ML13317B116
| ML13317B116 | |
| Person / Time | |
|---|---|
| Site: | Robinson, San Onofre, 05000000 |
| Issue date: | 04/15/1982 |
| From: | Mazetis G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13317B112 | List: |
| References | |
| TASK-A-49, TASK-OR TAC-47549, NUDOCS 8204300459 | |
| Download: ML13317B116 (69) | |
Text
i:.
-7
-7S
'PRESURIED-TRA-IGOR-Task Force Chairman:
G. Mazetis 44 04/15182 ROBINSON SER INPUT-:TC
NCStffAdit of Robinson 2 Pr6cedure.
-. and 1T1,ainn fo Pr1 e
ss urized Thermal Shock CONTENTS INTRODUCTION.....
1-1 1.1 Short-Term Effort Objectives and Scope of w
1-2 1.2 Current Status of Generic PTS Issue.....
1.3 Robinson 2 Configuration.................
2 SHORT-TERM CRITERIA USED FOR ROBINSON 2 AUDIT..............21 2.1 Transient and Accident Ana2yses......*******
.2-6 2.2 Criteria for Procedural Reviews.......**.
2-7 2.3 In-Plant Training Program......-
- 3 3
K FINDINGS FROM THE ROBINSON 2 AUDIT.......
1.........
3.1 Transient and Accident Analyses............
3-3 3.2 Procedures.......****
3.3 Training.........................................
o...
o.................. 3- 0
-3.4-Summary.......****
4-1....
4 FRACTURE MECHANICS........4-1......
4-1 4.1 General.....*..***
................................. 41 4.2 Robinson 2 Fracture Mechanics.....
5...
-1 5
RECOMMENDATIONS.......-.**
1......
5-1 6
APPLICABILITY TO REMAINING SEVEN PWRS...... *.............
0.
ROBINSON SER INPUT 04/15/82
a 1-INTRODUCTION and Scope of Review
.O.
Task Force was established to evaluate for. Robinsofl.
hcertain aspects ofthe Pressur ied Thermal Shock(PTS) issue r
n The question that the Robinson Task Force focused on was ARE CORRECTIVE ACTIONS REQUIRED THAT MUST BE INITIATED BEFORE THE E LONERTER PS POGAMPROVIDES GENERIC RESOLUTION AND ACCEPTANCE CRITERIA?
Emergency procedures and operator training were the only Robinson Task Force applied the abov6 general question.
As noted n the NRR March 9, 1982 presentation to the Commission:
to Teiy that existing operating.
...we will undertake a program tc sa ftprentngortigt
_-.procedures contain the seps necess~aYoprvnad/rmtgeP ents, and to verify tht operator education/training programs regarding PTS are acceptably thorough.'
ith CP&L the week of March 15th and, during a
tonference call on Marchw 9ta, th details of our expcted reviewareas were discussed.
Also discussed was a planned visit to the site.
ith-the limitation of a 30-day response, the scopecof reviewhaducobe narrowed
. so that meaningful conclusions and recommendations could be produced.
Therefore, resolution to the varied technical questions on PTS (thermalFydraulic analyses, fratur mchaic, pobailtie) as not part of the Task Force charter. Also,,
fracture mechanics, probabilities) wa no ~ecio
)i ujc o coordination implementation of any recommendations (see Section 5) is subject t.
and consistency with the longer term generic program (USI A-49).
ROBINSON.SER INPUT SEC 1 04/14/82
A visit to the.Robinson 2 site took place on April 5-7, 1982, during which time the Task Group evaluated procedures and training. The key findings of the group are discussed in Section 3. In preparation for the Robinson 2 evaluation, the Task Force used the general criteria addressed.in Section 2.
1.2 Current Status of the Generic PTS Issue Efforts to pursue an integrated PTS program involving a variety of technical areas are continuing under USI A-49.
The sumner of 1983 is the current schedule for finalizing our generic regulatory requirements for PTS along with required corrective actions if the generic requirements are not met. Key issues are yet to be resolved and extensive.programs exist to provide the foundation for the generic regulatory requirements.
Before the above effort resulting in regulatory requirements is completed, however, we have committed to the.Commission to have developed an interim initial position for the summer of 1982 (June).
The interim initial position will consist of NRC evaluation of the safety of continued plant operation (and initial corrective actions required) for the eight plants previously identified as representative of plants having the highest RTNDT.
Technical assistance is being provided by a PNL muti-disciplinary team.
PNL has been contacted to work with the staff to provide recommendations regarding the June 1982 initial position on the safety of continued operation and to recommend any additional corrective actions that PNL believes should be initiated before the NRC generic resolution and acceptance criteria are adopted.
The June recommendations by the NRC staff to the Commission will also consider.the findings and recommendations addressed in Sections 3 and 5 of this report, as well as other Task Forces formed for related investigations (such as fluence reduction at the vessel wall).
1.3 Robinson 2 Configuration Robinson 2 is a three-loop Westinghouse PWR rated at 2200 MWt (700 MWe).
Normal pressurizer level is contr olled b ce the chemical and volume control system which contains three positive displacement pumps. The safety injection system (SI) utilizes three high head pumps which will initially discharge the boron injection tank (BIT) into the cold legs of the reactor coolant system.
04/14/82 1-2 ROBINSON SER INPUT SEC 1
The SI pumps have a shut-off head of 1500 psig and have a rating of 375 gpm at*
1080 psig.
The SI system also* contains three accumulators which discharge at 600 psig and two low head pumps (RHR) rated at 3000 gpm at 115 psig.
The Robinson 2 control room utilizes an L-shaped bench board which contains standard three-loop information..However, most_of the meters and controls are considerably smaller than other three-or four-loop control boards...The most important plant parameters are displayed on recorders whichave -anormal speed of-1.ih/hr.
The following table contains the. major parameters available to an operator at Robinson 2 which would assist in monitoring PTS events:
Parameters Display RCS pressure Wide-and narrow-range meters (1) wide-and narrow-range recorder RCS temperature Wide-range THOT and TCo0 meters and recorders for all three loops Rx head thermocouple (1) meter - will be functional after refueling Core exit thermocouples Normal () display using toggle switches Single T/C display Subcooling monitor Utilies "pressure to saturation" meter May select core T/C or wide-range RTD Push-button selection may display oF subcooledin a digital form May select Loop AT.
04/14/82.
1-3 ROBINSON SER INPUT SEC 1
................................. Z..~.
2 SHORT-TERM CRITERIA USED FOR ROBINSON AUDIT
- 2. Tansient and AccidentAnal~ses~
2,...
-IntroduCti~on of~seamline breaks. (exces!
events in PWRs may occur as a result of s
fol t accidents.
Overcool i ng ee-fol sive steam flow), feedwater system malfunctionst or loss evere accdeng Multiple failures and/or operator errors can et ini represuroatin events.
Of particular concern are those events in.which. reprehietion the primary system occurs following the severeovercoolicg his steothe addresses an overview of Robinson 2 overcooling events which occurred sjnce the plant was built. Aside from the primary mission of the Task Force to audit procedures and training, also provided (Section 2.1.4) is asu ary.of the thermal-hydraulic analyses av'ilable for evaluating pressurized thermal shock events.
Section 3.1 provides our comments and conclusions on these events and analyses.
Seto 2.1. 2rvdso 2.1;2 H.
Robinson Overcoo Events 2.1.2.1 Sem afety Valve LineBreak, Ap rl 28-1970.
On April 28, 1970, during hot functional testing (no fuel loaded), one of the steam generator safety valv'e connections fai-led due-to overloading.
A 360*
rcotial break allowedcthe safety valve to blow off the main steam line.
The plant conditions were:
5330F, 2225 psi primary 900 psi secondary 3 RCPs running 45 gpm charging/letdowe n
- no feedwater to the steam~generators As~~~~~~L ateuthftee-n shdl As a result of the-6-in. schedule 80 pipe break, and with no decay heat, the to a 320aF cold leg temperature. The operatc plant cooled down 213 RF ink -OS hourI 2-1 ROBINSON SER INPUT SEC2 04/14/82.
.T7
immediately pped the RCPs (30 seconds) and started the remainuing two coolant charn pds)
The minimum primary system Pressu withthe safety injectioc (SI)setpint attio occurred.
The p s recovered to a normal no-load condition of 2050 psig b1s heror--to, shutdowni.
- 1.
and charging/letdown reestabl ishe p
A, post-event review of the data indicated that the pressurilzer.ur line I not empty.
-Abase caeanalysis was performed for the event. oInaddi ti rin a ensitivity analysis was performed without RCP trip with hnlyone charging p and with a primary heat source. The analysis-showed that the pressurizer would drain and the primary system pressure would falp l belW the SIsetpoint in about 3 minutes. The cooldown was less and the pressures were lower than th e ase ca naysis. It is expected that the operator actions, based on current basoese woldbes milar to this sensitivity analysis. The safety valve stand-off pipingdwas redesigned to prevent any similar occurrences.
-2.1.2.2 Reactor Coolant Pump (RCP)Seal Failure
- Event, May 1, 1975 During full-power operation, RCP "C" seal.1 leakage exceeded the echnical Durpn fu limit of6 gpm.R A load reduction was..commenced at a rate of 10%
p mnute to 36% power and pump "C" was deenergized.
Reactor trip occurred der mtn trbine tp pwes ing from the load reduction.
The decision was made due -to a turbine trip resultngfo dntbreoedopus Aad to restart pump "C". when seal injection could not be restored to pumps eand B."
Shortly after restarting the pump, whil e l be g to decrea se and 3 failed on pump "C" and the pressurizer levelbegan to decrease.
The following chronology is provided:
2300 -
RC system at 1700 psig, 4800F RCP "C" running 0015 -
Stop RCP-"C," on high standpi pe level alarm Pressurizer level falling rapidly due to seal 2 and seal 3 failure
.2-2 ROBINSON SER-INPuT SEC 2 ftA/T&/;t--9
Ve 0016 - SI pump "A' manually started to supplement charging flow (injection
-toho.lg 0018 tsi pumps B9and "C" manually started, pressurizer level stops falling 00 an p to auxiliary pressurizer spray to
-r rd c prssure (1150 fsigom this timc colt.t erature below A000F) i 0C11 due to rising pressurizer eve
.039 -Stop-S P umP
.p g at -hi 048 S
acumulators partia time) f om 5OF to approxi etly 310 F inn-half' The cooldown for this event was from O to a o t lYr th e hou,.
iththepresure decreasing from 1700 Psig 9to abut 15 pi ve h
period of interest. The use of the auxiliary pressurizer spray rapi e e the pressure to 500 psig.
ed SI to stabiize pressurizer level and pressure while using the ma acondenser to cool down the plant for RHR entry.
There is.
no indication that SI was used to repressurize the plant.
2.191..2.-3 2.1.2.3 Stuck e
Steam Generator Relief.Vale
- vent' oebr 17 2.
- t. amc 36 t a hile at nominal full-power operating conditions, the operator was using steam Whelerato ia rifllver~t ovieSeperature control.'
One valve would gnteaore reeultif g i the equivalent of a small steam line break.
The nont.reclose result ed in aereactor trip and safety injection. The secondary side lowdown res 1570F over a 2-hour period, to 3890F, during the S
oue f the evn Insufficient information is currently available to addressoerato actions taken during this event.
2-3.
ROBINSON SER INPUT SEC 2 04/14/82
2.1.3 H. B. Robinson Termination Criteria 2.1.3.1 Reactor Coolant Pumps (RCPs)
The RCPs-are tripped when the primary system pressure falls to 1300 psig. in addition, the RCPs are tripped if seal cooling is lost, if excesive seal leakage occurs,.or if excessive -vibration occurs 2.1.3.2 Auxiliary Teedwater*
Auxiliary feedwater is isolated to the steam generator identified as faulted for steam line breaks or steam generator tube rupture. -The flow rate is limited to 400 gpm to any steam generator.
2.1.3.3 SI Termination During LOCA The termination criteria for safety injection during a LOCA addresses core cooling. No reference to pressurized thermal shock is provided. The termina
. tion criteria include a 2000 psig (and increasing) requirement.
2.13.4 SI Termination During Steam Line Break The termination criteria for safety injection during a steam line break are:
One RCS THOT less than 460*F, RCS pressure greater than 700 psig (stable or increasing),
Pressurizer level greater than 20% (heaters covered),
RCS subcooling greater than 40aF, and Heat sink available (U-tubes covered).
As shown, one of the criteria for terminating SI during a steam line break is one wide-range THOT reading less than 460sF, with wide-range primary coolant 04/14/82 2-4 ROBINSON SER INPUT SEC 2 04/14/82<r~-.t'-.
~
system Pressure greater than 700 psig and stable or increasing.
Th Westing house guideline value is 350aF, THOT This value includes all uncertainties and does imply reference to the downcomer temperature.
.Theuncertai nties inld e core heatup during natural circulation, ECC Mixin and instrument errors.
Westinghouse has reviewed their fracturen-data for
- rniidentrange of taand, for the most limiting vessel at end of life, they conclude that the 350OF Ti. would not result in vessel. failure. The 700 sg cocldethtth 3
HT ide LOCA does- -ot exist stable or increasing, pressure assuraes that a primary side the 350eF coincident with the steam line break..
Robinson 2 has inreased the 350*F value to 460OF to provide a combined assurance that 4QO* 5Lubcooling exists at a p
ressure of 700 psig,- con'cur'rent with a s uf f icientl y high.tempelrature to accommodate brittle fracture concerns.
