveillance testing and/or maintenance of the APd system.
The NRC staff has reviewed SP 210.01A and SP 210.013 to verify that the precedures contain specific directions to return each valve that was operated during the conduct of the surveillance test to its prcper position.
The local cperator has to close a valve (F45-055) when so instructed by the control _ roca operator i
)$
4 c.
t or if he loses communication with the control room.
The NRC staff has verified that the operators are familiar with this test procedure.
We conclude that the licensee has adequate procedures to assure that AFW will be aligned irt a timely manner to inject on all AFW demand events when in the surveillance test mode and therefore, is in compliance with this part of the trder.
It was ordered that:
i
[dEted "3.
Procedures will be devele plemented and Mr 1
to provide for emi vf steam p,srowr.devel:by uw.misda);y grade AFW typass. valves in the wee 4Nsteam w.udweitnrel 6n' trot' fail s. "
The licensee has developed Emergency Procedure 0.14 (" Loss of Steam Generator Feed") that describes the symptoms that would result from a loss of main feedwater control that may have been caused by an integrated control system (ICS) failure.
The procedure has been reviewed by the NRC staff.
The operator is directed to restore feedwater to the steam generators by one of three methods.
The preferred method is described in jggtimsh4.r =dymsIurf
/MURTTTir~y +enteh4ystem":).
Section 7.7 directs the operator to:
May suqiiCS corrtrolled AFW rarrrrul.1ralyes; Stact.Ihe.AStaumps; and d'ritertin Se'4%d generatw2evels, scecified in the procedure, by manually operating the strturr4riva5WW1:ryc. ass valves Aca the. control Toom.
In this mode the pumps and valves will operate independent of the ICS.
The operator is provided with AFW flow rate and steam generator level indications in the control rocm for each steam generator.
.,,c 6
l t-5 e
Since the AFW bypass valves will fully open on a safety features actuation signal (SFAS)*, the operator is provided with instructions on how to take manual control of the valves after a SFAS.
NRC staff has conducted an audit of the operator training and verified that the operators have been trained to carry out those procedures.
The NRC staff concludes that the licensee has developed adequate procedure:
and operator training to control APW flow to the steam generators to specified values independent of the ICS, should a failure of the ICS occur, and therefore, is in compliance with this part of the Order.
It was also ordered:
"4. dierification that Technical Specification requirementstAFW capac.ity are in accordance with the accident analysis will be conducted.
Pump capacity with mini flow in service will also be verified."
The licensee has conducted the verification thet Technical Specification requirements of AF4 capacity are in accordance with the accident analysis for the Rancho Seco Nuclear Station.
The Technical Specification states, as a
'[The safety features actuation system (SFAS) monitors variables to detect loss of reactor coolant system boundary integrity.
Upon detection of "out-of-limit" conditions of these variables, it initiates emergency core cooling (ECC) which consists of high pressure injection (HPI) and icw pressure injection (LPI), Reactor Building cooling and isolation, and Reactor Building spray systems.
Additionally, it starts diesel generators GEA and GEB, which are in standby reduncance with the nuclear service buses 4A and 48.]
i r
I I
i t
I limiting condition for operation, capacility to supply feedwater at a m l
fhr*sqmescendThgYa decay heat DarnhafMmecent:ncessttr.Pactee
~
WW*nst - -
' =- a6 Moving v.eansy (a) a condensate pump and a main feed pump, or (b) a condensate pump, or (c) an auxiliary feedwater pump.
A letter fren Babcock & Wilcox to the licensee, dated May 16, 1979, states that it has performed an analysis of the required AFd flow rate for the Rancho Seco Plant which shows that g 4 cqy.-heat is.EL uf 4.5 m
~2 Osma Wel',
@:tha.heak-input fr-os the ilCPsp:willmquir9 Mew % Wor both steam generators of owroximately-760 ppn.
liach of the two AW pumosare sized todeliver480 gpm-twaam: generators
- N gom mini-flow.ird service.. This rump capacity exceeds the minimum required AFW flow rate in the Rancho Se:o safety analysis and Tecnnical Specifi-cations.
AFd pump capacity, with mini ficw in service, has been w s.$5d ey performing the quarterly " dad Systa-SurveiMancalest" and the "MH273 BE 6A 1M icCindi catm functtN45TP i l 2 ).
