ML20064L271
| ML20064L271 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Robinson |
| Issue date: | 04/15/1982 |
| From: | Mazetis G NRC |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20064E577 | List: |
| References | |
| FOIA-82-389, TASK-A-49, TASK-OR NUDOCS 8205050480 | |
| Download: ML20064L271 (38) | |
Text
d u-7, g
APR 151982 MEMORANDUM FOR:
H. L. Thompson, Jr., Acting Director, Division of Human Factors Safety FRGl:
G. R. Mazetis, Chairman, Robinson PTS Task Force
SUBJECT:
ROBINSON 2 SHORT TERM TASK FORCE ON PRESSURIZED THERMA 1. SHOCK (PTS)
Your menorandte dated March 16, 1982 appointed a Task-Force to make a detailed review and prepare a report on the status of efforts on PTS.at the H. B. Robinson Nuclear Plant. A site visit was arranged on April 5-7th, during which time audits were conducted on procedures and training, specifically with regard to. PTS. The Task Force msnbers consisted of the following individuals:
Gerald R. Mazetis - DSI/RSB - Chainnan H. Brent Clayton - DHFS/PTRB Joseph J. Bury - DHFS/LQB Edward Throm - DSI/RSB Raymond Klecker - DE/MTEB Roy Woods - DST /GIB (ex-officio)
Our instructions were to provide a report in 30 cays (April 15, 1982) character-tzing the problan(s), methodology of resolution, bases for conclusions', and recom-mendations regarding the adequacy of in-place training programs and opeiating pro-cedures. In addition, you requested that the report attspt to characterize the applicability of this effort to other like facilities and propose review schedules and triteria that can be used in reviewing the other facilities of special concern.
The enclosed evaluation provides the requested report. The site audit of training programs was corducted by Joseph J. Buzy. The site audit of procedures was con-ducted primarily by H. Brent Clayton. The evaluation of the overcooling history at Robinson 2 was performed by Edward Throm. Ray Klecker and Neil Randall, although not part of the on-site audits, contributed to the fracture mechanics assessment. Roy Woods, although also.not part of the site visit, assisted with the report to ensure consistency with other ongoing PTS programs.
j As indicated in Section 3.0, " Key Findings frts the Robinson Audit," it is clear #
that the control room energency procedures rs:ain weighted toward core cooling and do not go far enough in addressing the PTS issue. Pending generic resolution.
of TMI Action Plan Ites I.C.1, such a procedural shortecc:ing could have been tempered during our audit. of plant personnel by a strong awareness and knowledge
"~
of the PTS issue; however, the audit produced a varied response frca good to poor..
The reason for the' varied response is due in large part to the need for closer validation by CPal. of operators retention of the material covered in the classroom training sessions on. PTS.-A-more_ complete discussion of the interviews is presenteo in Section 3.3r-a'nd our-reccrunendatfons are addressed in Section S.C.
~
vaEOMT<.D
/
._. m c" n A
..Q.......J.
...............\\
......M!
.......z.C.......'................
f
-a >
DATE %
{
....,.............q...................
2
., H.,L. Thomps:n.
APR 15 1982 Based on your direction in the memorandum dated March 16, 1982, the enclosed report completes the Charter of the Robinson 2 PTS Task Group.
Original signed bys G. Mazetis, Chairman Robinson PTS Task Force
Enclosure:
As stated cc:
H. Denton (w/ enclosure)
E. Case J. P. O'Reilly DISTRIBUTION R. Pattson S. Hanauer D. Eisenhut R. Vollmer J. Zwolinski D. Zicmann T. Speis B. Sheron y
R. Woods T
R. Klecker E. Throm B. Clayton J. Clifford
~
J. Buzy L. Defferding (PNL)
T. Novak S. Varga B. Requa S. Weise (Robinson IE Resident)'(2 copies)
J. Laaksonen R. C. Lewis (IE)
Docket Files RSB R/F GMazetis R/F r-N/O W
TW Dii@
ric q.
.g................................................................................................................
......[.........
.... H..T..h o.,p p s..o. n Quis:d1 m }.
...../.... 8 2 4/W82 4/
,nj
.. ~ - - -
n.
~
a --
s.
f
- .~.. =... -. z =.. s....,,
.c..,, -. a.
1.. J
.a
.s.
.i ENCLOSURE
...., ~ c. 3 z, s.n '.... =. r.'. i.c " i d.. i. c,* ~.0
=C c.,>=.c
.=.-.-.t..=.'..=..-
~'
Ic
...'. : *~.