Also, it is noted that the Westinghouse 350oF/700 psig values would violate the Robinson 2 NDT limit for 100oF/hr cooldown events.
a 21.4 The 2.1.4.1 FSAR Analyses
.FSAR analyses assumptions are developed to demonstrate compliance with current NRC regulations concerning fuel.:design limits, pressure boundaryprotection (overpressure protection), and radiological releases. These assumptions do nol necessarily result in the most severe overcooling. The analyses are typically carried out for only a few minutes and do not provide enough data to perform vessel integrity fracture analyses.
2.1.4.2 WCAP-1001 9 Vessel Integrity Anal yses The analyses provided in WCAP-10019 are typical of FSAR-type design bases.
events.
However, the boundary conditions have been selected to enhance the vercooli weg.
Maximum safety injection and feedwater flows are assumed, minim water temperatures are used, and heat sources are either omitted or are conse atively underestimated Large and small LOCAs have been addressed, as well a large and small steam line breaks.
In addition, the Ranch Seco overcooling 04/14/82 2-5 ROBINSON SER INPUT SE
event was included. Westinghouse indicates that the dynamics of this eyent would be similar to a low probability smallsemln bek(nldigad-tionalfailures).
Oper ator action is identified for two eventspresented in;'
WCAP-10 019.* For the. -is olatable LOCA (stcopenIRVX. 'it i*asedtt CAP-0019 Forutes Forthe large steam-line-break, th poperator isolated the break in 30 mi nutes..For
- e.
oh aulted stem generator-and it isasumd that auxiliary. feedwater to th auld semgnrTran makeup Injection i is terminated within 10 i nute.
2.1.4.3 Westing us Procedural Guideline Analyses-In response to Item I.C.1 of the TMI Action Plan, Westinghouse has performed a series of "best-estimate" analyses-to support their current program-for operato uidelines and procedure deelpment. These analyses indicate-that considerabl conservatism exists in the WCAP-10019 vessel integrity analyses.
2.1.4.4 NRC Independent Audit Analyses Independen tea line break have been performed by Indeendet audit analyses of a large sThs anlssaeinareetw LANL with the TRAC-PD2 computer programs. These analyses are in agreement wit the Westinghouse guideline analyses..
Independent audit analyses are also being performed at INEL with the:RELAP5 computer program for small steam line breaks. The results of'these analyses.
will be available at the.end of April 1982.
2.2 Cieria for Procedural Reviews The procedures to be reviewed were selected based on the perceived likeli hooC of conditions occurring that might subject the reactor vessel to pressurized thermal shock conditions and based on the potential consequences of less lik transients. Such procedures selected included normal heatup and cooldown, tamsgenerator tube rupture, steam line breaks, and loss of coolant steam geeaotuerpre accidents.
The audit criteria for the content of procedures was somewhat flexible to account for the operator knowledge interface and to identify which.proceduri must be used to respond to a certain transient. In addition, detailedoper was-b soehaseeilet 2-6.
ROBINSON SER INPUT S 04/14/82
knoesdge of acdions for preventing or mitigatio oTg cl toffset s wak nessS i proedues. Wihthis in mind, the followingcrtia eresblhd for the procedures audit:
- 1)
Procedures should not-instruct operators to take actions that.Would 44iolate NOT imits; ur s
uld provide guidance on recovering from transient or accident conditions without violating NOT or saturation limits.
(3) Procedures should provide guidance on recovering from PTS conditions.
- 4) PTS procedural guidance should have a supporting technical basis.
(5) High pressure injection and charging system operating instructions should reflect a consideration for PTS.
(6) -Feedwater and/or auxiliary feedwater operating instructions should reflect PTS concerns.
(7) An NDT curve and saturation curve should be provided in the Eontrol room.
- {Appendix G limits for cooldowns not exceeding 100*F/hr).
2.3 In-Plant Training Program The effort of the task force to determi ne the effectiveness of CP&L training in
- TSbga y eelpigtriin riteria which would be used in evaluating the.
training material, onterviewing Robinson 2 shift personnel, and assessing the evaluation CP&L made after completion of the training.
The criteria developed into three general areas:
(1)
Training should incldde specific instruction on NDT vessel limits for NORMAL modes of operation.
(2) Training should include specific instruction on NOT vessel limits for transients and accidents.
04/14/82 2-7 ROBINSON SER INPUT SEC 2 04/14/82..
(3) Training should rticularly emphasize those eves known to require operator response to mitigate PTS.
More specific criteria were also developed to aid in the review of the training progm-and in parato f-intervews with operatipg personnel.
teq to furnish an outline of the Training progra n'P an VCP&
wat requested the lesson planwhichwasusd i s
the ting cIases. 'They ere also ques t iond o themeth used t6 e aluate the effetiveness of the training sessions.
Preparation for review of the training program included a review of CP&L correspondence with the Commission, including a report on vessel integrity of
.Westinghouse ope rting plants (WCAP-10019), normal and emergency procedures furnished by Robinson 2, the Robinson 2 license, technical specifica tions, and the FSAR. An interview plan was developed which used the general training criteria and the specific subjects which were included in the CP&L training material; Each interview was preceded by a discussion of the reason for the audit, acknowledgement that the individual could use allr material available -in the control room, particularly the -followup or recovery steps in the emergency proceduries, and a request that the individual not inform other operators of.the questions asked-in-th interview-;Several interview aids were prepared to provide the-operators a point of reference for discussion and to allow them to predict responses or execute recovery strategies to mitigate PTS or challenges to other limits.
04/14/82 2-8 ROBINSON SER INPUT SEC 2
3 KEY FINDINGS FROM THE ROBINSON AUDIT 3.1 Transient and Accident Analyses 3.1.1.
Introduction:
This section presents our comments and conclusions based on the aterial provided in Secti-on 2.1 of this report.'
3.1.2-Robinson 2 Overcooling Events CP&L reviewed the Robinson 2 operating history and presented three events where the cooldown rate exceeded 100OF per hour. The minimum cold leg temperature measured was approximatelY 310*F during the cooldown for the reactor coolant
- pump seal failure event of May 1, 1975.
In each case reviewed where --operator.
data was available, the operator actions were different than would be expected with current plant emergency procedures.
For-example, for steam line break events, the cooldown transients would-be less severe using the current reactor coolant pump trip criteria (continue to run until 1300 psig). Insufficient current procedural guidance exists to evaluate whether the operator would continue to run additional charging pumps during the small steam line break-for an extended'period.-. For~a given overcooling event, particularly if the pressurizer does not empty, continued use of-additional charging pumps could result in rapid repressurization.
For small-break LOCAs, repressurization to 2000 psig may not be a evisable following a severe overcooling event. CP&L and Westinghouse believe that repressurization to 2000 p.sig will not compromise vessel integrity.
04/09/82 3-1 ROBINSON SER INPUT SEC 3
3.1.3 Robinson 2 Teminatio riteri 3.1.3.1 rmnation 1CA
.".f t::13 c,o~urnga LOCA--are:
The terminrati on cri teriafr aeyjjcilldrn OAae a 20 sig.and r ocreasing, RCS pressure greW e2 PressuiZer eve a oevand respond1l Pg Heat sik av aiable (U-tubes covere an RCS subc oled at ieasto40F.
These criteria are weighted to core cooling concer ansddo n
e dtht Thddess theresurze thermal shock ssue.
The licensee has iniae-ht aedre the press ue tralsoes under review by the staff, no PTS concerns based on the Westinghouse analys.
exist during a LOCA.
LOCA is that the primary On ftecriteria for termination of' SI during a LOC Tis that te pridearYh One of the r ere 2000psig and increasing. This value provides the coolanit system press~r
.s followi-ng information:
The break has been isolated or the Si flow is equal to or grea the break flow.
(sts to termina te SI before the PORV would be challenged.
(2) Some margin ex)i 40aF subcooling margins (3) Repressurization to 2000 psig further assures a m
jncluding uncertainties.
d Robinson di d not have the At the time the emergency procedure was developede Roin n issue. To subcooling meter instaled, and core cong wa nstrument readingsua To verify sbcooling, and include unc ertainties i n insrmn rdit anoflo itary system pressure Pf 2000 psig was adopted.
that the Robinson high head safety injection pump cut-off head is 1500 psia.)
.Ros~t4so I
SER leUT SEC 3 3-2 R
04/09/82.
.S Termination Critea During Steam Line Breaks Teo termination criteria for safetyinjection during a steam line break, as present nStitoned add the prtssurized thermal o k issie by a,
.cang t the tA r i ntieria e s sd oin ue ay sion es3.1.3.1.
c n fit i eden t hear ros a nwdA ar e r re se nttve e seonc e d t n tCAP -10019 ar re a o n e a s a e be t e or-..
d niet e
a n le a__r rh e
e rass es rae assmptios e
in r
chse beStie it nalye ndct ta h
we~~~~~-
u A-a1ac~.
wen W C A
-1 0 1 c a uatot 6e s t h a 3 5 F f o t h s t a l n b r a sp c r u m.
M l c o o l d oin g a n d no c o nw i t h r e a r t
maks 3.-1.4 ThermlaVlydratilkc Analyses s de tinaes a esasuited to the evaluation of vesselintegrity.
- Insuficient carrygut, in tim eittopromfaure analyses. Thi evenlt presented in WCAP-10019 are boundingovercooling events, and are representative Tof.design bases events (single failure).
These analyses are suitable s
for vessel integrity studies. Analysest berformed by Westinghouse, usi estimate" assumptions, indiate that considerable conservatism existsin the WCAP-10 0l9 calculations.
These best-estimate analyses indicate that the cooldown would not be less than 350F for the steam line break spectrum.
While 2 some uncertainties exist with regard to mixing for small-break LOCAs,. these l o s s o f RPS i n v e n t o r y e v e n tda e s, a n a n b
y O a di t n e d
p sectrum.
T he*NRC independent audit thermal-hydraulic calculations for the large steam line break addressed in -Section 2i.1.4.4 support the above observation on: the Westinghouse analyses.
'Additional audit cal cul ati ons to be performed during.
April are expected to provide further confirmation of the estinghoue therma hydraul ic.
analyses.
3.2 Procedures 3.2.0 Our audit included a review of procedures selected as discussed in Section 2.2 discussions with licensee and Westinghouse representatives on the instructions relating to PTS and their bases, and an audit of the control room copy of the procedures to determine their legibility and currency.* Our audit included the 04/09/82
.3-3 ROBINSON SER INPUT SEC
e
- 9.
following Emergency Instructions (Els), Abnormal Procedures (APs), and General Procedures (GPs):
EI-1 Incident Involving Reactor Coolant System Depressurization EI-6 Loss of Feedwater EI-7 Station Blackout Operation EI-14 Reactor Trip.(Part-A) Turbine and Generator Trip Pr AP-19 Malfunction of RCS -Pressure Control Syste AP-24 Loss of Instrument Bus AP Spurious' Saf eguards-, Actuati-on:
G-2---Heatup--(C 61d -Sol -id -tt-Subcri tica a otadTAVG)
GP~B ReactorTi eoey GP-5 :Shutdown (Normal. Plant Shutdown From PowerOpertions to Hot Shutdown Conditions)
GP-5A Pl ant Temperatue an rssure-Control1 UsingNaulCicato GP6Cooldowl-(Plaflt-Cooldown From-Hot,Shutdown-to Cold ShutdownConditions) 3.2.2 Comparison of Procedures With the Audit Criteria (1) Procedures shouldrot instruct operators.to take actions that would violate NDT limits. The procedures audited generally did not appear to contain instructions which would cause an operator to violateNDTlimits; most of the procedures referred to, or included cautions to stay within, the limits of -the NDT curves.
These curves.,are consistent with the technical specification heatup and cooldown limits.
The only area where the procedural -instructions may violate these limits (even tbough cautions exist). is' the safety injection termination criteria and charging' pump operating instructions -in the loss-of-coolant accident procedures.
The-termination criteria require RCS pressure greater than 2000 psig and increasing.prior to.terminating high head safety injection (shutoff head approximately 1500 psig). There are no explicit instructions for pressure control or operation of the charging pumps until a controlled cooldown depressurization is begun using GP-6. Discussions with Westinghouse representatives indicated that the SI termination criteria are under review as part of the generic procedural guideline development and it is anticipated that they will be changed to a lower pressure, at least for the plants having intermediate head SI pumps like Robinson 2.
04/09/82 3-4 ROBINSON SER. INPUT SEC 3
(2 pocdues shll prvie uidance on r~e il froM transient. or 2e s ee item (1) above for discussion Olimitsocure fr.dproesu zation events (EI1) refers the ope raorcooCivg. If proides pup instructionto a~ita at least 4atura circlion itractr thean pumpto are tripped, the procedurefor naT e rcatto n edu rc tepo to mainltain! at least 5 0pFubcooe 3 5Tsa presurestempeott poie-.
maximum 5 bcooling limits 40rF s ubc oe d ure. f er re covryi howing a saturation curve and oa nt rupturedinsr uct The roverto t instructions for a secondary coo dtute genstuteoerator toz establish steam dump from the "gooresr steaear to sncrase oz t
otdet stea rator.
(3) prcdrssolbrvd ialeonrc Whie te rocdues roideintuc.ion for itiin h
Swti conditions allowed by the NDT curvests acnot mapparetthepo edures r escogni e tha t t some trani ts or cc idns may oe st a nP t
o o dito n s t oa e o t e
ra t c n b to er faulted steam generator an r
steam line break.
fo on mainana (4) d oeua iac hud ae in uTepoceua udnei eeal cu on
.-i nt r it th perdd t he
~~5~i heOwesGr procedure udlnsTeegie liestarehbsed onestGruemaergencalyses of transients.