The resu?ts of these tests demcnstrated that each of the two AFd pumps has the capability to deliver a ainimum of 780 gpm into the steam generators, with mini flow in service.
~83'2TCenNWTireconff rm the'miniarum AWTiow' rate tete staa:rgenemo s 4
g, EUthWWikt*Ty%T164 Tog-startup.
[ki.tEl wwa x;... J y g i a.
t
Cased on our review of the AFW flow rate test results, performed to date, we concludethatthelicenseeisincch.pliancewiththispartoftheOrder.
It was also ordered that:
"5.
Mcdifications will be made to provide gcifipth 1-e ntr5T7 rsauf AFW flow to each. steam generator.f To verify that AFW is being purped to the steam generators, the licensee has installed Clampitron Ficwmeters on both of the AFW injection flow paths, downst? cam of the AFW control valves, so that the actual ficw rate t'o each stea::. generator will be measured.
The QDmitron_-Elowmeters consists of transducers, attached to the AFW piping, connected to a flow display computer.
On command from the ficw display computer, the transducer transmits an ultra-sonic beam through the water inside the pipe and the velocity of the beam, as affected by AFW ficw, is analyzed by the flow display computer, which calculates the AFW flew rate in gpm.
The AFW ficw rate is displayed in the control room.
SectTibrathn. test (STP-612) was.csacucted by the licensee to functionally test the performance of the ficwmeters.
Performance of this test demonstrated that the indicated ficw rate agreed with the calculated flow rate within the 120% acceptance criteria specified in the prccedure.
2ased en our review of this design modification and test results, we conclude that the licensee is in compliance with this part of the Order.
-g-t'
~
i i
i It was also ordered that the licensee:
l "6.
Review and revise, as necessary, the procedures and training fon-l providing attarnata sources ef wter tttthe,suctienTf tha *FW l
pu:::ps."
i Control rocm alarms are available to alert the operator to perform the manual transfer of the AFJ supply source from the condensata storage unctuT) to the giant reservoir.
The CST is de:igned to seismic Category I criteria.
The licensee has reviewed and revised his Emergency Procedures 0.10 (" Loss of Reactor Coolant Flow /RCP Trip"), 0.14 (" Loss of Steam Generator Feed"), and Operating Procedure A.51 (" Auxiliary Feedwater System") to. provide guidance for the operator to obtain an alternate source of watec for the suction of the AF4 pumps.
The revised procedures require the operatcr to break condenser vacuum when the level reaches a level alarm point of epproximately 29 feet and
" to shift the AF4 pump suction to the plant reservoir when the CST level is-down to a second alarm point of approximately 3 feet frcm the bottom of the tank.
Tartrapacity vi me tST 4s-targe lneogn'to prov.ua xsolin @ m.
amers&fo~ e this transfer is tequired.
The shifting to an alternate source r
of AFJ pump suction is accomplished by manually operating four isolation valves at a local valve station.
Tasmeurawr rmre-mLmu> eyigeety abe w er.
The NRC staff has reviewed the revised Emergency Procedures 0.10 and 0.14 and Cperating Procedure A.51 and concludes that these procedures provide sufficient guidance to the operator for a timely shifting to en alternate water source for the AFJ pumps, before the CST is emptied.
S.
i
'iV
_g_
r
i i
i The NRC staff has verified that the control rocm operators are properly trained to carry out,these procedures. 'de conclude that the licensee has complied with the requirements of this part of the Order.
It was also ordered that:
"7.
Design review and modification, as necessary, will be conducted to sprovide control rocia annunciation for ali auto start conditions of the AFV system."
~
The licensee has provided indication for all auto start conditions of the AFW system on an annunciation panel inside the control room.
The cuxttt4 W " W.4 pil actuate the annunciator are:
(a) bss of_all reactor coolant pumps, or,
(b) bw discharge pressure.(850 psig) on both main feedwater.pumpf,'dr
~
(c) penual -start of the mter driven MV pcmp.
A safety features actuation signal, which will also automatically start AF#,
had already been annunciated in the control roca before the current modifications.
Based on our review of this design modification, we conclude that the licensee is in compliance with this part of the Order.
It was ordered that:
.4 r
"8.
Procedures will be develoced and implemented and tecining conducted to provide guidance for timely apernt-i+ir't=*
-f w 4 e taattiatiorrefMFW. "
~
The conditions that will automatically initiate auxiliary feedwater are adequately i
described in reperating Procedtrre ~A;51,W" Auxiliary 1eecwatarW').
