ML13317A340
| ML13317A340 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/20/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Dietch R Southern California Edison Co |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TAC-47549, TASK-A-49, TASK-OR LSO5-82-09-064, LSO5-82-9-64, NUDOCS 8209240329 | |
| Download: ML13317A340 (39) | |
Text
September 20, 1982 DISTRIBUTION Docket NRC PDR Local PDR Docket No. 50-206 ORB Reading NSIC LSO 09-064 DCrutchfield HSmith WPaulson Mr. R. Dietch, Vice President OELD Nuclear Engineering and Operations OI&E Southern California Edison Company ACRS (10) 2244 Walnut Grove Avenue SEP GVissing Post Office Box 800G Rosemead, California 91770
Dear Mr. Dietch:
SUBJECT:
PRESSURIZED THERMAL SHOCK AUDIT REPORT San Onofre Nuclear Generating Station, Unit No. 1 An! audit team consisting of NRC and contractor personnel visited San Onofre Unit 1 on June 3 and 4, 1982 to audit the adequacy of operator training and emergency procedures with regard to pres surized thermal shock. Enclosed is a copy of the audit report for your information.
Section 5 of the enclosed report lists five recommendations resulting from this audit. We request that you review this report and provide a response within 45 days of receipt of this letter that addresses the action being taken with regard to-these recommendations.
Sincerely, Original signed by Dennis M. Crutchfield, Oief Operating Reactors Branch #5 p5 Division of Licensing
Enclosure:
Audit Report cc w/enclosure:
See ndxt page 8209240329 820920 PDR ADOCK 05000206 P
PDR DL: ORB#
DL ORB #5 OFFICEO...
SURNAME P.,,...
P.......
.X DATE
.9
/82...
/82
//82 F.........* **.
- 80) R.020O F C A R E OC O
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se FORM 318 (10-80m uNRC 02n OFFICIAL R ECOR D COPY USGPO: 1981-335-960
Mr. September 20, 1982 cc Charles R. Kocher, Assistant General Counsel James Beoletto, Esquire Southern California Edison Company Post Office Box 800 Rosemead, California 91770 David.R. Pigott Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 Harry B. Stoehr San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Resident Inspector/San Onofre-NPS c/o U. S. NRC P. 0. Box 4329 San Clemente, California 92672 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego San Diego, California 92101 California Department of Health ATTN:
Chief, Environmental
..Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 U.-S. Environmental Protection Agency Region IX Office ATTN:
Regional Radiation Representative 215 Freemont Street San Francisco, California 94111 Robert H. Engelken, Regional Administrator Nuclear Regulatory Commission, Region V 1450 Maria Lane Walnut Creek, California 94596
ft San "fre ULni t 1 ProcedulreS nd Tr~aini n' for PresurizedThe~rml Shock C:ONTENTS INTRODUCTION 1
1 Short-Term Objectives and Scope of Review.
.1
.1.2 Current Status of the Generic PTS Issue.
2 1.:S SONGS 1 General Characteristics.
SHORT-TERM CRITERIA USED FOR SONGS 1 AUDIT 7
2.1 Transient and Accident Analyses.
7 2.1.1 SONGS 1 Overcooling Events Summary 2.1.1.1 Event 1 -
April 30, 1972.....
9 2.1.1.2 Event 2 -
October 21, 1973
.9 2.1.1.3 Event 3 September 3, 1981 10 2.1.2 SiS 1 TermiCnat1io'Cr r
r n
ia..
2.1.2.2 Iain Feedater Pfumps (MFWPs).
11 2 1.2.
- Charging, Saey Iniecti-on (SI
.~
Auxiliary.
Feedater (AFW) Sytm 11 2.1.7 Thermal Hydraulic Analysis 17 2.1.7.1 SONGS 1 Plant Specific Analysis.
17 2.1.7.2 W4CAP-100:.19 Vessel Integrity Analysis 1s 2.2 Criteria for Procedure Reviews 19 2.3 Criteria for Training Program Review 20 DESCRIPTION OF AUDIT
- 2.