The actions swith he bund and alysespresentedin acP11S atghouse rePr e senatith stad thgase gieeine are also being reviewed against besaitaiate acture mchncanlssadtttisefrwllb cobee tedtmay 192.Se2Sctos an d 3.1 f ort a discuonsionofthe gsat ction termination criter.
AR '0/82 ROBINSON SER INPUT SEC 3 uctio f
3-5 04/09/82 tamgeeatrst sa
(5) High pressure injection and charging system operating instructions'should
. reflect a consideration for PTS. The 700 psig SI termination criteria for steam line breaks reflect PTS concerns. The SI termination criteria for
- loss-of-coolant accidents -would allow repressuirization to above2000 psig with-a cool vessel.-There are no-specifi instructions f.or operation of he charging pumps following the depressurization transients.
- 6)
Feedwate Fw ano auilaryfeedwater (AFW) operating nstructions should reflet PTS onerns.
Instructions are provided in.the-steam generator tube-rupture and the loss-of-coolant accident procedures to terminate-FW/AFW flow to the faulted steam generator These and other procedures provide instructions to maintain steam generator levels in the good steam generators within a defined band.
(7)
An NDT curve and a saturation curve should be provided in the control room..
These curves are provided in the Curve Book located in the control room and are referenced in the applicable procedures.
Each of these curves is on a pressure-temperature plot. Curves 3.3 and 3.4 show the technical specification heatup and cooldown limits. Curye 3.5 shows the saturation curve and a 40*F subcooled curve.
The control room copy of the procedures and curves that we audited was iegible and current.
3.3 Trainng S3'3.1 Introduction The site audit of CP&L's PTS training program consisted of a review of the lesson plan used for classroom training and personnel interviews with five Senior Operators (two of these SOs were Shift Foremen), and two Shift Technical Advisors.
04/09/82
.3-6 ROBINSON SER INPUT SEC 3
- 3. 3.2 omariston of training with the Audi Crii NATA wriDT vesset lits and the bases for normal
_ paptheaUPand cooldownrestrictionsr h e~
id odfoW nd hes~
i i cur reklR ent classroom; ~~ iigh~r mphas 1zed the board indications and controls.
- 2 rini ng should i ncl ude
.specific instructions on MDT vessel l i fr os Traninand accidents, however, the training ofase limittocs droom tnstuction.
The training included discussions*
was lheiteminton clss m.
for LOCA and.steamline break accidents.
our.
NofK thes emno n citraa witbPs concrns Ss e
f the two STAs were familiar durin accidents.one of the STAs had not attered the classroom training.
(3) ons gaetPTS taroo trai ning required by the operatorst miigt.T evens; hevers atrinong2 was conducted in the control room, nor werd dpnt enclud at iscussons o
reviewed in detail. In addition, training theno incSl(oter icusins the events in which a steao bb couldedevel o
in t he steanlhe nor th e oent.cal fo competni§:
-rne5i h taln and veselrimit cuves pbreasuk proeure rbetween attempting to control RCS temperature and pressure while not worsening the cooldown.
Three of the five 50s had recent simulatorna triingandureaed gthastea they could adequately control RCS presrecand thepertils dureing uste line break. The other two Ss
- ntws rec og th e tha t pre ws steamline break simulator exercises.
e iz er s
'istutin.Te ranng3nlde-7 04/09/82m
eode do os tunt o
s e
limte insr~flntatiOn (wide range pressur rcde)taer covtry operatr of rtes of pressure rise during thesemiebrarcvey 3 sof el Inerviews
-3.3.2,P icate an excellent
-w i
-p r o rsil i ~ r ves, in adti to good know e x -
n once aendt how pledg cof ti ons could lead to PTS events.- Tey onibtro ! a n he Peent ed is-o cntrol room instruments and equipmen cotrl w uingadt t oe aT evector cussion, which included single-and two-phas ocein s and ict r tior vessel steam bubble, they wer abl to hlloengedurelms.dpeic oto of the recovery procedure which wo
_One of these two.S0s was concerned with the operr'siedt to antcia rapid ate of ressure change anng rs; 8 displ recorder was the only instrument whccud display tnceast.
prh enot
- transient, and adequately depict any apideatn oari icreae. The ormb operator had recently trained at the Sheats Harring stmlin e reke(Sb) the teais' concern on core ubcooln limcotsringsaft ecin bre)and r events and that they could adequatelyus cofo safeay inpsjeothe 50l ad RCS temperature and pressure ris1 bytm use ftneds. aSaThthe SOmau recall specific details of t the heactnesse a d theld ineg Both were concerned that a bubble in the reactowevessl thead elced at control of pressure after termination ofa a hapvd rtey bfieaed ste control secondary plant teaming t temeraureor pressure increase.
with regard to the interviews witha twoa Shf Techndi Advisorsea oSTA) had attended trai ni ng in PTS and ha aso aw~aei of racocor dur limits during normal operation.
e He ad deiofcuT coern sun accidents and events leading to PTrocedures deifclY jHDT nItrmin h temperatures to Tmonito for PTt patcT s mre o T inSI but concluded after discussions that 0eac D o ec did have some probles sd i considers onbty cofse bu l for
-general location. Heddntcnir 05blt ofsambbefo 04/OIBZ ROBINSON SER o
redict rpo 4/09/82
ndtions afterRC
. t headand the posibilitY of twop st-reactor pump rerivess e
did not know the manual actionse rdt anary e w tr bran tprheed ue hot a e
te Si temrt ur to 9essu the S
-rip nrfin o t e procedural manual a sse, to oraseu v v th e r sc omin d o oe foi n g 51aFs th-cted:
explicit)cedre e minot. on a p c ri te r that sduty (. s tand 2t.h hS po c e r ee e on strt rol He fls d ste p r excessu r usings, s team nump i t anc~ in o pee~ t the S1 we n o nt ons dol nt ansient on the ess as mo e
t ee r e t n low and PUMP con m owe erb e a s roev ue serstanO n ofc pdin imits ;
u h ow e not hadl cov sie e d o or O
a n g disu i n t sf iat on r t
ie n
Re abe o d tf f udo not ma dfic ulty h
s sub ooledor caooo mam a rrn a uar Y 198sa u as did n o ree ca l si u a o c ises wh c SI pu e
r co trohs e aeo app o h ed e sT im i s nt wecal i a n u s on 2levents th c h hav e uactr e se T l.im ts. ThshhifT id n t tten at mo t o lec t s
n aa h o w th e r e v s se zth e s ub tr R e p o r teo n He ndiated. aein ba i-f P limits o e e, he i w r i i e d m o
r eo Wf lI Cd g i n
- 'T S -~ro n a n -0l e v n loop temperature indic~~ ~~devationstadi ~ inndgrssucleorpe b t joltrto h aatitincre.
H a t s h licnse operat
-SIpup'hed/ lw le anHlne e ais ta in atn steam ch controls.nt Hde as o a o a p e i t o f. 5 i auilar feeciwatere simuroclator ~ann nSB" achiethee oas H did not recall nyu couldc help him3 in"Se e, no rvos Robinson 2 even tashl ratrvsel PT limit o ea in~ ~
~~
~~~i RStmraueatradcesino Stmatre an prsue.
t hpr a h e re s r z e s u T e e o t e is T o n t e c le g a d a t o b e l ed 4/09 82 e-a R0BIH s w th
-g a
difficulty) to reeval.uate his statement.
During discussions on the qteam line
- brea, heattempted touse the steam generator tube rupture ST)poeuei break,h-tepdt she esempetoalmost..2 minutes to determine his lieu of the SLS recovery procedure.
He ook a n step nc i n c tr ol of error. He did not appreciate possible competing steps concerning ctra f d -pressure increase coupled ith trna
-RCS-tempertr a p.ocduefrsmtm.
n t s-a obvioS -.that he has' no.t 'walked thru" the proce uror t
or Itwa oviusof Lwhen helast had siblao additoii, edidInot recall specifics of thfeuSLrB adlar ing He did ecall (sftY.VaVe falure and large a innthat chalenged reactor ves 1eak -in an-RC pumpe they could have been helpful inreviewing PTShistoY wtly licensed and had received addi.tional simulator training in February 1982. Both werevery knowledgable about reactor vessel P/T limits and the PTS issue; however, both stated that the PTS training was pfTTiitsan te PS sse; owve, Te adworked as a team with other conducted after the simulator training.
They hadTwmets a
of thotr SO candidates and did consider reactor vesselPIT. limis manin n wa godir exercises.
Although they considered that PTS classroom training was good, they did ot eceve pepaed raiing material.
(They apparently were not aware of did not receive prepared trainihbtr cnl lcdi h oto om the PTS reference material which had been recently placed in the control room.1 Both S0s were'exceptionally knowledgeable in predicting SLB responses nd aar ofposbl rpesurztinwith and without steam bubbles in the vessel'head They ecognized that the SLB model at'the Shearon Harris Simulator may not respond to the same event at Robinson 2.
The Robinson 2 PTS training outline wasreviewed riornoath siteerisit on April 5-7, 1982 and found to be acceptable with thegenerdl criteria as well most of the specific criteria. The CP&L training was conducted ver a 2-mont period and consisted of six classroom sessions.
A atendane not required to attend the training sessions; however, STA attendance was not mandaory' No foripal evaluation of the effectiveness of the training was onducod; however, the instructor did question individuals during the class romsessions.
r00m SeS0S 3-1 ROBINSON SER INPUT S 04/09/82
.*g a
3.4 Summary Sside ar hat operator training, specifically on the On ithe positi e
d tea claCP&L. A genefal awareness of brittle fracture.
PTScersuexisad and some personnel interviewed were very good on procedural adontrol bo f
dge(indic ations contr c
procedures used inthe -control roomaf equ hlY efere c do ol ts particularlY those procedures used for
- pnu, peor cool down voluti ons..
Some accident procedures address the PTSissue specifica1y the modifiedo :
mination ritra in El Appendix On -the negative side our audit of seven plant personnel in-the contro1 room produced a varied response from very good topoor. snowled eaOf th t u location of key control room tindicators andeve ntrols, and procedural wal kthe were particularly weak with three of the seven individuals. With regard to the control room emergency procedures, there is no explicit mention of potential brittle fractureconcerns-in the LOCA instructions,' and a relatively high preue (2000 psig) remains as one of the.four SI termination criteria. We also noted that no emergeny procedures addressed strategies on what to do once alooedthtno eimereln procve PTS condition. (specifically-,trying to the operator found himself in a severe
)
I reduce pressure or minimize repressurization)
In addition step 2.9 of Em-p Appendix B, provides minimal guidance to th operator on using stam dump valves to stabilize temperatures.followin~g a steamli e or _feedwater i
ne.
bak. Exce sive ing of steam -could extend the coodown transient. With regard to the PTS classroom training, STAs were not.required to attend the esions and.the absencrof CP&L validation of the learning process were large esons r the ation in PTS knowledge.
The revious overcooling history of Robinson 2hprovides a particularly valuable training tool jwhich was not
..emphasized sufficiently.
the existing 'procedures remain weighted toward core cooling cncerns.
While cl etins prfored conservatively to bound PTS concerns (WCAP-1001 9) have caluerit(nla ogs tomendi re cooling calculations), the use of only cosertivenanalsso Ai not cessarily a sound approach in writing operator conservative analyses is not necs teidsrsnethTI-acdntn guidelines. As has been endorsed by the industry since theo s-2 ent n 1979, more rigorous "better estimate" analyses' are needed to splmn n 1979ppor su grou btguidance.' Such an objective (currently underway as support such procedural udne
.3-11 ROBINSON SER INPUT SEC 04/09/82
part of TMI Action Item I.C.1) is intended to provide a better balance to safety functions needed to migitate the consequences of transients and accidents.
Based on-the expectation-thaItuffint procedural inadequacies will be corrected within approximately one year under TMI Action Item I.C.1 (b6th-from a-technical--
and a human factors staidpoint); we concludethat with t6 exceptions, pro cedural changes should await completion of this program.- Those exceptions are reducing the 2000 psig SI termination criterion and providing additional guidance for stabilizing temperatures following a steam line or feedwater line break.
Also, additional operatbr training should be conducted prior to restart to address the key procedure weaknesses discussed in Section 3.2 (see Section 5.0, "Recommendations").
- P 04109182 3-12 ROBINSON SER INPUT SEC 3
- 4. FRACTURE MECHANICS 4
General e
o udi procedures and. traiig, F rc omincluded in teoli g sections a eoss3f-P tu e ehani c and a s i ay of Robi oh
.. e diet s aesse cp onf!ifoe 'tha cta ue mechanic analyses ad tera s cn expritethas t e y c ne grow t
rltively shallow preexisting crack cfath intaesurate o they cylndrwis.
deeper into a cylindrical metal wal ifl thecring rsftempetre to thder s u b j ec t e d t o a t h e r m a l s h c k b r a p i t r a n t e a t e u p e
r s e r et t h
e d transition region betwee hduhcertosmore brite maeialhi neurncrdathen of the materalwo NOT idocooling7o In adition to the thermal shock which couldsocur sdue so sa nai oling of the primarn regn presre s
and/or the system is repressurized a aria i c o p prsresr For vessels with a relatively high RTNDT' al esikely approach than icosw transent were a new vessel. Therefore,
.partoni dehrationshprescrb tanbs represesu rizuon snhould sbeoasoidedato mnamszeat che potential.for jeopardizing. vessel inhegritS Thisdownsnd seatin t repres to an overall objective of minimi lnt the R co rema ns acool.
q en ep e surization while stll ensuring th.