The operators are directed, as an immediate action, to verify that the AFW flow has autcmatically started on loss of both main feedwater pumps in Emergency Procedure 0.14 (" Loss of Steam Generator Feed") and on loss of all reactor coolant pumps in Emergency Procedure 0.10 (" Lass of Reactor Coolant Flow /RCP Trip").
Both procedures require the following immediate actions by the operator:
grifyJ. hat _the auxiliary feedwatersumps have.antomaticaliystarted; that there is gelow to the. steam generators; and that thtStoper25 team gene.eator W ls are being maintained.
The NRC staff has performed an audit and verified that the operators are trained in these procedures.-
Based on review of these procedures, we conclude that the licensee has provided guidance for timely operator verification of any automatic initiation of APd and therefore, is in ccmpliance with this part of the Order.
It was also ordered:
"9.
Verification will be made that the air operated level control valves (a) Fail to the Sc% coen oosition.woondess of.electricabpower to
- i
-n-I
the electrical to pressure converter, and (b) Fail to the MP position upon learxrr-se5 ceiir.
The APd by pss valves are _ W.
gcade."
The licensee nas pepletedjAs yptifjcatiottet for the failure mode of the air operated level control valves.
The test results show that both air operated level control valves fail to the IH)0% upensosition-sn ; css,fm m trre s
at the valve operators.
On tests for loss of control signal to the electric to pressure converters, one level control valve failed to the 5CRaparrpesTtten and the other one failed to the 60% open m it.icn; which are acceptable.
The
~
Q byptssplves:arepfetyJrade, gd;pt hated valves which are operated independently from the ICS as discussed in Part 3 above.
8.ased on our review of the test results on the air operated level control valves and the safety grade design of the bypass valves, we conclude that the licensee is in compliance with this part of the Order.-
Based upon our evaluation, we conclude that the licensee has upgraded the timeliness and reliability of delivery frcm the AF4 system by carrying out the actions identified in Enclosure 1 of the licensee's letter of April 27, 1979, and therefore, is in ccmpliance with Item (a) of the Order.
Item (b)
It was cedered that the licensee:
J/ 'i g.
I i
" Develop and implement cperatin,gcqadure,s for sii.uamem -% n.s I
w #'# = '= M 7 ;cdepe
- t=at W egra W lantee %'d151-2 % 1. "
e-We have reviewed the revised procedures for the AFW system to assur? that there is sufficient guidance for the operator to actuate the system if the automatic initiation failed, and to control steam generator levels at the required values.
The review of the procedures focused on verifying that the operator is directed to observe the proper instruments and that the operator is directed to maintain specific values of parameters by manual control, such as steam generator levels.
The review also determined that the operator should confirm the validity of the instrument readings of certain ke'y parameters, such as steam generator levelc.
The necessary modifications to the procedures to satisfy these requirements 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 in part 3 of Item (a.)
Tusevdicensee will tencuct.-aMartupr test 4ulspawr M5%}-tar-demchrace-thewenastlity-to cronce 2nd cent.colGow. Ao tne steam generators,semmythe M tryears valves.
During the visit to the site, the NRC staff walked through the AFW procedures with the-operators to evaluate whether the procedures were functiona'ly adequate.
In addition, the NRC staff audited a sample of Rancho Seco operators to determine if they were familiar with the revised procedures and cculd imolement them correctly.
Cased on the NRC staff audit, we conclude that the revised procedures and operator training are satisfactory a~nd therefore, the licensee is in comoliance with Item (b) of the Order.
jdd
~13-1
Item (c)
~
~
The original Rancho Seco design did not have any direct reactor trips that would be initiated'by a malfunction in the secondary system.
To obtain an_
anticipatory reactor trip (rather than delaying the trip until a primary system parame.ter exceeded its trip setting) the licensee committed to install a hard-wired, control grade, reactor trip on loss of all main feedwater and/or turbine trip.
The Order requires that the licensee:
" Implement a hard-wiresi centrol grade reactor trin that would be actuated on Jess of maindeeowsi.cr aml/ar.turetne' trip'."
The licensee has added control grade circJtry to Rancho Seco, which is designed to provide an automatic reactor trip when either the main turbine trips ~or all main feedwater is lost.
The purpose of the 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).