.. 1 w.e *w.
.3
.. e.. :.
.J -, 3.
I
.r...,t. <, n. j :..,. s
.t.
.t
....a
...........so..
64.3.m..:...
- e.....a
.s.
...a.
....s NRC' STAFF AUDIT OF ROBINSON 2 PROCEDURES AND-TRAINING FOR-. -
.PR SSUR
- H L. SHOCK.
..a
.a
...s..
2 2.:..
ca....
==
=2....
...w...
Task Force Chairman:
G. Mazetis
.=.
Y e
a e-04/15/82 ROBI.NSON SER INPUT TC
. ~.
NRC Staff Audit of Robinson
- 2 Pro'cedures
~
and Training for Pressurized
~
Thermal Shock CONTENTS Page 1
INTRODUCTION........................................................
1-1 1.1 Short-Term Effort Objectives and Scope of Review...............
1-1 1.2 Current Status of Generic PTS Issue............................
1-2 1.3 Robinson 2 Configuration.......................................
1-2 2
SHORT-TERM CRITERI A USED FOR ROBINSON 2 AUDIT.......................
2-1 2.1 Transient and Accident Analyses................................
2-1 2.2 Criteria for Procedural Reviews................................
2-6 2.3 In-Plant Training Program......................................
2-7 3
KEY FINDINGS FROM THE ROBINSON 2 AUDIT..............................
3-1 3.1 Transient and Accident Analyses................................
3-1 3.2 Procedures.....................................................
3-3 3-6 3.3 Training.......................................................
3-10 3.4 Summary........................................................
4 FRACTURE MECHANICS..................................................
4-1 4.1 Genera1........................................................
4-1 4.2 Robinson 2 Fracture Mechanics..................................
4-1 5-1 5
RECOMMENDATIONS.....................................................
6 APPLICABILITY TO REMAINING SEVEN PWRS................
6-1
== -
D4/15/82 iii ROBINSON SER INPUT TC
'T O'
~
1 INTRODUCTION 1.1 Short-Term Objectives and Scope of Review On March 16, 1982, an interdisciplinary Task Force was established to evaluate certain aspects of the Pressurized Thermal Shock (PTS) issue for Robinson 2.
The question that the Robinson Task Force focused on was:
ARE CORRECTIVE ACTIONS REQUIRED THAT MUST BE INITIATED BEFORE THE LONGER TERM PTS PROGRAM PROVIDES GENERIC RESOLUTION AND ACCEPTANCE CRITERIA?
Emergency procedures and operator training were the only areas in whkch the Robinson Task Force applied the above general question.
As noted in the NRR March 9,1982 presentation to the Commission:
...we will undertake a program to verify that existing operating procedures contain the steps necessary to prevent and/or mitigate PTS events, and to verify that operator education / training programs
.regarding PTS are acceptably thorough."
Initial informal contacts were made with CP&L the week of March 15th and, during a conference call on March 19th, the details of our expected review areas were discussed. Also discussed was a planned visit to the site.
With the 1 imitation of a 30-day response, the scope of review ha,d to be narrowed so that meaningful conclusions and recommendations could be produced.
Therefo're, resolution to the varied technical questions on PTS (thermal-hydraulic analyses, fracture mechanics, probabilities) was not part of the Task Force charter.
- Also, implementation of any recommendations (see Section 5) is subject to coordination
=r -
and consistency with the longer term generic program (USI A-49).
04/14/82 1-1 ROBINSON.SER INPUT SEC 1
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.
CurrentStatusoftheGenericPTS[ssue 1.2 Efforts to pursue an integrated PTS program involving a variety of technical areas are continuing under USI A-49.
The sumer of 1983 is the currewe schedule for finalizing our generic regulatory requirements for PTS alo'ng 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 ^
Technical assistance is as representative of plants having the highest RTHDT.
PNL has been contiacted to being provided by'a PNL multi-disciplinary team.
work with the staff to provide recomendations 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 reduc' tion at the vessel, wall).
1.3 Robinson 2 Configuration 2
Robinson 2 is a three-loop Westinghouse PWR rated at 2200 Mdt (700 M4e).
Normal pressurizer level is controlled by the chemical and volume control The safety injection system which contains three positive displacement pumps.
system (SI) utilizes three high head penps which will initially discharge the boron injection tank (BIT) into the cold legs of the reactor coolant system.