4.2 Rbs 2
rk tectaoiss In the fracture analyses of pressurized thermaln shocke thEe furectons at is obtained from curves g is a fucton tel materelative to the reference temperature, RTDT ls ofthe Sm o uatiese te initial RTND e
d corndd asorhearules qie Ae no and the RTDT caused byi 10 CFR Part 50.
.RBNO E E 04/14/82..
4-RBXSHSRSC4I subject--------------t-------------------------
d oial noanith suu For Robinson 2, the welds are the cont ling mat e se there sensitivewtoeneutronradiation by virtue of their higher pp content. -Although the longitudinal welds have low nicke c t t radiation), both ongitudinal a c
r i
al w considered since-pressure stresses and the-theril -srse-a ep cks are.-
higher f0 flaws mein v elongitudinal welds..ar o
1
.7 notRob because tb:ese ws:
.ia l
. s 1 w e r e v a s ue d f o r Ros b y t uV
-fbicatedbefore the ASME-Cde
'ieswer n pla e.Forthe ci rcumferenti al':
-welds,.there were three:Chiarpy tet-t 10F. "Fromths reuts a 0o.rv p ant s t e o f N T f
hn i
t i a l R Ta i
w a s2o0u t e t h given in SRP 5.3.2.-From generic data on similar wel ds, wlsmaewt 9e Linde 109 2 flux, a mean value-of -56F and an upper sgma be estimated; hence, the latter is used as a best estimate. For the longitudit welds, there are no records Available, except that they were made with ARCOS B-5 wel d fl1ux.
From a limited amount of information obtained from other plants, the initial RTNDT values were.assumed by us to be the same as those fo the circumferential weldsO*F for the conservative estimate and -20*F for the best estimate.
-The only measurement of copper c a6tent for Robinson 2 welds-is a value of 0.3 for the surveillance weld, which matched the circumferential weld near* the tol ofo the core, but not the weld where fluence was greatest.
Consequently' for our pedcon of nt the copper content of the longitudinal welds was est mat redtobe 0 fbet Testimate and 035% conservative estimate. For the analysis of the circumferential -weld,.0.34% copper was used for the best estimate. For the conservative estimate, the calculated value of shift usin
.34% copper exceeded the upper' limit of Regulatory Guide 1.99, Revision 1, which bounds all known surveillance and test data in this fluence region; hence, the Regulatory Guide prediction was followed, as given below NickeIl b 0.1% and 0.75%, respectively, for the longitudinal a content was taken to be d02 n
.%frtecn circumferential welds (best estimate values) and 0.2 and 1.2 for the consc ative estimates.
4-2 ROBINSON SER SEC 4 1 04/14/82
Ae ous-re jv.in the al5o day" report Fluence values for-the various weld locations are given i to ee vlehut from CP&L dated January 25, 1982 (7.2 EFPY), For the longr D. G.
mated -to a
tudinal weld, the fluence as of Dec er31 1981 wasa o
h
- .3~ 019.nc2
( > MeV) t th 1nie.
5rface qof the wql1 Fo h
a0 1
-( h ai i -e u r ccnfreflt a weld teale s1.4x10'snc 2 (E 4V.
Tect thefluence ocatio.
iclyled i
l wel t'_____
.me end ~
-. us tocalculate. AR was deve pe j n3~Sof T
cur e ue by
-Z T -E i
e cm su illne at G Guthrie of whh t
s f rcent copper, "Cu," nicke ND 40Cu+ 270 Cui (,f/1O'*)o 2 2 ART.-ART[5
-a~
The standard deviation s 22oF.
The mean curve was used by best estimates" and the mean plus 2sigma was ca c
e estimates."
- b"
-is Sub ttutinl te appropriate values in the Guhi ormula, our crrnt vaues of RTD for the Robinson -2 welds are:
NDT Best Estimate Cm Longitudinal
-10 0
-290F Circumferenta 220F Thes values were ty nCommisn tmas tce were compared ith e lpitensales consevtv siae o the longiudiia od crcuf ren ial welds of 183oF and 290rFe respectively.
Current pressuretemperature Appendix Gallimt benedb Robisoy2wheeNR submitted by letter of January477 andherves previouslendeted appy theor in a letter dated January 25, 197od A ekee veTa en t apply for 20 EFPY, or about 13 EFPY beyond tday.
teio R DT as information available today regarding fluence accuuuelaion an R20 ate confirmed our acceptance of the pressure-temper5tFrerron limits92 ad anuary 11, 1982 alerted the RC to a possib e 5 e
r 04/1/82
-3.-ROBISON SER SEC 41H1
_-3 Ths4/14/82 r r tdia
e resolution of this issue is not expected to change the genera conclusion.)
These limits do not aply to cooldown rates les f K -ther hur A nlyha Thae nt h
eese lis tr ap eooling rat the ral stresses produce values ofK -therm that are only a fat preuof Ksprass rai on more severe(postulatn :
prctopaaton ted rever seosr ttrue..h J.7.
2,7-7 T e ce e s
n ca.t e as t
t e f r t he on2 e
ef ini ti vea s
o on an Sinc
_dei vendisse How veeiitsyha e et-c a
too.abl-low cred f r a t:
e p e yi s e v e re r c o o l do w n a n d s ub se q u e n t re p r es s u r i a t i o n l t e i n t preclded ater acooldwn an chr9cmmisn eting,3 thdees oues o ensuring that the coresremains cool. The e
Sestion willsno
- udt f heoperationsstaff at Robinson to detefiine-their. level.of-awareness 1- 'nh-hthe, control" roomk.
of this-concern, and the procedural guidance aalbe.l.
es and training on PTSwere evaluated against:
Preventing or minimizing the potential for overcooling events..
(2 urn a vecon
- ieet, should one occur,limiting RCS pressure to minimize the probabilitY of crak_ intiation.
(3) I
() r 2) aove, is not possible _(ev r apid-overcooling acci dent) limiting RCS pressuret.
propagation.
Thelicnse ha inicaed hatfor the conservative overcooling scenarios.
- aaze inCA-10019, at least 31.EFPY remain f or the Robi nson-2..reactor vese.
Hoevr kytechnical questions on assumptions for theseanalyses ar,
- notyet resolved. An example is when toalwreifr rtpetes(S
- which is dependent on defining the events which create PTSrik Cuen exprienalinoration suggests that the beneficial effects of WPS could be
- precluded after a cooldown and subsequent repressurization later in the transient.
As addressed at the March 9 Commission meeting, the above questiol and uncertainties are being pursued intensively, but final resolution will no, be available for the June 1982 reassessment.
04/412
-4ROBINSON SER SEC 4 IN 4
04/14/82-.~-
0-0 Aside from the primary mission of the Robinson 2 Task Force to audit-.procedures and training, as discussed in previous sections of thii report, the Task Force also discussed what parts of these unresolved questions are of most imnediate interest for Robinsoh 2 pendi ngresolution in 1983.
While conservative worst case PTS scenarios are being-sought and analyzed, our attenton focusedfon the more probable overcooling scenarios (anticipated operational occurrences).
Ptevious staffevaluatiorhasbenared the Ranc co 1978 event as historical reference to a severe overcooling scenario. Given that.
siml ar event is postulated at Robinson '2, WCAP-10019 indicate that at least five_ addi tional-years remain before their defined acceptance criteria for thermal shock transients are exceeded, evenwithout credit for WPS.
Ongoing. staff fracture mechanics evaluations using conservative Robinson vessel properties support a period of at least one year and, using a best estimate RTNDT (see page 4-3) support the five year value.
As indicated in Sections 2.1 and 3.1, recent "better estimate" thermal-hydraulic analyses by Westinghouse to. support proposed procedural guidelines indicate thatthe more likely scenarios-. (such as a stuck open PORV or steam dump) would be bounded by the analyzed Rancho.Seco-cooldown and repressurization scenario.
These Westinghouse calculations are under-review as part of TMI-2 Action Item I.C.1.
4 4
04/14/82
.4-5
,ROBINSON SER SEC 4 INPUT
5rig'ECOMMENDATIONS
- 7.
ro e
ural an i
n tg area.
-~~
jr~4"a
.ndrtheV
-r that additi onal Actn Y.
CP&iswarrant dP u
e to o
aec dtis aare.pr(vi ver
- trestart,
-and pending longer term generic resolution. of the PTS issue, all Ronso 2 operators and STAs should be retraiped in the following areas:
(1)Reiewofprvio~percoln~events at Robinson 2. This includes all available strip charts, event summaries, and review of operator response to mitigate the events.
(2 Reviewthe emergency and abnormal procedures which-challenge core and P/s limits and sketch the typical progress of key pwatran il at is chieved.
This exercise should consider a RCS wit and without a steam bubble at locations other than the pressurizer.ons temigate shfta should review their sketches andn operator response the recog ry th ta i This includes instrume ion and controls during the recovery phase, with a complete walk-thru until conditions stabilize.
Emphasis ad f n
discussing alternatives for recovering from a PTS condition, and alterna tives for minimizing RCS overcooling and subsequent represSuriation.v while still ensuring that the coree remains cool. The shift should proVide feedback of any questionseo or comments arising from these dillsto plant management..Resolution to these questions or comments shoud then follow, with revised procedures and additional trairing as necessary.
(3) A CP&L audit of the shift's ability to cope with a PTS event should be made after the above is completed. This includes a short quizand a drill or demonstration atthe console.
In the longer term, an independet audi+/- of the ability to cope with PTS using the new I.C.1 procedures shouldbe made to verify an acceptable leve of 04/14/82 5-1' ROBINSON SER SEC 5 INPUT
training.- Also, CP&L should review the Shearo atr e
PTS events to verify that the models are reasonve ad demingforced flo bubble(s) in the reactor coolant system (.e.,esselbhead) during
-flow
.-- Idntii
.-anolwmalies-e5 e the £1! Iatrad.;
and natural circulation. Identis d epocess Robsinson.2:responsesishould be.discussed during the trainingp t
e gency poceduresfor safety injection termifation:
ith-
-re ard-tohecurrent emer (1)
We recomend that-prior to restarta the SIotpination critera: a2n00 e_
psig be modified to lower the pressure atgshichithen operator can secure 2 S,..hil stll ~ 5~vjg aequte ~~~ol~gheat sink, and pressurizer SI: while still -observing'adquate subtgoseidcaetatti level. Discussions with the licensee and Westinghouse indicate that this vaue could b t ssaet injection p p cut of head, plus-uncertainties (about 1600 psig).
(2)
-we recommend that prior to restart step2.9 of E eud
~'eaie.
eovr Poedr-Steam Lineor Feed, Line Ruptre,"be revised "DeaiedReovryPrce os orcotrl)i(pg temperature and pressure.,
-to provide clear instructionsfor ontrou n r s following dryout of the faulted steam generat or. Such instructionso should include recognition. of.t potential for ext transient.
cnideration. be given to (3) In the longer term, we recommend more ierion e ghan
-owering the RCS pressure SI-termination criterion further than-s-o (1) above. -For example, an acceleration of the schedule for conversion of the ucoolinlg eter to temperature indication would provide a direct subcooling indication. Such an indication, with asaety grade subcooling meter, should reduce the nee tO acc teiat ncrites with as high-a pressure reference in t oc erminatiol criteia.
Crteiasiilr o hesteam line break procedure (suitably weighted for Crbt e coinan te saconcerns)could then be adopted in the other both core cooling and PTS cnen~o accident procedures.
57148 ROBINSON SER SEC 5 IN 04/14/82
RNS
-T R E'I NA IN 3~
t u th n
t* fe ep
.s Oconee (B&W)
San Onofre (SW)
Turkey Point ThI (B ey p
a Since it is likely that San ofre n Tu kety n p ontse re n e p oedhcurse Rs based oaios gf main aban one, dWP~ O1 sey no app to San o fre due to the absence o
s t e a m l i n e i s o l a t i o n v a lv e s T h i Sdq a n e p r o cre d e s i ag ain p~g r o g aold tn d
.to increase the importanc our findngs oedreRosndb irecgtoy copped to secondde breaks.
fcn a n s nno oeons st audits are plnt speci theans sta SOnofre plants.
Turkey pocedran ntified in Section 2 can be MainewarrantedCto Sappliedto eacho heedr a nt e auditd.
Rviecabof referece transient accident analysesiei Basedon.
- d review, itPPears neaessary toeudit pr msis o
e reainig snplaonts wtht or vesel preessary to dthea Commfision~
riefing in Juneo (Telams beomposed o asi poe rare s n o o tn e e reco m~end that a tea m o re s and p aosed o f orci Nthesrnt LaeratryP.)
personnel' audit the proc~eres~fs and Maine o SnonoresL or ot. Calhoun, Turkey Point, Oconee,num Ce fprocedur alit Sane e e should consist ofa iin e
SinceTh tems
~fsppr 0a/14ee
.1BNSOM SER SEC 6 11 61lation........
training...........
04/14/82ac o dqut poeursan r-r n
S cialist (preferably an operator licensing exapniner,.a reactor systems speci al i st f nor ailyit eaThativ,-and a fracture mechanics rea sste sheci-(a 1neceasal) vsit each.site to expediti Theteam-membersa s
__necssr shou e
thecia personnl and.td discuss questions with
-the licy e aneoessar forall team members (e.g., the at-mcec ialist) to visit each site.
The teands)L(4Pt c.9ndtct an evaliation of each plAnt's training program for SPTS and conduct a technical and human engineering review of each plante procedures-used-durin sible PTS events.
These reviews will use criteria developed from the Robinson 2 evaluation co cf l 982..