The main turbine trip is sensed by an existing, normally deenergized relay in the main turbine / generator protection system.
The relay is energized by the protective trips of the turbine and/or gene ntor. 4tbwer;is3upplied_by_aq esentar;catteryarcurce.
The loss of all main feedwater is sensed by amowlyaiaWied.y.vuure m was m en'ar m,
- mindeadot <>mw-uii@- U nes ).
The J T'
pressure switches actuate (close) cn icw pressure in tne header.
Pcwer is supplied by the same gj te--bd 44.cy-we.
In order to prevent an inadvertent reactor trip' during startup or shutdown, the bass,4durma Ehkw pnr um,sme.nf W.Ch.q i,j ranytagenzwitch.
The key for this switch is maintained in the custody of the shift supervisor and is located in the control rocm.
When the switch is placed in the " cut-out" position, it is gammeiaIm2..crLthe main corrtrol voard.
The operating procedures specify when the switch is placed in the " normal" or " cut-out" position.
Either signal (turbine trip or loss of all main feedwa'ar) will actu' ate a reactor trip relay, which in turn provides an input to both of the shunt coils of the AC reactor trip breakers.
Energizing both of the shunt coils causes a reactor trip.
The licensee has analyzed this additional circuitry with respect to its independence frca the existing 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 VCC supply and cperates independently frca the 120 VAC undervoltage trip coil which receives the safety grade reactor trip signal.
dagCdaspector.4asefirmed that -tne enecz oct. tests-fm-uus-arcumi Sve.beegrempJeted. success fully.
In addition, the licensee has committed to perform a--~ dy peticdic. test'en the added circuitry in order to descnstrate its ability to cpen the AC reactcr trip breakers via the shunt coil.
f Le
= -.. -
Based en our review of the implementation of the trip circuitry, with respect to its independence from the existing reactor trip circuitry, we conclude that this addition will 'not ' degrade the existing reactor protection system design.
Based on the licensee's design modifications and commitment to perform a monthly test on the new circuitry, we conclude that there is reasonable assurance that the system will perform its function.
On the basis of the evaluation above, we conclude that the licensee has complied with the requirements of Item (c) of the Order.
Item (d)
This item in the Order requires the licensee tc:
"Cemplete analyses -for potentiaE1 mall-breaks and develop and implemeot operating instructions to define operator action."
In the licensee's letter of April 27, 1979, the licensee committed to providing the analyses and operating pr0cedures Of this requiremen*,
Babccck and Wilcox, the reactor vendor for the Rancho Seco plant, submitted analyses entitled, 'bEvrluauvu vi hous e Beharf ur.- -5matthewr i.uolant System Weits Jin the :177 fuel Assembly Plant" and sunclements to these analyses (References 1 through 6).
The major parameters used in this generic study 4
l"
't
?
bound the Rancho Sec'o plant.
The staff evaluation of the 8!.'4 generic study has been completed and the results of the evaluation will be issued as a NUREG report in June 1979.
A principal. findirg of our generic review i4-n a h J a =J =rhme jnWe@dGCA}-analyses -of breaks-.at,4he;4owee-ewef&satH-c: mat :xsectww healier.;than 0.04 sq. - ft.) demonstrate that,a..combinatierWJ2 eat. remysi.ty i pe -steam generators, the high pressure injection system and spusi.ur4mn etnsure adequate core ccoling.
The APd 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.
The ability to remove heat via the steam generators has always been recognized to be an important consideration when analyzing very sr.all breaks.
Separate sensitivity analyses were performed assuming permanent loss of all feedwater (with operator -
initiation of tN high pressure injection system at 20 minutes) and loss of feedwater for 4.y the first 20 minutes of the accident.
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.
These results are applicable to Rancho Seco considering the ability to manually start the redundant AF4 pumps from the control rocm, assuming failure of automatic AF# actuation.
Another aspect of the study was the assessment of 'recent design changes on the lift frequency of pressurizer safety and relief valves.
The design changes
- 7-t
included:
a change in the setpoint of the mweim Atcyalaled;;;gitief valve (PORV) f ry; QhWw1450g+
-t srse in the %W' n. - J trip setpoint fromag15i_ - 43M nsi; and the installation of an anticipa-tory reactor trip on turbine trip and/or on loss of all main feedwater.
In the past, during the turbire trip or loss of feedwater transients, the PORV lifted. With the design changes the initial pressure increase of these tran-sients do not result in lifting of this valve.