7.1 Procedures 2
-. 2 Traninrg -
25 KE4 FININGS FROM THE SONGS 1 AUDIT 26 1
r eo Of Procedures with the Audit Criteria
.26
4.1 1
Procedures Should Not Instruct Operators to Take Actions That Would Violate NDT Limits
.26 4.1 2 Procedures Should Provide Guidance on Recovering From Transient cr Accident Conditions Without Violating NDT or Saturation Limits 6
4.1.7 Procedures Should Provide Guidance on Recovering From FTS Conditions 27 4.1.4 PTS Procedural Guidance Should Have a LSupporting Technical Basis.
.2 4.1.5 High Fressure Injection and Charging System Operating Instructions Should Reflect a Consideration for PTS 27 4-.1.
Feedwater (FW) and/or Afux iJliary Feedwater (AFW) Ope-rating Instructions Should Reflect S TS Cocerns 8
4.1.7 An NDT Curve and a Saturatiion Curve Should
- e Provided in the Control Rom 4-2 CompalFrison of Tr~.-
ning, &.ith kidit CriteriE-2?.
4.,l Trainino Should Include Specific instruction on NDT Vessel Limits for Normal Modes of Operation 4.2.2 Training Should Include Specific Instructions on NDT Vessel Limits for Major Transients and Accidents 29 4.2.3 Training Should Particularly Emphasize Those Events Known to Require Operator Response to Mitigate PTS.
9 4.3 Summary o-f Findin,2s.
4.7.
1 PrC:edC!UreG-S 4.3.2 Training.
.1.
4.:.2.1 Training Material....
4.-3.2.2 Operator Inter..iews..
31 5
FRECCM1EiD ATIONS REFE ENCES.
.1 T.hugt-Term Ubiective a.na Srcooe or-~'
r*e,'
On June 3 and 4, 1982, a m.ti-disciplinary audit team visited San Onofre Nuclear Generating Station Unit 1 (SONGS
- 1) to evaluate certlin aspects of the Pressurized Thermal Shock (PTS) i sue.
The question that the audit team focused on was:
ARE-CORRECTIVE ACTIONS REQUIRED THAT MUST BE INITIATED BEFORE THE LONGER TERM PTS PROGRAM PROVIDES GENERIC RESOLLTON IAND ACCEPTANCE CRITERIA?
Emerget-ncy proce ures nd operator training qwere the only areas in
.which the SONGS 1 audit team applied the above general question.
As noted in the NRR March 9. 1982, presentation to the Commission:
we wil.
undertake a
program to verify that existing operating procedL.res contain the steps necessary to prevent and/or matigate PTS
- events, and to verity that operator d ducati on./trai ni.n programs regarding PTS are acceptably 1
T co of: the review was limited by several factrr:
- 1) The laes mr, c
proc edures w ri tten T(or SONGS 1
have not comfpleted the In-li nt rev iw cyc Ie and therefore have not been approved for use.
These procedures were reviewed by the audit team because they were based on the latest Westinghouse Owners Group (WOG) technical guidelines and because they incorporated PTS considerations based on FTS scenari os rom "Summary Report on Reactor Vestel Integrity for Westinghouse Operating Plants" (WCAP-10019).
- 2)
Some of the operators ha-e had little or no opportunity to train on the new procedures.
Z3)
Recommendations made-in Section 5
of this report do not include changes which ma'y-oe requirea as a result of the longer term generic program (USI A-49)
+/-.e Current Status of the Generi c F'TS Issue Efforts to pursue an integrated PTS program involving a
variety of technical areas are continuing under LSI,r-49.
The summer of 1983 is the - current schedule
-for finalizing the generic regulatory reuirements for F TS along with required corrective actions if the e-ri.c req'uirehents are not met.
Key issues are yet to be resolved and extensive programs exist to provide the foundation for the generic reqIintry requirements.