It is anticipated hat the site visits will r quire 3-5 ay each.
Therefore t-is at cte tat it n early June, the site visits should be conducted at to-complete th audits_1n 9 Adraft eauto hudb rate of one a-week, beginning April 19 1982.
A valuatio n should be
~the week fol-lowing each evaluation-. It appears that provi ded-at -the -endbofnthep e
ia io usin or more teams will be needed to meet this schedule.
Because of questions S
r e
l rvieof San Onofre-,I we recommend that it be the fi plant to be audited.
The OR project manager for each plantshould attend tt p-lant -
t at.
lie aibseon be tween the rev iew team and the plant, sinct is most familiar with any particular plant problems and iEththe.Resident s
role will primarily be to ensure that the necessar Inspetor ue O LPM' rol JE o ensure an effii documentation and personnel are available at the site, ficie The reports will be submitted to the Generic Issues Task Manager, who may, depending on the findings, request additional evaluation by PTRB, LQB, RSB MTEB..The final evaluation will be summarized by the Generic Isues Task Manager for presentation to the Commissioners.in June.
Should the above multi-team effort not be practical an alternate option I liiin hesteadisto threeor four of the relmaining six plants, wil lestine te svenud coete by June.
This would leave Ft. Calhoun, Ocol and San Onofre As the next-three candidates. - Assuming a team effort is utilized (PNL), the enclosed schedule outline is proposed.
6-2 ROBINSON SER SEC I 04/14/82
Prior to further site audits, however, copies of this Robinson 2 report should be made available to the six ilants.
Inquiry of the licensee should then be made as to whether the key negative findings on training (Section 3.3) at Robinson 2 would apply. A response that similar problems exist should dictate initiation of the training recommendations -in Section 5 prior to any site visit. A positive response (no similar problems) would verify that a meaningful site audit could then be conducted.
9
- 4.
04/14/82 6-3 ROBINSON.SER SEC 6 INPUT
April May June
- 1.
Robinson Review Complete
- 2.
San Onofre Review
- 3.
San Onofre Site Visit
- 4.
San Onofre Report
- 5.
Ft. Calhoun Review
- 6.
Ft. Calhoun Site Visit 2
- 7. Ft. Calhoun Report
- 8.
Oconee Review
- 9.
Oconee Site Visit 1.Oconee Report Summary About 3 weeks each plant (total) 3 day site visit About 1 week writing report 04/10/82 6-4 ROBINSON SER SEC 6 INPUT
3 KEY FINDINGS FROM THE ROBINSON AUDIT
-.3.1 Transient and Accident Analyses
-3.1.1 Introduction This section presents our comments and conclusions based on the material provided in Section 2.1 of this report.
3.1.2. Robinson 2 Overcooling'Events CP&L reviewed the Robinson 2 operating history and presented three events where the cooldown rate exceeded 100aF per hour. The minimum cold leg temperature measured was approximately 3100F during the cooldown for the reactor coolant pump seal failure event of May 1, 1975. In each case reviewed where-operator data was available, the operator actions were different than would be expected with current plant emergency procedures.
For example, for steam line break events, the cooldown transients would-be less severe using the current reactor coolant pump trip criteria (continue to run until 1300 psig). Insufficient current procedural guidance exists to evaluate whether the operator would continue to run additional charging pumps during the small steam line break for an extended period. For a given overcooling event, particularly if the pressurizer does not empty, continued use of additional*
charging pumps could result in rapid repressurization.
For small-break LOCAs, repressurization to 2000 psig may not be advisable following a severe overcooling event. CP&L and Westinghouse believe that repressurization to 2000 psig will not compromise vessel integrity.
04/09/82.
3-1 ROBINSON SER INPUT SEC
3.1.3 inSon 2 Termination Criteria SI ermnaion Durin -QLOCA 3.1.3.1 The temction ring a LOCA re p ereaterthan200 psig and increasing, Pressurizer level at no-load level and responding, Heat sink available (U-tubes covered), and RCS subcooled at least 40 E.
dto core coolingconcerns and do not explicitlY These crte pre wied t ock issue. The licensee has indicated that, based on the estinghouse analyses under review by the staff, no PTS concerns exist during a LOCA.
f SIdurig aLOCA is that the primary One of the criteria for termination of iduring This vaue pri the
- colat ystm pesureis2000 psig and icreaing-This value provides.th
.coolarit system pressure i following information:
The break has been isolated, or the SI fl equal to or greater than the break flow.
n exists to terminate SI before the PORV would be challenged.
(2) Some margin 40OF subcooling margin, (3) Repressurization to 2000 psig further assures a including uncertainties.
At the time the emergency procedure was developed, Robinson did not have the 5~colig eer installed, and core cooling was the dominatinlg issue. T subcooling meterisald n
oecote in instrument readings and floW verify subcooling, and include uncertainties i ins a edit as not conitinsa pimay sstem press)Jre of 2000 psig waadpe. (tinod hadt theRobinsoninmh head msaety injection pump cut-off head is 1500 psia.)
ROBINSON SER INPUT SEC 3 3-2.
04/09/82-
3 ?3. SI TerminatAion Criteria During Steam Line Breaks 31.3.2 termination ri for safety et during a steam-line break as
~ddsste~rssu ze thrma soc k isu ya, presented in Section 2.
4 dp c hange to the LOCA riteria d h~Se nnth seedy 11tf 3wed.The criteria dues the presurat
_raso e
wie conrcud thtb teeciej~
rvd.3ra e lac ewe.c~
.ooli nTS c once rns 3.1.4 T-I I..
A n
a l
y s
e s
4sent FSAR design bases analses are not suited to t evalu on f vessel tegr Insufficient carryout in time exists to pecon events, andare reeents presented in WCAP-10019 are boundinguro Theenlsesnd re rt reseor
.*of design bases events (single falure).rmese aayesnhoue, suiabestr vessel integrity studies5.
Analyses pC siermdab ionsevtsm exing ab ethe estimate" assumptios, indicate et-csidale onlsesrviatmextsat the hl WCAP-1 0019 calculations.
Thesefo bsethae steamysee indce speatrum.
cooldown would not be less than 3xing the smlnbreak Lstr.
hle
-some uncertainties exist wihregard toobmixione for thelsteam LOC, tese loss of RCS inventory eventaperobebuddyth'samiebea Sspectrum.
Th Rhla.nfor the large steam Thlie rea dresendn indSetion 2.1.4.d asucportth aabove observation on the linhobre analese Addietiona 2.udit calculations to be performed duringV r
A re expctd to prode furthr confirmation of the estinghouse thermal' hydraul analyses.
3.2 Procedures 3.2.1 d c o
ed d
e s e al ted as d Isc sse In pSec t 2 rel n to PTS and the r bases, and anO audit cluded tl roedre so d r legibility and currency.
Orad pr9/ce s t t3e ROBINSON.SER INPUT SE Ins~~~~~~
nt3ad-3 ersettv 04/09/eve
following Emergency Instructions (Els), Abnormal Procedures (APs), and General Procedures (GPs):
EI-I Incident Irvolving Reactor Coolant System Depressurizaton, EI-6 Loss of Feedwater EI-7 Station Blackout OperationoTi
>E'4ReacoTrp (at A) Turbine and Generator i'Pat)
Ap-19 Malfunction of -RCS -Pressure Control System AP-2 Loss of Instrument Bus
- GP-2 Heatup -(Col d Sol id to Hot Subciia at Nolbd~G GP-3B Reactor Trip Reovd GP-5 Shutdown (Normal Plant Shutdown From Power Operations to Hot Shutdown Conditions) aio ntod Ps Control Using Natural Circulation GP-PCldon Tempratur Co doFreureH Shutdown to Cold -Shutdown Conditions) 3.2.2 Comparison of Procedures With the Audit Criteria (1) Procedures should not instruct operators. to take actions that would violate NDT limits.
The procedures audited generally. did not appear to
-. contain instructions which would cause an operator to violate NDT -limits; most of the procedures referred to, or included cautions to stay within,
-'the limits of tihe'NDT curves..These curves are consistent with the technical specification heatup and cooldown limits. The only area uwher
-the procedural instructions may violate these limits (dventhough-6autions exist). isthe safety injec'tion termination criteria and charging' pump operating -instructions Aiii the loss-cif-coolant accident Procedures,
-The termination criteria require RCS pressure greater than 2000 psig and increasing prior to.terminating high head safety injection (shutoff head approximately 1500 psig). There are no explicit instructions for pressure control or operation of the charging pumps until a controlled cooldown/
depressurization is begun using GP-6. Discussions-with Westinghouse representatives indicated that the SI termination criteria are under review as part of the generic procedural guideline development and it is anticipated that they will be changed to a lower pressure, at least for the plants having intermediate head SI pumps like Robinson 2.
3-4 ROBINSON SER INPUT SEC 04/09/82
(2) prce~re sal roid udanlce on recovering from transient Or accoiden codtol ihu jltn O r strto jiS. See
-item.
(1 aov fr jSUS1Ol n OTljjits., The procedure for depressuri ev nt)( apr er te operator to Curv. 5 n p oie
.zation eet ig fr-
. n s t r c t i n m a i n t a i n
,: -at e a s t 4.0 0 -F
-uC ~ ~ - r e a t o c o n t u S are trpetepoeuefrntrlcrul ati on
-i nstructs-ithe Opeao D rescu rto nmte
-tr l ci r eu~
pro e0ue-is~ a.
Th-. - Iov: -
a9 saturation curveaurPo sowmingana et when tem~eruia pressufolwn temperatures-i
- ah st-.t nc ea conii onm-alow doylhed curves, i s n taprn h ttepo con iti ln che r e n lci n tr c i n to0 th-p r t r o t o faulted stea geeao h Itr int rterihe per torlm t
followingfo a semlnbe Ownrs'Grop
~
guielilesThee gid hspeifi d um no the dlnswic ol mat Pt S re as o ositn sentat e n sttedptatutre gidei s ar loben eiwe gis t ep rtesti t frcu e ehais aayss adtatti for il b crotomplted in ay 1982 stea Setos 21ad 31fo icsin o h safety inecio terinaio crteia codtin ro-i eo tOBheO R SE w tI~nPU E
procdure shold_ ov 3-5_
id,u109/82 1
(5)
High pressure injection and charging system operating instructions 'should reflect a consideration for PTS. The 700 psig SI termination criteria for steam line breaks reflect PTS concerns.
The SI termination. criteria for loss-of-coolant accidents would allow repressurization to above 2000 psig with a cooT vesse.
h-Tere are no specific instructions for operation of the charging pumps following the depressurization tranent.
(6)
Fedwat r (FW) and/or auxiliary feedwater (AFW) operatinqjinitructions should reflect PTS concerns.
In-ructions are provided in the steam generator tube rupture and the loss-of-coolant accident procedures o terminate FW/AFW flow to the faulted steam generator.
These and other procedures provide instructions to maintain steam generator levels -in the good steam generators within a defined band.
(7)
An NDT curve and a saturation curve should be provided in the control room..
These curves are provided in the Curve Book located in the control room and are referenced in the applicable procedures. Each of these curves is on a pressure-temperature plot. Curves 3.3 and 3.4 show the technical specification heatup and cooldown limits.
Curye 3.5 shows-the. saturation curve and a 40OF subcooled curve.
The control room copy of the procedures and curves that we audited was legible and current.
3.3 Training 3.3.1 Introduction The site audit of CP&L's PTS training program consisted of-a review of the lesson plan used for classroom training and personnel interviews with five Senior Operators (two of these SOs were Shift Foremen), and two Shift Technical Advisors.
04/09/82 ROBINSON SER INPUT SEC 3
.04/09/82
.-- ~-'r
3.3.2 Comparison of training with the Audit Criteria (1) Training should include specific instruction on NDT vessel limits for NOAL modes of operation.
All senior operators (SOs) and Shift-Technical Advisors (STAs) were aware of IDT vessel limits and the.bases for normal 1 t heatup and cooldown restrictions.- The-SOs exhibited a good knowled9i in the use ofplant procedures, control board indications and 'control S.1 and selimi cu es Recent classroom paipigg:ad.re-ehasized be reason for-these limits. Both STAs lacked a familiarity with control board indications and controls.
(2). Training should include specific instructions-on NDT vessel limits-for.
transients and accidents. Training was conducted to emphasize concerns of vessel limits during transients and accidents, however, the training was limited to classroom.instruction. The training included discussions of the termination criteria for LOCA and.steamline break accidents.
Four of five SOs and one of the two STAs were familiar with PTS concerns during accidents.
One of the STAs had not attended the-classroom training..
(3) Training should particularly emphasize thse events own to require operator response to mitigate PTS.
Classroom training included acti os required by the operators to mitigate PTS events; however, no training was conducted in the control room, nor were past events at Robinson 2 ed in detail. In additin, training did not include discussionsa events in which a steam bubble could develop in the RCS (other than the pressurizer), nor the potential for competing concerns in the steamline break procedure between attempting to control RCS temperature and pressure while not worsening the cooldown.
Three of the five SOs had recent simulator training and recalled that they could adequately control RCS pressure and temperature during a ste line break. The othertwo SOs did not recall the details of previous steamline break simulator exercises. It was recognized that there was 04/09/82 3-7 ROBINSON SER INPUT S
limited instrumentation (wide range pressure recorder) alert.the operator ofrates of Pressure rise during the steam line break recovery a.7 3.3.2.Pronl ntvew The i 1 a.1ntv. s-with two SeniorOperators (S indicated an excellent onda oftvesse esure/teperature DT (P/T) limits and basis for acurvesoin addition-to a good knSowldge of-pTS concerns and how plantcondi' tions couldlead to.PTS events.- They-exibited an excellent knowledge-of control room instr.uments and equipment controls.