Hcwever, the consequent depressurization could initiate safety injection which in turn could repres-surize the system and lift the relief valve.
It is expected that the operator would terminate HPI before the relief valve or safety valves lift, since the
~
50 F subccoling criteria would be satisfied at pressures below the PORV setpoint.
Also, lifting of both the PORV and safety valves might occur in the case of control rod withdrawal or inadvertent boral dilution transients, using the normally conservative assumptions found in the Chapter 15 safety analyses.
The above design changes do not effect the lift frequency uf the valves for these Chapter 15 safety analyses.
Based on our review of the small breck analyses presented by S&W, uness:caffI ph ned -that e-foss vf all main -freersterMth;h) er-tseistett.-PORY;'
bar.csafety wives t:oermuredchshg as-engneti,mrdiGs- %dufa?
@ pes e -result -ia com..,.-evy, vi vdded eitherrAFVwrtWI ptaros7s' auitiated withh -20 minutes.
Based on the acceptable consequences calculated for small break LOCAs and loss of all main feedwatar events coupled with the expected reliability of the ATW and HPI systems, we conclude that the licensee has complied with the analyses portion of Item (d) of the Order.
e r
I To support langer term operation of the facility, requirements will be developed for additional and more detailed anaiyses of loss of feedwater and other h et24 bdanalysessf salL*reakdisCalm,4jtu4 anticipated transients.'
Wso Tt!$itd ? L Jis.purposcAccording1y ~-the ;i icensee vim.*Ew a r.d 1.0 e
grovide:.theenalyses4iscussad 4a.SecticaJ.4. hand AA.4,1nthe-cecentdRC.;
f htafi. Report of the Generic Assessment of Feecwater_IransientsdnJressurtzed
%1ter seactors Designed by the Babcock and_Wilcox Ccmpa @ @ REG'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 assisti.the staff in developing more detailed guidance on design requirements qf relief and safety valve reliability during anticipated transients', as gdiscussed in Section 8.4.&of the NUREG report, Se4iceaseevilbhrecuired ta provide analyses of thedift frequency and mechanical reliability of the -
cressurizer relief and safety valves of the Rancho Seco 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 conmitted to memory by the operators which must be carried out as scon as the problem is diagnosed.
Follcw up.
actions require operators to consult and folicw the steps in written and approved procedures.
These precedures must always be readily available in the control roca for the operators' use.
Guidelines were develcped by B&W to assist the operating S&W facilities in the development of emergency procedures for t.9e small break accident.
m
/
-1C-r-
l
' The "Carating Guidelines for Small Breaks" were issued by 3D on May 5,1979 and reviewed by the NRC staff.
Revisions recommended by the staff were inc,.porated in the guidelines.
In response ~to.these guidelines, the staff at Rancho Seco made substantial revisions to Emergehcy Procedure 0.5 (" Loss of _
Reactor Coolant / Reactor Coolant System Pressure") and Operating Procedure B.4
(" Plant Shutdown and Cooldown").
These procedures define ths 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.
Guergency Procedure 0.5 (EP D.5) is divided i".to three sections.
The ditst :
c:ection deals with a small leak pithin :the capaoility of as;;axeup pumo In this case, the operators proceed with an orderly plant shutdown unless pres-surizer or makeup tank levels fall below prescribed limits.
If these limits are exceeded the reactor is manually tripped and high pressure injection is initiated.
The gecond section of EP 0.5 definet the required operator action for a small break got within the capanility st+eautm pump.
This section provides the operator with the guidance necessary to achieve a safe hot shutdcwn condition for a variety of degraded conditions.
If all feecwater is lost, a heat removal path is established by the high pressure injection system through the break and the pressuri:er pcwer operated relief valve or the safety valves.
Once feedwater is reestablished, the steam generators can be used as a heat sink.
~
If the reactor cociant pumps are not available, the operator is directed to Cperating Procedure 3.4 (CP B.4) which defines the actions necessary to cool down the plant by natural circulation.
Additional guidance is provided in OP B.4 if natural circulation is not immediately achieved.
_2g_
i, The M t..C u of EP D.5 defines the actions necessary in the event of a M
rd.