!E-li t e +
C.
ti n n
regula
~tory reoqu renCts
~
colTated
- hoever, the staff.
has committed to the Commission to have ceveloped a7kn interim initial position for the summer of 19B2 (June).
The i nteri m
i n iti al posi ti on Aill consist of NRC evatiOF of the oContinued plant operation (and initial corrective actions required) for all of the eight plants previously identified as representative of plants having the highest RTNDT.
PNL has been contracted to work with the staff to provide recommendations regarding the.une 1982 initial position on the safety of continued operation and tc recommend any additional corrective actions that should be niti ated before the NRC generic resOlution End acceptance criteria are adopted.
The June recommendations by the NRC staff to the Commission will also consider the findings and recommendations addressed in Sections 4 and 5 of this
- report, as well as the.findings of other audit teams formed for rel ated i nvesti gati ons (such
-as
.f I uence reduct i on at the vesse wal).
1 1.3 SONES I General Characteristics 5Ojb6 1
utilizes a
4c6 Mke Westinghouse three-loop design.
plant control is provided by a combination of rod control, pressurizer (Pzr) pressure and level
- control, and steam generator (SG) level control systems.
The reactor core is protected by the Reactor Protection Eystem (RPS) and the Safeguard; Load Sequencing System (SLSS),
- hich can ctuate t.e
.afet-, Inrecticin (SI
- system, used for long term core
a 1 a r,'
nd h~~
he Ez removahR a~ rdte L rztEflffeft Spray ctutio System CSA)
.used for mcontrol of containment
- pressure, tenE-rature, and radiological releases.
The S1 S stem 1s cmpri sed od t o 5 pu mpc which take water from. the Re ueli ng Water Storage Tank (RWST; approx imately 4 0 DO ppm boron For reactivity control) and discharge to the Main Feedwater Pumps (MFWPs)
- which, on a Safety Injection Signal (SIS),
realign from their normal flowpath to inject water into the Reactor Coolant System (RCS) cold le. s Although the SIS is actuated at 1735 psig, the shutoff head: of the AFWPs (1200 psig.
prevents injection until the RCS pressure decreases below that point.
The SONGS 1
design does not include passive accumulators or high-head injection, although charging pump pumpo suction is aligned to the RWST on a SIS.
Feedwater is de-liv,,,ered from the condenser hotwell to the U-Tube steam gener.ators by four condensate pumps and two motor-driven 'MFWPs..
The uxiiliary Feedwater (AFW) sy.sten is actuated on an SIS since the MFWP's are realigned for cold leg injectien.
The AFW system consists of one motor-driven.
pump and one turbine-driven pump which both provide water from the Condensate Storage Tank (CST) to each steam qenarator.
The three steam generators feed a
common header which Cupplies steam to the turbin e via two main steam lines.
Each steam line has two atmospheric steam dump valves (10%
total steam relief CEmacit+:,
five code safetY valves (110% total steam relief capacity)
nd to conde~ner sta CdUIp Valv eS (10.
total steam rli ef capaci ty)
Ther areno main steam isolation valveE.
Th-,e E.OGS c con tr ol rqo om gives indication of the foll1owi ng major param eter to the operator fo:r assistance in monor events:
Pa rameters
-lr --
RCS Pressure One wide rane (WRa) ad onhe narrow ranQe (NR) recorder One WR nd three
!R indicators RCS Temperature One aver'age T-averige and
- thre, WR T-cold tone on each loop) recorders.
One NR T-average and one delta-T indicator for each loop.
One T-Mot indicator for each loop displayed at subcooling meter In-Core Temperature Meter selectable to any one of 55 in-core thermocouple positions,
.ubcoing~ M~rgi
~:Readout on subcooling margin in degrees F
based on RCS pressure and
?uctioneered highest hot 1 eg or in-core thermocou rple temperature Steam Generator Level One NR recorder and-one WR indicator for each SG.