ROBINSON SER INPUT SEC 1 1-2 04/14/82
i 2 SHORT-TERM CRITERIA USED FOR ROBINSON AUDIT e
2.1 Transient and Accident Analyses _
~
2.1.1 Introduction Overcooling events in PWRs may occur as a result of steam line breaks (exces-sive steam flow), feedwater system malfunctions, or loss-of-coolant accidents.
Multiple failures and/or operator errors can result in more severe overcooling' Of particular concern are those events in which repressurization o.f events.
This section t
the primary system occurs following the severe overcooling.
addresses an overview of Robinson 2 overcooling events which occurred since the
~ -
Aside from the primary mission of the Task Force to audit plant was built..
procedures and training, also provided (Section 2.1.4) is a sumary of the thermal-hydraulic analyses ava'ilable for evaluating pressurized thermal shock events.
Section 3.1 provides our comments and conclusions on these events' and analyse 2.1.2 H. B. Robinson Overcooling Events Summary 2.1.2.1 Steam Safety Valve 1.ine, Break, April 28, 1970 during hot functional testing (no fuel inaded), one of the On April 28, 1970, A 360*
steam generator safety valv'e connections failed due to overloading.
circumferential break allowed the safety valve to blow off the main steam line The plant conditions were:
533*F, 2225 psi primary 900 psi' secondary -
~
3 RCPs running / letdown 45 gpm charging no fee 6 tater to the steam generators As a result of the 6-in. schedule 80' pipe break, and with no decay heat, the The operator plant cooled down 213*F in I hour, to a 320*F cold leg temerature.
ROBINSON SER INPUT SEC 2 2-1 04/14/82 i
l, imediately tripped the RCPs (30 seconds) and started the remaining two coolant charging pumps (70 seconds). The minimum primary system pressure was 1880 psi; with the safety injection (SI) setpoint at 1715 psi.,.no safety infection
~
occurred. The plant was recovered to a nomal no-load condition of 2050 psig and charging / letdown reestablished prior to shutdown.
A post event review of the data indicated that the pressurizer surge line did not empty.
A base case analysis was perfomed for the event.
In addition, a sensitivity analysis was perfomed without RCP t' rip, with only one charging pump,,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 redesignsd to prevent any similar occurrences.
?
2.1.2.2 Reactor Coolant Pumo (RCP) Seal Failure Event, May 1,1975 During full power operation, RCP "C" seal 1 leakage exceeded the technical i
specification limit of 6 gpm.
A load reduction was comenced at a' rate of 10%
per' minute to 35% 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 Shortly after restarting the pump, while at 1700 psig and 480 F, seals 2 j
"B."
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 "C," on high standpipe level alam Pressurizer level falling rapidly due to seal 2 and seal 3' failure ee i
.. e o
0016 - SI pu=p "A" manually started.to supplement charging flow (injection l
l
~
to hot leg) 0018 - SI pumps "B" and "C" manually started, pressurizer level stops falling i
0036 - Divert charging flow from "B" loop to auxi,liary pressurizer spray to reduce pressure (1150 psig at this time, coola'nt temperature below
~ ' '
400*F) 0039 - Stop SI pump "C" due to risihg pressurizer level 0048 - SI accumulators partially inject prior to isolation (500 psig at this time)
The cooldown for this event was from 450*F to approximately 310*F i I
hour,' with the. pressure decreasing from 1700 p period of interest.
the pressure to 500 psig.
~ ~
The operator used SI to stabilize pressurizer level and pressure w the main condenser.to cool down the, plant for RHR entry.
There is_no indication that SI was used to repressurize the plant.
2 Stuck Open Steam Generator Relief Valve Event,_ November 5, 197 2.1.2.3 While at nominal full-power operating conditions, the operator was us One valve would generator relief valves to provide RCS temperature control.
The not reclose, resulting in the equivalent of a small steam line. break.
The secondary side blowdown resulted in a reactor trip and safety inject overall cooldown rate was 157*F over a 2-hour period, to 389 F, du Insufficient information is currently available to course of the event.
address operator actions taken during this event.
I 5?
e s
t me ROBINSON SER INPUT SEC 2-3 04/14/82 i
b 2.1.3, H. B. Robinson Temination Criteria 2.1.3.1 Reactor Coolant Pumps (RCPs)
The RCPs are tripped when the primary system pressure falls to 1300 psig.
Ir.
addition, the RCPs are tripped if sea'1 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 flow ra'te is
.... limited to 4.00 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 themal shock is provided. The termina-J tion criteria include a 2000 psig (and increasing) requirement.