During the PTS event dis cussion, which included single-and two-phaseflow in additiont to a reactol vessel steam bubble, they were able to follow procedures and predict portic of the recovery procedure which would challenge P/T limits.
One of these two SOs was concerned with the operator'snability to anticipa rapid rate of pressure change'using meters.
drecognied that the widet recorder was the only instrument which could display the past and present transient,. and adequately depict any rapid rate of increase.
The other operator had recently trained at the Shearon Harris Simulator..
He rememb the team.' concern on core ubcooling limits during steam line break (SLB events and that they could adequately control safety injection (Sl)'and T RCS temperature and pressure rise by use of steam dumps. The-other SO d recall specific detailsofutheias time he witnessed an SLB at the simu rBoth were concernedthat a bubble in the reactor vessel head could.negat control of pressure after termination of SI; however, they believed they control secondary plant steaming to negate a rapid rate of primary syste temperature or pressure increase.
With regard to the interviews with two Shift Technical Advisors (STA),
had attended training in PS and had a good understanding of reactor ve alimits during normal operation. He was also aware of PTS concerns duri accidents and events leading to P15.
He had difficulty Sdentifying whi t monitor for PTS (procedures identify HDT in SI terin but concluded after discussions that is more of interest than TH udid have some problers identifying meters on the console, but knew the d general eoctipon.
He did not consider possibility of steam bubble for
.23-8 ROBINSON SER It 04/09/82.
the reactor vessel headad ddohekossibility of twophase.olditlons after RC pump aretriped.
He dd nt know the manual actions reqie o n e~o tipmprs d he-find the procedural manual actions to terminate auxiliary
- feedwaterin the zaffected loop for a steam.line break. (Procedure step s no nd. e cIPe)
SLS p rocedutes did not appreciate-that two stepst(2A and hs p ou
.oncerning control of RC$ temperaturesse an cod mps h the S
~ivoveaotercolngVnsient on the essel and could compete" wth h 7 "volve another cooling tranit ont'ticI o
upontrol - was termination criteria.
Some difficulty soctiwnthe Shift Foreman' that demonlsrated.- He feels that his duty isnto imts but oesno bF that may be violating procedure stepsnor exceeof liats o
es n strate is ready to contribute to any discussion of deviations or changes n Rateg whenconitins o nt mtch procedures.
He is in training for.a Olcn when conditions do match proHedd not recall simulator exercises which approyapply essel uary 1983. n1 significant Robinson 2 events which approached vessPITlimits The her STA did not attend the PTS lectures; have challenged P/T-limits. The otr nVse nert y
-01
- howve, e asreieedth Suniary Report on Vessel Integrity (WJCAP10019 however, he has reviewed the SumayR~
iis oeeh saaeta Heidicated a basic understanding of P/T limits; however, he is aware that He idicted aedge basPc ue acgund and possible events.
He had considi
-acedsimoretknoledge in ea und specific controls, and meters on the abedifficulty in locatingeqpmnstinrrtngheRSwdra ng bor.
He alsoclY-ihiterrtn theRCS wide-ran c o n t r o l b o r.I S a contop te aur e inaion and in termining degrees subcooled or pres top aturca tration curve. He had to ask the licensed operate tsI u hadow o
valuesand also needed assistance in locating steam duml Sumplary edw aue ondrols. He also had no appreciation of possible auxcompinga sepsd terc nation of SI nd controlling RCS temperature and competing steps of termintno hotocnrlhesodayytm pressure increase during an SLB, nor how tosconto te iinon syst achieve these goals. He did not recall any Rob ine ng ona cLlerw could help him in PTS events, nor previous Robinson 2t reactor vessel P/T limits.
nnterviewedo has not been on shift for almost two months.
One Sha rece ve wed trainig h obelieved that the PTS concerns were an he ~
~
~
~-
ia reevdPSnriig RCS temperature and pressure.H in RCS temperature after a decrease in co teg and pre le. (
tht heprsrze srg i is.
on the cold leg and had to be led (W that the pressurizer surge Sne 3
ROBINSON SER IN 04/09/8.2
difficulty) to reeval.uate his statement During discussions on the team line break,he attempted to usethe steam generator tube rupture (SGTR) procedure in lieu of the SLB recovery procedure., He took almost 2 minut to determine his.
errorf He did not appreciate possible competing steps concerning control of RCSro r H e di d pressre e pos ed withterminatjon criteria for SI.
te pedtr
-p e s r n r a ecouple1 w t RS tp that e he a
walked thr
.the procedurea forsome tme.
n a
iohe did not recalI specifics of the SLB hen he 1 ast had simulator ditin h
ddno rcllspc
~e failure and large r ining. H
-re 1call
-two -Robinson 2 events (safety fau a
larg leak in an RC pump) that challenged reactor vessel P/T mits.1-e ieves they could have been helpful in reviewing PTS history.
The final two SOs were recently licensed and had received additional simulator training in February 1982.
Both werekvery knowledgable about reactor vessel P/T limits and the PTS issue; however, both stated that the PTS training was conducted a the simulator training. They had worked as a team with other SO candidates and did cnsider reactor vesselP/T limits in. many of their exercises. Although they considered that PTS classroom training was good, th, did not receive prepared training material. (They apparently were not aware the PTS reference material which had been recently placed in the control rooff Both 50s were exceptionally knowledgeable in predicting SLB responses and awi of possible repressurization with and without steam bubbes in the vessel he They recognized that the SLt model atthe Shearon Harris Simulator may not respond to the same event at Robinson 2.
The Robinson 2 PTS training outline was reviewed prior to the site visit on April 5-7, 1982 and found to be acceptable with thegenera l criteria as -we most of the specific criteria. The CP&L training was conducted ver a 2-mo period and consisted of six classroom sessions. Al licensed personnel wer required to attend the training sessions; however, STA attendance was not mnaory. No formal evaluation of the effectiveness of the training was conducted; however, the instructor did question individuals during the cla!
room sessions.
04/09/82
- 3-10 ROBINSON SER INPUT
3.4
-Summary positive side, it was clear that operator training, specifically on the t
e ssv hd be conducted by CP&L. -A general awareness of brittlefracture So a ciul rnyt s procedu es s e d t fo oT i s e p c fically the modified SI Pterinse ati d brtena inE,Apendi v
~os o ebry goo-ontpr On-the negative sid, ur audit of seven plantpersonKel ndtheocottPT r ose produced a varied response from very c orodar o poor.osonde oft he wal isu, location of key control room d o
a n inr ls
.procedrgar toithe were particularly weak with three of the sen expiitas m thno r tedti control room emergency procedures, ere adicit aeltion ptei ecthe LOCA instructions,a re presste fractur op emns asn n of the four SI termination criteria. We alssred that o emains aprocedures addressed strategies on what to do once alse oetr fond imerelfn psevetresPTS condition (specificallyaltrying to rede opessueor founimsfZe nreprsessuization). In addition, step 2.9 of ElI, ped pr ovdes mini 1 guidance to the operator on using stgam dump appnlxvpoies tosaiietmpneateg foliwi~g a stam-line or feedwater line brak.e Excsbile dumpng team could extend the coidown transient.
With bregard Excee dupngao tam 5TAs were not required to attend the segardio anthe Sasnceoom CPa validation of the learning process were lagi resons andrthe vabs n if PTS knowledge.
The previous overcoli g hstor rofasRobinsoon 2 provies aparticularly valuable training tool which was not' emphasized sufficiently.
e w
hted toward core cooling cncerns.
While Thelexstins prforedesratiniel to bound PTS concerns (WCAP-100 19) hav calculatogs tormedevcore cooling calculations), the use of only coseratiannalouse Aen necessarily a sound approach in writing operato guidetine alys s eno edred by the industry since the TI2 accident 1979elmore rigoos bete e st e"needed to supplement and csupror such procedura gudae. e Such an objectve (currently underay as e3-11 ROINSON SER INPUT 04/0/82
part of TMI Action Item I.C.1) is intended to provide a better balance to safety functions needed to migitate the consequences of transients and accidents.
Based on-the expectation-thatcurrent procedural inadequacies will be corrected within approximately one year underTMI Action Item.I.C.1 (both from a technical and a human factors staidpoint),cwe 6nclude that with tw6?exceptions-pro-cedural changes should await comletion of this program.- Thos exceptioire reducing the 2000 psig SI termination criterion, and providing additional guidance for stabilizing temperatures following a steam line or -feedwater line break.
Also, additional operator training should be conducted prior to restart to address the key procedure weaknesses discussed in Section 3.2 (see Section 5.0, "Recommendations").
04/09/82 3-12 ROBINSON SER INPUT SEC 3
- 4. FRACTURE MECHANICS 4.1* General e mudi t*
p rocedures and traiiflg, the Task r also incuded in th -11loing sectionSa discusiO fracture cl~
nd, a n
u~ r
.f R b n on..tct rv se properties ',
cr mechanics ana s a therma1 shock experiments ave confred tat Fracture mechanics alyestand cr can initiate, that s they can grow<
relative y shallow pre-existing cac sf the inner surface of the cylinder is ee ito a-cyli rical etay wapidl dereasin its temperature to the rueoneo to hermal s i.dctit trnti temperature or lower.
This it io between ductile to more brittle material s reence b the transito reio bteni eoe nmanitude with neutron irradiation.
RTHD of he materia, which increases in magniuewt etrnirdain RT Of thmaea, naddition to the thermal shock hich could occur due etoa rapid cooling of
- thebeltlin regionof a reactor vessel,pressure stressescaaloeitf th periar coIant pressure is maintained and/or the system is repressurized the primary coolant pressure For vessels with a reiatively high RTNDT' after an raint s
re likely to approach the transition tempel particular cooldown transient wreto occur in a new vessel.
Therefore, P ature than if the same transin wr iztoshould be avoided to minimize considerations prescribe that repressurization sh s coidedato translat the potential.for jeoparditin vesse herity.
down and subsequent repres to an overall objective of.inimizng the core remains cool.
surization while still-ensuring 4.2 RObinson_
ractureehnics ne shcthe fracture toughne In the fracture analyses of pressurized thermal shoASE Code as a functio o is obtained from curves given it s ahuto temeraurereltive to the reference temperatue RTT. tisheumo R
measurd according to the rulesualte AMEpp adte !&RTH caused by radiation damage and measured as required by Appel 10 CFR Part 50.
0411182-
-1.ROBISON SER SEC 4 4-1 04/14/82
For Robinson 2, the welds are the controlling material now and in the future because they are more sensitive to neutron radiation by virtue of their higher c pper content. Although the longitudinal welds have low nickel content (less sensitivity to radiation), both longitudinal and circumferential welds. must be considered since pressure stresses and the thermal stresses at deep cracks are higher-for flaws in longitudinal welds.
Init T
v were not measured for Robinsol 2 becaus the vessel wast; fabricatedibefore the ASME Code.rules were in pFor theci rcumferential welds, there were three Charpy tests at +100F. From these results, a conserva tive estimate of GOF for their initial RTNDT was obtained by using the methods givenin SRP 5.3.2. from generic data on similar welds, welds.-made with Linde 1092 flux, a mean value of*-56*F and an upper 2-sigma value of -20*F can be estimated; hence, the latter is used as a best estimate. For the longitudi welds, there are no records available, except that they wtere made with'ARCOS B-5 weld flux. From a limited amount of information obtained from other plants, the initial RTNDY values were.assumed by us to be the same as those fc the circumferential welds--0OF for the conservative estimate'and -20OF for th, best estimate.
The only measurement of copper content for Robinson 2 welds is a value of 0.3 for the surveillance weld, which matched the circumferential weld near. the tc of the core, but not the weld where fluence was greatest.
Consequently, for our prediction of RI.T the copper content of the longitudinal welds-was es'i mated to be 0.30%.best estimate and 0.35% conservative estimate. For the analysis of the circumferential weld, 0.34% copper was used for the best estimate. For the conservative estimate, the calculated value of shift usin
.34% copper exceeded the upper' limit -of Regulatory Guide 1.99, Revision 1, which bounds all known surveillance and test data in this fluence region; hence, the Regulatory Guide prediction was followedasf given below Nickel content was taken to be 0.1% and 0.75%, respectively, for the longitudinal circumferential welds (best estimate values) and 0.Z and 1.2 for the cons' ative estimates.
04/14/82 4-2 ROBINSON SER SEC 4
Fluence values for the various weld locations are given in the "150 day" reprt to D. G. Eisenhut from CP&L dated January 25, 1982 (7.2 EFPY).
For the longi tudinal weld, the fluence as of December 31, 1981 was estimated to be l.30 11
>.I MeV) at theinside urfacE fthe wel.
orThe was 101 n/'
dI F-the.--
circumferential wel the yalue was 1.24
- 101 (E>1MeV). (The cri i al 1 is eloi t eak axial fl uen ocation.)
e trend cur.e used byus to calculate ARTNDT-was dev pe analysiS of v136 PWR ruveianedata points by G-t e :of HEDL His mean ty,,
- o.
smau which has terms for percent copper, Cu., nickel, "Ni," and fluence ' is:
NT=[
+
480 Cu.+
270 Cukij
(/10 e
o*2.
The standard deviation was 220F. The mean curve was used by us to complete t
-"best estimates" and the mean plus 2-sigma was calculated for the "conservati estimates."