In this case the system depressurizes to the point of low
~
pressure injection.-
~
ikirallmases_.in-which_bigh.Sressure injection as sc.nuallyfore44tomaticakjy initiated,ethe_rcerators are specificaUyinstru.i.cd,Mfb5 txr.arurtain gr.aximts _HPl. flow unless one of the following critaritate. met:
(1) The 1.PI system is in operation and providing cooling at a rate in excess ofyl600 cpm and the situation has been stable forattmeg, or (2) All hot and cold leg temperatures are at least 80cf.egrecsrbelow the pituration. temperature for the existing. RCS pressure.
If the 50 degrees subcooling cannot be maintained after HPI cutoff, HPI shall be reactuated.
A 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," and " Loss of Steam Generator Feedwater."
Each of these procedures, in addition to the " Loss of Reactor Coolant / Reactor Coolant System Pressure" procedure, provide additional instructicns to the operators in the event of faulty or misleading indications.
<i T'
A subsequent action st2tement directs the operators to check alternate instru-The Ranc*o Seco staff mentation channels to confirm key parameter readings.
r has made revisions to all of their emergency procedures to include this require-ment. 55"hthstarc.:is_r.at., initially availabitt:follArinGA#ansieggrMdent, T; fire.cooliM -is MintAined.hy flN frca.twojiPI:pu=ps and. relief.4hrJugh ;tt*13 ADORV, etcA.is,coaned.by-AhaAperator. Am?ct;2,hese -degraded.SooM99muditimas,
Wie pressure-temperature. limits considered.in.Eigure,.,LM;2 ofnthelechnicals 45pecifications are not applicable to.the ensuing depressurization and cooldown Ebecause these limits were developed for normal and upset operating conditions aply.
Density differences between the downcomer and reactor core will cause recirculation ficw between the core exit and dowr.ccmer via the vent valves.
Mixing of the hot core exit water with the cold HPI sater will provide suffi-ciently warm vessel temperatures to preclude any significant thermal shock effects to the vessel.
lyosequent-restoration of AFW would depressurize.the'
- pactor toolant system to below 600 psi wnere pressure vessel integrity is assured for any reasonable thermal transients that might subsequently occur.
We conclude that kr.ther;re, liability. analyses are.needed as partaf2the.
' gig-term requi rements -of tne -Grder-*c.ct:cfirm ~that AFW < ant be-rastomif
'est) in a reasonable period of time.
Ski _+.magreed to. provide. e.-etailed tyrmai-mechanical report xn the behavior of vessel materials dor:these=es.m CAnditiCns7 to -De applicable generically to the OConee Class of plants,NiCh mcludes ' Rancho Seco.
The "Less of Reactor Coclant/ Reactor Coolant System Pressure" procedure was reviewed by the NRC staff to detarmine its conformance with the S&W' guidelines.
Ccmments generated as a result of this review were incorporated in a further revision to the procedure.
A :e 1ber of the NRC Maff walked ; rough this mg o esrw
>p r ~ rr em.,. T rr ju
((
l
,I b qq
.g(
a n i,, e y g @L t
a v \\s un
(
emergency procacure in the Rancho Sece control room.
The procedure was judged to provide adecuate guidance to the operators t0 cope with a small break loss-of-ccolant accident.
The instrumentation necessary to diagnose the _
break, the indications and controls required by the action statements, and the administrative controls which prevent unaccectable limits frca being exceeded are readily available to the operators.
We conclude that the operators should be able to use this precedure to bring the plant to a safe shutdcwn condition in the event of a small break accident.
An audit of seven of 14 licensed operators and senior operators assigned to shift duty (22 total licensed personnel) 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 Rancho Seco staff has conducted special training sessions for the cperators on the concept and use of EP 0.5.
The audit revealed that, except for one deficiency, the operators had sufficient knowledge of the small break phencmenon and the requirements of the procedure.
This ceficiency, verification of natural circulation, was brcught to the attention of the plant staff.
Each licensed individual received additional training in this area by the plant training organization anc General Physics Corporation.
They also received training on the revisiens made to E? D.5 as a result of the NRC review.
This additional training has been c:mpleted and verified by the NRC staff.
The audit of the coerators also included questioning about the TMI-2 incident and the resulting design changes made at Dancho Seco.
The discussicns covered i
E*
~
the initiating events of the incicent, the response of the plant to the simul-taneous loss of feedwater and small break LOCA (PORV stuck open), and the operational actions that were taken during the course of the incident.
We identified a deficiency in interpreting the initial sequence of the TMI-2 incident on the part of several of the operators.