Two safety-grade NR indicators on each SG for AFW system use Steam Generator Steam/Feed Flow One steam -flow and one feed flow recorder for each SS (on same recorder as NR SG 1-evel Safety Injection Flow Flow meters on each
.loop injection line Ch ring
'S t
yestem Fl ow 0ne
+fow meter on normal anjection patn.
One flow meter on each of the thre-alternate injection paths 1,+/- 1 Fecow--ter Fl. ow F.oLw meter on com:mon header
(total f
and one flow meter on each SG injection ine 2
SHORT-TERM CRITERI.
USED FOR SONGS I AUDIT The criteria for procedure and training reviews were based on transient and accident analyses.
The analyses and audit criteria are dis cUased below.
2.1 n
en n
ccidentn AnalySe Overcool i ng events in PWRs may occur as a resul t of 'main steam 1i.ne breai:s (MSLEs).
feedwater system malfunctions, steam generator tube Srupture (SGT)R' loss-of-coolant accidents (LOCA) or any situation which leads to the injection of cold water into the reactor vessel.
Multipie -Failures and/or operator errors can result. in more severe oivercooling events.
0+ particular concern are those events in which repressurizti on of the primary.
sy'Stm occurs following the severe overcooling.
Section 2.1.1 presents a summary of SONGS 1 overcooling events.
Terminatic-n criteria
-or primary and secondary injection systems are given in Section 2.1.2.
Aside from the primary mission of the. audi t team to examine procedures and training, a summary of the
heralshckE*vents is also provided (Section 2.1-73).
_:FSOiF-1 Lvercooli~ nEvents Summary A
review o
the operatina history of SONGS 1 has resulted in the identification of three overcooling events that could have led-to exceeding the cooldown limit if not. mitigated by automatic plant controls or operator action.
These events are discussed individually below.
Al1.
three events were the result of excess -feedwater flow to the steam generators causing overfill and cooldown of the primary eystem.
A 1l three events were terminated when the SI setpoi.nt was reached, since feedwater injection by the. MFWPs was automati call y isolated and realigned fpr SI (see discussion in Section. 1.3).
Termination of the couldown prevented RCS pressure from falling
-below the shLtoff head of the 1FWFs so no SI flow occurred.
In all three etentE, the total cooldown was limited to less than 100 F.
.. The emergency operating procedures require the operator to take immediate action to verify the SI valve alignment and to open or close.valves Manually as ap ropriate.
Operator action to terminate the cooldown for.
these events was not required.
In our appraisal, these events are n,-
significant as PTS PrecursorS.
ith the unit at 55 MWe during startup from an outage, a failure of the "C" feedwater controller resulted in a reactr trip on high "C"
steam generator level Overfilling this generator caused average-RCS temperature to fall from 550 F to about 460 F in 16 minutes and pressure to drop
-from 2035 to 1550 psig.
The event was terminated when the SI setpoint (then 1685 psig) was reached and SI was initiated.
Nine minutes after actuation the safety injection system was secured.
2.1.1.2 Event 2 -
October-21 19 Unit load was.being gradually reduced from 450 MW e to perform plant maintenance when a turbine trip and resultart reactor trip occurred.
At that time, the feedwater regulating system was programmed to open the regulating yalves to 80% open on any trip.
This resulted in a rapid filling of the steam generators and a cooldown of the RCS from 54-F to -470 F in about eight (8) minutes.
Initiation of SI at 1685 psi g. terminated the event..As a result of this events the feedwater regulating system was programmed to provide 5% flow on a reactor trip, thrbprev'/enting a recurrence oF this event.
O C
With the unit operating at 390 MWe, a
failure in the
- 1 Regulated Power Supply cauaed several alarms and the loss of several-plant aramster incacitions.
As a
- result, the operator manually tripped the
- plant, bur feedw.ater
+10o continued, resultind in an overfilling of the st-am generator.
This resulted in RCS temperature falling from 550 F to 430 F and pressure falling from 2077 to 1700 psig in about five
- 45) minUteS.
SI initiated and terminated the event at the new setpoint of 1735 psig.