Substituting the appropriate values in the Guthrie formula, our current valui of RTNDT for the Robinson 2 welds are:
Best Estimate Conservative Estimate Longitudinal 140 F
-240PF Circumferential 220aF 290OF These values were reported b us in a Commission meeting-on March9 19,a2 i were compared with the licensee's conservative estimates for the longitudin and circumferential welds of 183oF and 290*F, respectively.
Current pressure-temperature Appendix G limits being used by Robinson 2 wer submitted by letter of January 4, 1977 and were previously accepted by the in a letter dated January 25, 1977.
The curves are intended to apply for 20 EFPY, or about 13 EFPY beyond today. A recheck of these limits against information available today regarding fluence accumulation and RTNDT has confirmed our acceptance df the pressure-temperature l imits.
(An LE& datei January 11, 1982 alerted the NRC to a possible 5*F error in the PIT limits 04/14/82 4-3 ROBINSON SER SEC 4
resolution of this issue is not expected to change the general conclusion.)
Thee lmit donot apply to cooldown rates exceeding-1000Fper, hour. At that These limits do noe apoyf K tera that'are only a cooling rate. the thermal stresses produce values of h o fraction of Kl-pressure, whereas in more severe (postulated hermal shock transients the reverseis
_rue.
Sdentrit t!e fracte citeria beyondh nce definitive cooldow tr t
ApedxGlimit5have yetto be. decided, it is therefore of nters t&
Appendix Gre 1 nt-hsuiair tln minimize any severe RCS-cooldown and subsequent r-erctioadd sse uti ;
'~uigthat the coreremains cool.
Th preceding Section3 addtre5S~u~
audit of the operations staffat Robinson to determine their level of awareness ou this concer and the procedural guidance availab e in the control room; The procedures and training on PTS were evaluated against:
(1)
Preventing or minimizing the potential for overcooling events.
(2) During an overcooling event, should one occur, limiting RCS pressure to minimize the probability of crack initiation.
- () I
()
o (2,
- bov, i nt possible (severe, rapid overcooling accident),
limiting RCS pressure to minimize the probabilit o roughWlaccrack propagation.
e licensee has indicated that for the conservative overcooling scenarios hyzed in WCA-10019, t 31 EFPY remain for the Robinson 2.reactor vessel. However, key technical questions on assumptions for these analyses are voee.solved. An example is.wheh to allow credit for warm pre-stress (WPS) whic i ependent on defining the events which create PTS risk. Current experimental information suggests that the beneficial effects of WPS could be precluded after a cooldown and subsequent repressurization later in the transient. As addressed at the March 9 Commission meeting, the above question and uncertainties are being pursued intensively, but final resolution will not be available for the June 1982 reassessment.
4 44 04/14/82
.ER SEC 4 INPL 04/14/-
82
Aside from the-primary mission of the Robinson 2 Task Force to audit,procedures and training, as discussed in previous sections of this report, the Task Force also discussed what parts of these unresolved questions are of most immediate interest forRobinson2 pendin resolution in 1983.
hile conservative worst case PTS scenarios are being sought and analyed our attention focusedfonthe t
iore:probable overcooling scenarios (anticipated operational occurrences).:.
Previous staff7evaluatiori has GbIench arked the Rancho Seco 1978 event as historical reference to a severe overcooling scenario. Given that a similar event is postulated at Robinson 2,-WCAP-10019 indicates that at least five-addi tional-years remain before their defined acceptance criteria for thermal shock transients are exceeded, even without credit for WPS.
Ongoing staff fracture mechanics evaluations using conservative-Robinson vessel properties support a period of at least one year and, using a best estimate RT T (see page 4-3),
support the five year.value.
As indicated in Sections 2.1 and 3.1, recent "better estimate" thermal-hydraulic analyses by Westinghouse to. support proposed procedural guidelines.indicate thatthe more likely scenarios (such as a stuck open PORV or steam dump) would be bounded by the analyzed Rancho.Seco-cooldown and-repressurization scenario. These Westinghouse calculations are underreview as. part of TMI-2 Action Item I.C.1.
04/14/82 4-5
-ROBINSON SER SEC 4 INPUT
-,;7.7r 7.......T
SRECOMMENDATIONS Based the summary f
in Sec 3.
tc ns ahe keuy procedural andtrain!ig shortcomings the Robinon 2P T
a a.
that additional
_actonby CP&
s warrantedjpartcu atlyj - to Thelfollowing recommendation g-rez rovided?
ZZm edve yta Prior to restart, and-pending longer term generic resolution of the PTS issue, all Robinson 2 operators and STAs should be retraiped in the following areas:
(1) Review of previous overcooling events at Robinson 2. This includes all available strip charts,. event summaries, and review of operator response to mitigate the events.
(2) Review the emergency and abnormal procedures which challenge core and PIs limits and sketch the typical progress of key parameters until recovery is achieved.
This eercise should consider a RCS with and without a steam bubble at locations other. than the pressurizer.
As a team, each shift should review their sketches and operator response to mitigated the transien.
This includes instrumentation and controls during the recovery phase, with a complete walk-thru until conditions stabilize. Emph4si shbuld focus on discussing alternatives for recovering from a PTS condition, and-alterna tives for minimizing RCS overcoolifg and subsequent repressurization d
while still ensuring that thecore remainscool.o The shiftlshould provide feedback of any questions or comments arising from these drills to plant management. Resolution to these questions or comments should then f
onaow, with revised procedures and additional trairing as necessary.
(3) A CP&L audit of the shift's ability to cope with a PTS evenf should be made after the above is completed. This includes a short quiz and a dril or demonstration atzthe console.
in the longer term, an independent audit of the ability to cope with PTS using the new I.C.1 procedures should be made to verify an acceptable level of 04/14/82 5-1
.. ROBINSON SER SEC 5 INP
training.
Also, CP&L should review the Shearos Harris SimulatOrresponse for PTSevets o vrif tat he odes ae easonable and can demonstrate steam PTS events to'verify thatth vessel aead dur ing forced floW bubes) the reactor coolant system (I.e., vessel head) durin foce0 fo abdbnltua i rculati on.
Identified anolmalies between the slulator and Robs na2 responses.should.be.discussed during t training process.
-tre are d to tbe curret emergency proceduresfor sat injection terminati n (1)
W recginedthatprior to restarth SI termi nation cri teri a--of 20 psig be modified to lower the pressure at wh eat sink an ecurer SI, while still observing adequate subcoolinghe s dpressuris level. Discussions with the licensee and Westinghouse indicatetatie value could be the safety injection pump cut-off head plus. uncertainties (about 1600 psig).
(2)
We recommend that.prior to restart step.2. 9 of E-i, Appendix B,
-"Detailed Recovery Procedure-Steam Line or Feed Line Rupture," be revised to provide clear instructions for controling temperature and pressure.
following dryout of the faulted steam generator. Such instructions..
should include recognition of the potential for extending the overcooling transient.,
(3)- In the Tonger term, we recommend more consideration be given to' lowering the RCS pressure S1 termination criterion further than--r (1) above. -For example, an acceleration of the schedule for conversion of the subcooling meter to temperature indication would provide.a direct subcooling indication. Such an indication, with a saety grade 5ubcolig meershould reduce the need td accommfodate uncertainties subcooigh ma pressure reference in the LOCA SI termination criteria.
Criteria similar to the steam line break procedure (suitably weighted for Crieri siila tod ten e adopted in the other both core cooling and PTS concerns)could then b accident procedures.
5/
ROBINSON SER SEC 5 IN 04/14/82 7.2.....
0ICABIY TO REMINING SEVEN PWRS seven PWRs which have been ide ified as repo plants having a relatively high RNDTe..
ft. Calhoun (CE)
Oconee (B&W)
San Onofre (W)
Turkey point (W)
-Calvert Cliffs (CE)
THI (B&W)
Since it is likely that San nfe an Ture int emergency procedures are, l i k e~u k e Ro i r a l t ncou opeatinsoat San Onofre pants.r~
thn e itur ke y t W sigos gtiieline ct on,,anb l i k e R o b i n s o n, b a s e d o n s i m l a r i n it n
dt ineO e g u d e ri n c trn ouse conlusonswould probably equallY apply.
aobtinne of suprtnaWinos a
pnl s ie oes of m a y pt a p p l y t o S a n O n o f r e d u e t o t heapb s e
.anaccden (W analyss is stem in iolaio v lv s. This San Onofre design co.figurat o
pe wd ten ct of in reast ht i opo r sanc o doeq uae procedures and saining toe e
ecd ary t oe br fndigs on bref g
tthe Ro inson traim cing progrs
'brr Ep and San Onofre plants.
04/14/82 n-eto_2cnb The general procedural and triigciei dniidie t ansb appled t eac ofthe lant tobe audited.
Review of referencetrnit accidnt anlyses is warrated to veritfy appialtYtpan and acdn n
c0nf igurati on.
Based on the problems disclosed during the Rbnoreewors itapeass necssry o udit six of the remaining seven plants wit wort es e d pretesro to th Connsinbiefing in june (TMI-1 may. b e e.xclue the taes nopratn)Woer
~ilndta team or teams composed of Pa Norhwet Lbortor (PL) ersonnel'audit the proceduresad~ann o
SNo ore 1,boFtCoun, Turke Pont cnee, Calvert Cliffs, and Maini Sank. OnTe t al)houldcnis f as a minimum'. procedures evaluat 6-iROBINSoH SER SECI 04/14/82
~e e
plant operations specialist (preferably an operator licensing exapiner), a reactor-systems specialist for analysis evaluation, and a fracture mechanics specialist. The team members (as necessary) should visit each site to expedit the audits, to interview operations personnel--and to discuss questions with the licensees.- It may-not-be-necessary for all team members(e.g., the fracture Imechanics specialist) to visit each site.
The team(s) will conduct an-evaluation of each plant's training program for PTS, ah-duct-a-technical-.-and-human engineering review of each plant's procedures used during possible PTS events.
These reviews will use criteria develo ed from-the -Robinson-2 evaluation. conducted April 5-7, 1982..
It is anticipated that the site visits will require 3-5 days each.
Therefore to complete the audits in early June, the site visits should be conducted at rate of one a week, beginning April.19,1982. A draft evaluation should be provided at the end of the week following each evaluation. It appears that or more teams will be-needed to meet this schedule. Because of questions raised during the SEP review of San Onofre 1, we recommend that it be the fi plant to be audited. -The OR project manager for each plant should attend th plant visits to provide liaison between the review team.and the plant, since is most-familiar-with-any -particular plant problems and with. the Resident Inspector.
The OR LPM's role will primarily be to ensure that the necessary documentation and personnel are available at the site, to ensure an efficier evaluation..
The reports will be submitted to the Generic Issues Task Manager, who may, depending on the findings, request additional evaluation by PTRB, LQB, RSB, MTEB.
The final evaluation will be summarized by the Generic Ishues Task Manager for presentation to the Commissioners.in June.
Should the above multi-team effort not be practical, an alternate option is limiting the site audits to three or four of the remaining six plants, with least one per vendor complete by June. This would leave Ft. Calhoun, Ocone and San Onofre as the next three candidates. - Assuming a team effort is utilized (PNL), the enclosed schedule outline is proposed.
04/14/82 6-2 ROBINSON SER SEC 6
Prior to further site audits, :however, copies of this Robinson 2 report should be made available to the six plants.
Inquiry of the licensee should then be made as to whether the-key negative findings on training (Section 3.3) at Robinson 2 would apply. A response that similar problems exist should dictate initiation of the training recommendations -in Section 5 prior to any site visit. A positive response (no similar problems) would verify that a meaningful site audit could then be conducted.
4S 04/14/82 6-3 ROBINSON.SER SEC 6 INPUT
April May June 1.- -Robinson
?Review.Complete--.-T.
- 2.
-San Onofre7 Review
- 3.
San Onofre Site Visit,,
4.: San Onofre
- Report,
- 5. Ft. Calhoun Review
- 6.
Ft. Calhoun Site Visit
- 7. Ft. Calhoun Report
- 8. Oconee
-Review
- 9.
Oconee SitelVisit.
1d. Oconee Report Summary About 3 weeks each plant (total) 3 day site visit About 1 week writing report 04/10/82 6-4 ROBINSON SER SEC 6 INPUT ROISNSR
.2.13 H. B. Robinson Termination Criteria 2.1.3.1 Reactor Coolant Pumps (RCPs) are tripped when the primary system pressure 'falls to 1300 psig n
addition, the RCPs are tripped if seal cooling is lost, if excessive seal leakage occurs, or if excessive vibration occurs.
2.1.3.2 Auxiliary Feedwater Auxiliary feedwater is isolated to the steam generator identified as faulted for steam line breaks or steam generator tube rupture. The flowrate is limited to 400 gpm to any steam generator.
2.1.3.3 SI Termination During LOCA The termination criteria for safety injection during a LOCA addresses core cooling.
No reference to pressurized thermal shock is provided.
The termina tion criteria include a 2000 psig (and increasing) requirement..
2 SI Termination During Steam Line Break The termination criteria for safety injection during a steam line break are:
m One RCS THOT less than 4600F, RCS pressure greater than-700 psig (stable or increasing),
Pressurizer level greater than 20% (heaters covered),
RCS subcooling greater than 40aF, and Heat sink available (U-tubes covered).