Additiooal training has been ccnducted in this area by tha plant staff and their consultant and verified by the NRC staff.
Otherwise, we found their level of understanding sufficient to be able to respond to a similar situation if it happened at Rancho Seco.
We also concluded they have adequate knowledge of subccoling and saturated conditions and are able to recognize each in the primary coolant system by various methods.
The AFW system was also discussed during the audit to determine the operators' ability 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 normal feedwater.
The long term operation of the-system was examined to evaluate the cperators' ability to use available manual controls anc water supplies.
The level of understanding was found to be sufficient to assure proper short and icng term AFW ficw to the steam generator;.
In addition to the oral audit conducted by the NRC, the licensee administered a written examination to all licensed personnel.
Indivicuals scoring less than 90 percent on the exam will receive additional training and will not asst.me licensed duties until a score of at least 90 percent is attained on an equivalent, but dif ferent exam.
The written exam and the grading was audited by the NRC staff and judged to be satisfactory.
The staff will also review bs i
.i 3
V
~
I l
r all sucsequent results and rec:rds as part of the ncrmal inspection function of the Rancho Seco requalification program.
We c nclude that there is adequate assurance that the' operators'at Rancho Seco have and will continue to receive a high level of training concerning the TMI-2 accident and the consequent impact at their station.
Based on the foregoing evaluation, we conclude that the licensee has complied with the requirements of Item (d) of the Order.
Item (e)
The Order requires that the licensee:
"grovide for one seniardicensed operator-assigned.to.the-controLroom who has had THI-2 training on the B&W simulator."
Th'e licensee has confirmed that this item of the Order has been completed and has further ccmmitted that all reactor operators and senior reactor operators smaave cceoleted the TMI stimulator ' training at B&V by June 21,m/9.
This training consists of a class discussion of the TMI-2 event follcwed by a demonstration of the event en the simulator as it occurred and the proper actions that should be taken to control the accident.
The class discussion is about fcur hours long and the remainder of the session is conducted en the simulater.
The TMI-2 event, including cperational errors, is demonstrated to each cperator.
The event is again initiated and the operators are given
" hands un" experience in successfully regaining control of the plant by several F:
'J' t
methods.
Other transients which result in depressurization and saturation conditions are presented to the cperators and they must maneuver the plant to a stable, subcooled condition.
Based on the above commitment by the licensee, we conclude,that the licensee is in ccepliance with Item (e) of the Order.
Conclusion We conclude that the actions described above fulfill the requirements of our Order of May 7,1979 in regard to Paragraph (1) of Section IV.
The licensee having met the requirements of Paragraph (1) may restart Rancho Seco as provided by Paragraph (2).
Paragraph (3) of Section IV of the Order remains in force until the long term modifications set forth in Section II of the Order are completed and approved by the NRC.
Cated at Bethesda, Maryland this 19th day of June 1979.
9.
s 1
REFERENCES 1.
Letter from J. H. Taylor (B&W) to R. J. Mattson (NRC) transmitting report entitled, "E/aluation 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 (S&W) to R. J. Mattson (NRC) transmitting revised Appendix 1, ' Natural Circulation in B&W Operating Plants (Revision 1),"
dated May 8. 1979.
3.
Letter from c.
H. Taylor (B&W) to R. J. Mattson (NRC) transmitting additional information regarding Appendix 2, " Steam Generator Tube Thermal' Stress Evaluation," to report identified in Item 2 above, dated May 10, 1979.
4.
Letter from J. H. Taylor (B&W) to R. J. Pattson (NRC), providing an analysis for "Small Break in the Pressurizer (PORV) with no Auxiliary Feedwater and Single Failure of tne ECCS," identified as Supplements 1 and 2 to Section 6.0 of report in Item 2, dated May 12, 1979.
S.
Letter from J. H. Taylor (3&W) to R. J. Mattson (NRC), providing an analysis for "5 mall Break in the Pressurizer (PORV) with no Auxiliary Feedwater and Single Failure of the ECCS" identified as Supplements 1 and 2 to Section 6.0 of report in Item 2, dated May 12, 1979.
s
l t:'
g 6.
Letter from J. H. Taylor (SL'd) to R. J. F%ttson (WC), providing Supplement 3 to Section 6 of report in Ite: 2, dated May 24, 1979.
O g
e 0
e f