As a result of this event, Westinghouse was consulted about the cooldown of the vessel.
They con-firmed that vessel shock was not a concern in this event.
2.1.2 San Onofre 1 Termination Criteria 2.1.2,1 Reactor Coolant Pumpg (RC~s)
The RCF's autom.tically trip on a turbine trip or on an SI when main aenerator is above 40% load.
-RCS temperature is greater than 50 F subcooled and RCS pressure is
- 00 psi above main steam pressure.
During a
steam generator t ube ruptUre (e TR one FCF may be restarted if RCS pressure is greater than..50 'psig (#1 seal delta-P requirements).
In0
On an SIS, the MFWPs are isolated from the normal feedwater lineup as described in Section 1..
- 2. 1. 2.3 Charaina. ---
et Ineto gl uiiary F
.Eedwa4t!er
(
"AW The termination criteria for these systems are dependent upon plant conditions.
The following tables describe these indications, termination criteria, and SUbSEquent-action -For Spurious SI actuation, loss of secondary coolant, loss of reactor
- coolant, and steam cenerator tLbe rupture; inacations:
RCS pressure 184C) psig and increasing Normal contLainment:
(1)
C ontainment pressure 1.4 psig (2)
Contai nment ra,diation be.ow setpoint a nd not increasing C. ontainment S
l evel 11
el op on t
and not increasing No 'Sec.
Rad.
alarms SIT L=rmiarion:
RC S P're ue 184U10E S-,c psip
.and increasing Pr Level:
- 25.
RCS Subcooling:
40 F Secondary Heat Sini::.
>-250 gpm AFW flow to S Gs or>
107. NR in at least 1 S G AFW Terminantion:
Throttle AFW flow to maintain 50% SG level Cha'rging Termination-Realiqn charging pump Suction from RWST to Vo1l.me Control Tank (VCT) and resume normal etdown Subsequent Action:
Proceed to hot shutdown 5 f eondary Coolai~nt Indications:
Main steam pressure 00 psig or 12
steam flow/feed flow mismatch prior to trip on at least two SB's SI Terminat-Lon:
A.
All RCS Temp-.
> 350 F
- 1.
Normal Containment RCS Press.:
.250 psig Pzr. Level: > 20%i RCS Subcooling: > 40 F Secondary Heat Sink:
25 gp m AFW to each unisolated SC and stable or decreasing RCS hot leg temperature
- 2.
Abnormal Containment Same as
- 1) except pzr. level of.
50%
due to possible degradation in pressurizer le\\vel indication B.
RCS Temperature < 350 F Same i!
e.cept RCES pressure
a 1100 psio to avoid loss of pressure vessel inteority
-F Trmi nation:
Throttle AFW 4lo to maintain BO NR e
level BUT DO NOT EXCEED RCS CODLDOWN FATE OF 100 F/HR Changing Termination:
Realign charging pump suction from RWST to Volume Control Tank.(VCT) and resume normal
.etdown.
Hold RCS cooldown rate to 50% F/hr Subsequent Action:
Depressurize RCS to 350 psig and initiate Residual Heat Removal (RHR) system Le s o'f Reacto'r Co"l ant Indications:
Abnormal containment or Pz r.
saFety valve StuCk open and RCS pressure -:.. 2485 psig or Fzr.
PORV stuck open and RCS pressure
< 2100 wig SI Termination:
RCS Fress-a:
1250 psig and 14
Fzr.
Level:.
50%
RCS Subcooling:
> 40 F Secondary Heat Si-nk:
250 g pm AF W flow t
'o SGs OR 10(A NR in at least 1 S
Throttle AFW flow to maintain SG level at 50 %
B.U.T D O NOT E',CEED RCS COOLDOWN RATE OF 100 F/HR (If possible)
Charging Termination:
IF SI has been terminated, realign charging suction from RWST to VCT and resume letdown.
Hold RCS cooldown rate to 50 F/hr.
If SI has.not been terminated, realign charging for cold leg injection (60C) gpm -flow)
A:bsequnt Actions:
If 5I has been terminated, proceed to hot standb'.