As shown, one of the criteria for terminating SI during a steam line break is one wide-range THOT reading less than 4500, with wide-range primary coolant 04/14/82 2-4 ROBINSON SER INPUT SEC 2 04/14/82 -.....-
system pressure greater than 700 psig and stable or increasing.
line value is 350F, T This value includes all uncertainties house guidelievlei HOT and does imply reference to the downcomer temperature.
r The uncertainties clude core heatup duting natural circulatione ECC Pixing and instrument errors.. Westinghouse has reviewed their fracture data for t
.%ide range of transients and,-for the most limiting vessel at end ofhlife, they oncludeithat--the--350*FTHO would -not result in vesseil ~failure.'- Thbe J7Q 2psigi stable otincreasing pressure as est a
LOCA doesnot exist coincident with the steam line break.. Robinson 2 has increased the 350*F cvalue to 460'F to provide a combined assurance that.40oF subcoolingexists at a pressure of 700 psig, concurbent with a sufficiently high temperature to no accommodate brittle fracture concerns. Also, it is noted that the Westinghouse 350oF/700 psig values would violate the Robinson 2 NDT limit for 100OF/hr cooldown events.
.2;1.4 The ialydraulic Ana 2.1.4.1 FSARAases
-FSAR analyses assumptions are developed to demonstrate compliance with curreni NRC regulations concerning fuel -design limits, pressure boundaryprotection (overpressure protection), and radiological releases. These assumptions do n ily result in the most severe overcooling. The analyses are typicall carried out for only a few minutes and do not provide enough data to perform vessel integrity fracture analyses.
- 2.1.4.2 WCAP-10019
_Vessel Integrity.Analyses The analyses provided in WCAP-10019 are typical of FSAR-type design bases events.
However, the boundary conditions have been selected to enhance the overcooling.
Maximum safety injection and feedwater flows are assumed, mini water temperatures are used, and heatsources are either omitted or are cons atively underestimated. Large and small LOCAs have been addressed, as well large and small steam line breaks.
In. addition, the Ranch Seco overcooling 04/14/82
.2-5 ROBINSON SER INPUT S
event was included. Westinghouse indicates that the dynamics of this eyent nwould besimilarto a
low probability small steam line break (includingaddi ctioa s
)i exest o ctionAP1009identified for two eventsprese line break haebenpefrmdb
.WA-10019. Fo h slatable LOCA (a stuckopen IORV)J, itis assumed that th e operator autedtalybreak n 30 mi nutes. For the-1arg e steam ine break, Si eth a
ilary feedwat r to the faulted steamgeneratdrand
- .n~upirecti'f nflowto atheRCS is termi nated within10 minutes.,
SWstinghoue ProcedUralGuideline Analyses n
response to-item 1.C.-Lof theTHIAction Plan, Westinghouse hasperformed a er ostimate"analyses to support their current program for operator
-gui del i nes anid procedure development.
These analyses indicate-that considerable T-conservatism exists in the WCAP-10019 vessel integrity analyses.
2.1.4'4 NRC Independent Audit
- Analyses, Ihdependentaudit analyses of alarge steam linebreak have been performed by
- LAN L with,the-TRAC.PD2-computer programs.
These analyses are in agreement with otheoWestinghouserguideline analyses.uo Independent audit analyses are also being performed at INELewith the:RELAP5 Computer program. f or ceua r s
esteam line breaks. The results of these analyses.
willbe available at the. end of April 1982.
2.e aditria for hocedural Reviews The procedures to be reviewed were selected based on the perceivd likelihood of conditions occurring that might subject the reactor vessel to prressuirized thermal shock conditions and based on the potential consequences of less likel:
Such procedures selected included normal heatup and cooldowli, steam generator tube rupture, steam line breaks, and loss of coolant accidents.
The audit criteria for the content of procedures was somewhat flexible to account for the operato'r, knowledge itterface andto identify which procedures must be used to respond to a certain-transient. In addition, detailed operat 04/14/82 2-6 ROBINSON SER INPUT. SEC
'knowledge of actions for preventing or mitigating PTS could offset some weak nesses in procedures.
With this in mind, the following criteria were established for the procedures audit:
S
- 1)
Procedures hould ot instru ctopetos to take actions that would violate DT mits (2) Pfocedure should provide guidance on recovering from transient or accident conditions without violating NDT or saturation limits.
(3) Procedures should provide guidance on recovering from PTS conditions.
(4)
PTS procedural guidance should have a supporting technical basis.
(5) High pressure injection and charging system operating instructions should reflect a consiieration for PTS.
(6) Feedwater and/or auxiliary feedwater operating instructions should reflect PTS concerns.
(7)
- An NDT curve and saturation curve should be provided in the optrol room.
{Appendix G limits for cooldowns not exceeding 100oF/hr).
2.3 In-Plant Training Program The effort of the task force to determine the effectiveness of CP&L training i PTS began by developing training criteria which,would be used in evaluating tb training material, interviewing Robinson 2 shift personnel; and assessing the evaluation CP&L made after completion of the training. The criteria developed into three general areas:
(1) Training should incldde specific instruction on NDT vessel limits for NORMAL modes of operation.
(2) Training should include specific instruction on NDT vessel limits for transients and accidents.
4/14/82
.2-7 ROBINSON SER INPUT SEC
(3) Training should poticularly emphasize those eve*
known to require operator response to mitigate PTS.
More specific criteria were also developed to aid in the. review of the training program and in preparation of interviews with operating personnel.
CP9 ftann program on riS ani agw wa requestedto iu h an outli t e ain ng th lessonpian ich was s ed i th tinng cas e: 'They were also ques tioed'on the method used et 4valuate the iffidtiveness of the training sessions.
Preparation for review of the training program included a review of CP&L correspondence with the Commission, including a report on vessel integrity of Westinghouse operating plants (WCAP-10019), normal and emergency procedures furnished by Robinson 2, the Robinson 2 license, technical specifica tions, and the FSAR. An interview plan was developed which used the general training criteria and the specific subjects which were included in the CP&L training material.
Each interview was preceded by a discussion of the reason for the.audit, acknowledgement that the individual could use all material available in the
-control room, particularly the followup-or recovery steps irith emergency procedures, and a request that the individual not inform other operators of the questions asked-in-the interview -Several interview aids were.prepared to provide the -operators a point of reference for discussion and to allow them to predict responses or execute recovery strategies to mitigate PTS or challenges to other limits.
04/14/82.
2-8 ROBINSON SER INPUT SEC 2
A visit to the Robinson 2 site took place on April 5-7, 1982, during which time the Task Group evaluated procedures and training. The key findings of the group.are discussed in Section 3. In preparation for the Robinson 2 evaluation, the Task Force used the general criteria addressed.in.
Section 2.
1.2 Current Status of the Generic PTS Issue Efforts to pursue an integrated PTS program pvolving a variet yof technical areas are continuing under USI A-49.
The summ-er of 1983 is the current schedule for finalizing our generic regulatory requirements for PTS along with required corrective actions if the generic requirements are not met.
Key issues are yet to be resolved and extensive'.programs exist to provide the foundation for the generic regulatory requirements.
Before the-above effort resulting in regulatory requirements is completed, however, we have committed to the.Commission to.have developed an interim initial position for the summer of 1982 (June). The interim initial position will consist of NRC evaluation of the safety of continued plant operation (and initial corrective actions required) for the eight plants previously identified as representative of plants having the highest RTNDT.
Technical assistance is being provided by a PNL multi-disciplinary team.
PNL has been contracted to work with thestaff to provide recommendations regarding the June..1982 initial position on the safety of continued operation and to recommend any additional corrective actions that PNL believes should be initiated before the NRC generic resolution and acceptance criteria are adopted.
The June recommendations by the NRC staff to the Commission.will also consider.the findings and recommendations addressed in Sections 3 and 5 of this report, as well as other Task Forces formed for related investigations (such as fluence reduction at the vessel wall).
1.3 Robinson 2 Configuration Robinson 2 is a three-loop Westinghouse PWR rated at 2200 MWt (700 MWe).
Normal pressurizer level is controlled by the chemical and volume control system which contains three positive displacement pumps. The safety injection system (SI) utilizes three high head pumps which will initially discharge the boron injection tank (BIT) into the cold legs of the reactor coolant system.
04/14/82
.1-2 ROBINSON SER INPUT SEC 1
fe The SI pumps have a shut-off head of 1500 psig and have a rating of 375 gpm at, 1080 psig. The SI system alsB contains three accumulators which discharge at 600psig and two low head pumps (RHR) rated at 3000 gpm at 115 psig.
The Robinson 2 control room utilizes an L-shaped bench board which contains standard three-loop information.'
However, most-of the meters and controls are considerably s maller than other three-or four-loop control boards.
The most important plant parameters are displayed on recorders which have a normal speed of 1 in/hr.
The following table contains the. major parameters available to an operator at Robinson 2 which would assist in monitoring PTS events:
Parameters Display RCS pressure Wide-and narrow-range meters (1) wide-and narrow-range recorder RCS temperature Wide-range THOT and TCOLD meters and recorders for all three loops Rx head thermocouple (1) meter - will be functional after refueling
- Core exit thermocouples Normal (W) display using toggle switches Single T/C display
- Subcooling monitor Utilizes "pressure to saturation" meter May select core T/C or wide-range RTD Push-button selection may display *F subcooled-in a digital form" May select Loop AT.
04/14/82 1-3 ROBINSON SER INPUT SEC 1
- 2. SHORTTERM CRITERIA USED FOR ROBIHSOH AUDIT...
2.oT rsient andus ooaYso AuT
- 2.
-as en and
-:ci
~
ntroduction 2f s e& -----------
(exce ay occur as a result oss.
orakt acces dressem sflo )
d e fr mt ma ry i so r s
coolant a cci oft h poveasesbfardtcrg csornare or enat ing phchrepsredtral sock t h e r m a r
- y r a u l m a n ae s v eely s e sif g
T h s e ti eve s a o ionscse ee n d sinc es mandunctonlsoso Section 3.1 provides our comentsn attil fal r sa/ra alo chnurboiofth zatin staf 1 e These ventt condihions were eet.
900 psriclrcn'cra eig.
T s secndar the ~ prm rCSs runnsincngh haargingoldo enrors a d r s e a nl
- o vr.e B. RobisoiEensso n
ta Sae fr avte prine a reak Api 28, 1970akFrc o ui OnApil2, a97,n ing ot funiedSction testing (no f uel aed) on te.o h
stea genrato sa eY ava al conetons fvalied due. ~ rodn. A30 The op leas connditond Rhr 0 900 psi seonar Se-o 31 RPSrning o As a reult of he 6-inischedue 80 pipe beakanwihodeyhatte plant ooled own 213 0Fin
.hu.ta32 0
codlgtneate Teprt 04/14/82nso O
- R B HSvSE HUT E i
t..
B.
R o b i n s o n*
tripped the RCPs (30 seconds) and started the remaning two coolan chargingpumps-70 seconds). Th minimum primary system pressure was.1880 psi occurred.- Theplant was recovered to a normal nsl ad conditi n of e2O5O psig and charging/letdown reestablished prior tQshutdown.
A post~event review-of-the data indicated that the 'pressuriter. surge line 'ld not empty. --A base-case analysis was performed for the event. In additiori, a sensitivity analysis was performed without RCP trip, with only one.charging apump, and with a primary heat source. The analysis-showed that the pressurizer~
would drain and the primary system pressure would-fall below the SI setpoint in about 3 minutes. The cooldown was less and the pressures were lower than the base case analysis. It is expected that the operator actions, based on current procedures, would be similar to this sensitivity analysis.
The safety valve stand-off piping.was redesigned to prevent any similar occurrences.
2.1.2.2 Reactor Coolant Pump (RCP) Seal Failure Event, May 1, 1975 During full-power operation, RCP "C" seal 1 leakage exceeded the 'echnical spei fication limit of 6 gpm.
A load reduction was.commenced at a rate of 10%
per minute to 36% power and pump "C was deenergized.
Reactor trip occurred due to a turbine trip resulting from the load reduction.
The decision was made to restart pump "C". when seal injection could not be restored to pumps "A". and B."
-Shortly after restarting the pump., while at 1700 psig and 4800F, seals 2 and 3 failed on pump "C" and the pressurizer level began to decrease.
The following chronology is provided:
2300 -
RC system at 1700 psig, 480*F RCP "C" running 0015 -
Stop RCP,"
on high standpipe level alarm Pressurizer level fallint rapidly due to seal 2 and seal 3 failure 4114182
. 2-2 ROBINSON SER-INPUT SEC 2
0016 - SI pump "A"l manually started. to supplement charging flow (injection to hot leg),.
0018 - SI pumps "B" and "C" manually started, pressurizer level stops falling a 00)
SI pump due to spressurizer e e 0038 - Dvr tag~gfo.01it-6 rtr 03 lat6n (50 psi g :at this 0048-SI accumuiators p~rpio
.J nlati{
5OO r-sg at -thi time) 7 The coodown for this event was from 50F to approximately 3100 F in one-half hour, with the pressuredecreasing from 1700 psig to about 15 sgoe h horiod of teres. rhede of the auxiliary pressurizer spray rapidly reduced the pressure to 500 psig.
The operator used SI to stabilize pressurizer level and pressure while using the mainacondenser to cool down the plant for RHR entry.
There is no indication that SI was used to repressurize the plant.
2..1.2.3 Stuck OpenSteam Generator Relief Valve
- Eve, oer 19 While at nominal full-power operating conditions, theo operator was usingwste
- -eneato reief~Vlve~~t prvid RC tmperature control. One valve would nteraor, resulting in the equivalent of a small steam line break. The seclo side rlowdown resulted in a reactor trip and safety injection. -The oyerall cooldownrate was 1570F over a 2-hour period, to 3890F, during the course of the event. Insufficient information is currently available to address operator actions taken during this event.
-3 ROBINSON SER INPUT SEC 2 04/14/82