I-F SI has not been terminated, depressuri RS to 350 osia and initiia ate RHRiF-sytem 15
I.7 indication:
High air ejector radiation gr:
High steam header radiation or High SS blowdown radiation SI Termination:-
F-CS Pressure:
1100 psig and i ncreasi ng Pz r L ev el
> 25 %.
AFW Termination SE NR level on ruptured SG 226 SG NR level on unaffected SG 50%
Charging Termination' Realign charging for cold l.eg injection (600 gpm flow)
SUbsequent Action Once steam header (and RCS) presSure below 350 psig, realign charging to normal injection lineup from VCT and reUMEm letdon.
Initiate RHR system
2.1.
aTrml H
ldraui I Lyi 7.. ~
3 1
']~g~
_E Lnt SiC23 H~~e
~
Plant spectic ana.1ys.es were made on the overcooling transients due to he los of secondary coolant Re.
1).
The cases analyed included MSLB at hot full power (HFP) and hot zero power (HZP) i ntermediate steamline break at HFP (0. 1 ft
^2/loop) and the credible steamline break at HZF (0.02 ft
^2/loop).
As study indicates (Ref.
2),
the downcomer ilUid temperature is the
.most sensitive parameter for the PTS analysis.
Low temperatures wi. 1 Cause a PTS concern even without system repressurization.
By studying the. SONGS 1
results (Ref.
1),
one realizes that among the cases analyzed, the most limiting one is theMSLB at HZP.
The cold leg temperature drops to about 220 F.
If incomplete mixing of loop flow and SI flow in the downcomer is allowed Ksuch as when natural ci rcu Lati on.is lost),
the dcwncomer temperature can be lower than 220 F.
This is below the RT NDT presented by the Westinghouse owners group
('O).
Depending on the probability of the occurrence of MSLB at HZP, the cooldown in this case could cause a potential concern.
in the sequence of events for the transient, the SI was terminated by the operator at 717 seconds into the transient.
This is a relatively shorr ime allow-.nce
-or operator action.
No discussions were made on 17
the enstivty o
sytempresuretemeratre espnseto the time allowamce for operator's termination of-, the S:!Ilw.
In hSouthern California Edison (SCE) mentioned T!at the -case uFd-F r
PTS an a Iyzsi E, "employed the conservative assumptions normally associated with design basis safety analysis".
The asmtions uSEd +or the anavsis of core heatup and cooling in an FSA.R USUly.i 1V not conservative for PTS analysis (e.g.,
decay heat level and SI tempEratUre -and 'lowrate).
The analyses provided in WCAP-10019 are typical of FSAR-type design basis events.
However5 the boundary conditions have been selected to enhance the overcooling.
Maximum safety injection and feedwater flows are aELSUmec q d,
miImumL.
water temperatures are used5 and heat sOLrces are either omitted or are conservatively underestimated.
Large and small LOCAs have been.addressed. as well as large and small steam line breaks.
In additin,.
the Rancho Seco o-rercooling event was included.
Westinghouse indicca tes that the dynami cs of this event would be si imqi l ar to l
_ow probability small steam line break (including addi ti onal f ai I ures)
Operator action is identified for two events
'resented in WCAP-10019.
For the isolatable LOCA StUCkI:
open POR )
it is assuimed that the operator isolated the break in TO minutes.
For the lIarge steam line
- break, it is assumed that
.ilio lE:..
A 0
the faulted SteaM generator and makeup injection fIow-r to the RCS iE terin te within
- 10) min.:ta
- 2.
riter
-for Procedure Revew Al ScOG i
emercency operating procedures, called Emergency InStructionS (Eis),
were reviewed.
These -included spu ious safety injection, steam generator tube rupture, loss of secondary coolant, and loss of reactor coolant accidents.
At the time of the
- audit, these instructions had not yet been approved by the On-Site Review Committee.
Plant
- personnel anticipated acceptance before -
plant start-up w.th minifmal changes.
The audit criteria for the content of procedures was somewhat flexible Jo rcount for lack of operator knowledge about the new procedures and to identify which procedures must be used to respond to a given.
'Tr addition, detailed.operator knowledge of actions for preventing or itigating PTS could offset some.
weaknesses in procedures.
With this in
- mind, the following criteria.
were established for the procedures audit:
Fr oCedures should not instruct operators to take actions that woul d violate NDT limits.
(2F ProcedureS should provide guidance on recovering from 19
transient or accident conditions without violating NDT or saturation limitS.
(7)
ProcedUres should provide guidance on recoveri ng from PTS conditions.
(4)
PTS procedura guidance should have a
supporting technical basis.
(5)
Safety injection and charging system operating instructions should reflect a consideration for PTS.
(6 Feedwater and/or aU.-i1iary feedwater operating instructions should reflect PTS concerns.
(7)
An NDT curve and sEtura.tion curve should be provided in the control room.
(Appendix 6 limits for cooldowns not exc eding 100 F/hr).
2.3 Citeria For Trdianing Proaram Review The audit team used the criteria developed by the Etif4:F as a. standard for 11l plant PTS audits.
These criteria cover three general areas.
20
- 1)
Tr aini ng should inc ude speci fi c instruction on NDT vessel 1mits f or norrmal modes o+
cperation.
Trining should include specific instruction on NDT vessel limits for transients and accidents.
Training should particularl emphasize those events known to require operator response to mitigate PTS.
Within. the general criteria, more specic criteria were used in r Ev I &iewng detailed training material such as lesson plans and in preparation for interviews with the training staff and operating personnel.
(1)
Training in NDT limi.ts should include the knowledge that irrad-iation adversely Effects fracture toughness properties of the reactor vessel.
Operators should know that the vessel and welds will lose ductile material properties and trend toward embrittlement.
C. e r a-t ors ahou.d be aare' that NR.C has =ent letters to
21
- perators should understand that a rapid reduction in rector vessel temperature can raise the possiollity of crack propgation.
particularly if pressure ris.after thne temper ture reaches its lowest value 4
Operators should be aware of the types of events which are known to involve PTS (such as steam line breaks and secondary side malfunctions).
- 5)
Operators should appreciate that other safety limits (such as core cooling and shutdown margin) must also be balanced with the PTS limits.
(6)
Training should emphasize the instrumentation
.available t'
observe key parameters as they. approach limits.
Strategi es/options -*which are Under operator control should be emphasized.
'7)
Oerators shoul d understand the basis
-for current emphasis on FTS.
specifically.
that more severe transients have occurred than expected (Rancho Seco, Crs ta1 Ri.'er)
Freparation for rei ew of the training program included a
review o
SCE orrespondence with the Commission, including a re-port on vessel.
integrity of Westinghouse operating plants (WCAP-I0019),
emergency PrcedureS furnished by
- SCE, and technical specifications.
An i.ntervie p'n was developed which used the general training criter i a and the specific subjects that were included in the SCE training material.
3DESORIPTIONJ OF AUDT Pr ior to the plant visitP PNL reviewed the SONGS 1 150 day response (letter from K.P.
Baskin (SCE) to D.G.
Eisenhut (NRC) dated 1-25-82)
And a more recent package supplementing that response (letter from P.
Bask in (SCE) to D.M.
Critchfield (NRC dated-5-20-82).-
The procedures and training audit are described below.
7e-.
I.
Procedures
.During the plant visit, the audit team reviewed the current version of a rious SONGS 1 Els.
As mentioned earlier, these procedures had not yet been approved.
Instructions relating to possible PTS events were d-iscused with the group of individuals responsible for -writing the El s.
The baes for PTS related instructions were di scussed in the course of.
w 4orking through various transient and accident scenarios.
27
7he uait team then visited the control room to review instrumentation
&n vilable pre sure/temper ture curves tht had been referred to in heprceou-res.
The
+olo Ing Els ere i ncluded in the audit PROcCEDU'RE 0N0S -
U1-E Reactor Trip or SI S01-1.2-1.0 SI Termination Following Spurious Safety Injection 501-1.2-1. 0.
-Lss o+
Reactor Coolaknt.0-1.2 SI Ter-minatipn Following Loss of Reactor Coolant S
50-1.2-1.11 Post-LOCA Cooldown and De7pressur i Zati4on 6,01-1.2-1.12 Loss of Secondary Coolant S01-1.2-1.2 1
Terrintion Following Loss of Secondary Colant SI01-1.2-1.21 24
-Stec,-
G~enerEator Tube Rupture 80 1-1.2-1.::
Response
t lcs
'of Secondary Hezt Sijnk 501-1.2-15
.2 Tranina The audit of-the SONGbS 1 training program began with a review of the PTS trainina outline and discussions with members of the training staff+
PTS questions used in an oral exam and on the requalification exam were reviewed.
Interviews were then conducted with the foll owing licensed operations personnel:
1 Shift Supervisor (SRO) 1 Control Room" Operator (RO)
I Asistant Control Room Operator (RC))
1 Shift.
Technical Advisor (STA) 1 Train.ring Rpcrvisor (5RO)
K IdINH FROM THE SOGt-!S 1
LD IT 4,1 C=rin tr Podr Wi th the AUdit Criteri 4.1. 1 Procedure Should Not InStcruct OeratorcS to Take Actions That ouLtLd Vio1ate NDT Limits The procedures th at were audited did not appear to contain inStruCt1ions that w.ould cause an operator to violate NDT limits.
The procedures referred to or included caL'tions to observe the limits o+
the NDT curve (Appendix,
100 F/hr) and the technical specification for hee.tup and coo1down limits.
4.21.2 ProcedureS Should Frovifde Guidance on Recov-rihn From Transient
- . Or Accident Conditions Without Violatina NDT or S atrti90 Li mi ts Qne El (Steam Generator Tube RuputUre) references the NDT technical specitic.tion curve mentioned above and other El's instruct the oper=t or to mai ntain specified subcoolIng margins (40 F, 50 F).
Two
<=s.'accli.ng meters indicate subcoo.ing margin.
4..
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- 1.
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Feed wter (FW) and/or Auxiliry_ Feeater (gFW) 2 rE tnO Instructions Should Reflect PTS rncerns MFWFs are realigned to function as SI pumps in the primary system when SI is required.
Instructions to throttle AFW pumps and maintain SG level minimiums reflect PTS concerns.
4.1,. 7 A.n NDT Curve and c Saturation Curve Should be Frovidd in the Control Ro~om An. NDT curve was available in the control room.
The saturation curve was misino but will be in place for start-up.
Subcooling curves oili 7lso be included.
42 Cmori-;
n oft Trai. n-. ng w,4i th Audi t Cri teria ran 1
i24ng hbuTa1dIlude Secific Instructi on on NDT'Ve&e
imi ts fo-~r Normal~ Modes~ of OFratio cn ction 2 of the Operator Requalification Pressurized Thermal Shoc:
course outline includes a
discussion of the PTS issue in
-jeneral and NDT vessel limits.
All interviewees showed good urerstrnding in this ar e a 4.2.2 Trainina Should Include Sqggkiic Instructions on NDT Vessel Limits for Maior Transients and Accidents Section 3 of the requalification training deals with PTS concerns for major transients and accidents.
All interviewees were aware of the Westinghouse.
analyses of various events for SONGS
- 1.
They all knew the ve-sel RT NDT and understood NDT cooldown limits.
4.2.3 Trini Should Particularly Emohasize Those Events Known to RE2tBie _O~erctor Re--------toMitignate PTS Section 4 -of requalification training -concerns operator actions.
Ope-tOrs are taught that if they fol low procedures they will not reach a PTS condition.
It is emphasized that SI termination criteria in thr:Is should be closely observed when PTS is concerned.
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