2.1.3.4 SI Termination During Steam Line Break The temination criteria for safety infection during a steam line break are:
One RCS T less than 460*F, H0T RCS pressure greater than 700 psig (stable or increasing),'
Pressurizer level greater than 20% (heaters covered),
RCS subcooling greater than 40*F, and Heat sink available (U-tubes covered).
f As shown, one of the criteria for teminating SI during a steam line break is reading less than 460F, with wide-range primary coolant one wide-range THOT m
2.4.
ROBINSO,N SER INPUT SEC 2
7
~
system pressure greater than 700 psig and stable or increasing. The Westing.
This value includes all uncertainties house guideline value is 350*F, THOT.
and does imply reference to the downcomer temperature.
The uncertainties include core heatup during natural circulation,-ECC mixing
~
and instrument errors. Westinghouse has reviewed their fracture data for a wide range of transients and, for the most limiting vesse'l at end of life, they w uld not result in vessel failure.
The 700 psig, conclude that the 350*F THOT stable or increasing, pressure assures that a primary side LOCA 'does not exist coincident with the steam line break.
Robinson 2 has increased the 350*F value to 460*F to provide a combined assurance that 40*F subcooling exists at a pressure of 700 psig, concurrent with a sufficiently high temperature to accommodate brittle fra'cture concerns.
Also, it is noted that the Westinghouse 350 F/700 psig values would violate the Robinson 2 NDT limit for 100*F/hr
'~
cooldown events..
2.1.4 Thermal-Hydraulic Analyses 2.1.4.1 FSAR Analyses FSAP. analyses assumptions are developed to demonstrate compliance with current NRC regulations concerning fuel design limits, pressure bcundary pr'otection These assumptions do r)ot (overpressure protection), and radiological releases.
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-10019 Va.:sel Integrity Analyses The analyses provided in WCAP-10019 are typical of FSAR-type design bases However, the bo,undary conditions have been selected to enhance the i
events.
Maximum safety injection and feedwater flows are assumed, minimum
=.
~
overcooling.
j water temperatures are used, and heat sources are either omitted or are conserv-atively underestimated.
Large and small LOCAs have been addressed, as well as large and small steam line brlaks.
In addition, the Ranch Seco overcooling 2-5 ROBINSON SER INPLIT SEC 2 04/14/82
p event was included. Westinghouse indicates thati the dynamics of this. event,
would be similar to a low probability small steam line break (including addi-j, Operator action is identified for two events presented in tional failures).
For the isolatable LOCA (a stuck open PORV), it is assumed that WCAP-10019.
For the large steam line break, the operator isolated the break in 30 minutes.
it is assum ed that auxiliary feedwater to the faulted steam generator and makeup injec' tion flow to the RCS is teminated within 10 minutes.
2.1.4.3 Westinohouse Procedural Guideline Analyses _
In response to Item I.C.1'of the TMI Action Plan, Westinghouse has perfom i
series of "best-estimate". analyses to support their current program for operato These analyses indicate that considerable guidelines. and procedure development.
conservatism exists in the WCAP-10019 vessel integrity analyses.
2.1.4.4 NRC Indeoendent Audit Analyses Independent audit analyses of a large steam line break have been per These analyses are in agreement with LANL with the TRAC-PD2 computer programs.
the Westinghouse guideline analyses.
Independent audit analyses are also being perfomed at INEL with th The' results of these analyses
. computer' program for small steam line breaks.
will be available at the end of April 1982.
- 2. 2 Criteria for Procedural Reviews The procedures to be reviewed were selected based on the perc of conditions occurring that might subject the reactor vessel to pressu themal shock conditions and based on the potential consequences Such proc,edures selected included normal heatup an'd cool transients.
steam line breaks, and loss of coolant,
e-steam generator tube rupture, accidents.
The audit criteria for the content of procedures was somewhat flexible account for the operator knowledge interface and to identify which.
In addition, detailed operator' cust be used to respond to a cer+ gin transient.
ROBINSON SER INPlJT,SEC 2 2.6 S MSVW29
a 4
knowledge of actions for preventing.or mitigating PTS could offset some we'ak-nesses in procedures.
Wit'h this in mind, the following criteria were e'stablished for the procedures audit:
~
'~
(1) Procedures should not instruct operators "to f.ake actions that would
~
~
violate NOT limits.
(2) Procedures 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.
I (4) PTS procedural guidance should have a supporting technical basis.
High pressure injection and charging system operating instructions should (5) reflect a consideration for PTS.
^
^
Feedwater and/or auxiliary feedwater operating instructions should reflect (6)
PTS concerns.
An HDT curve and saturation curve should be provided in the control room.
(7)
(Appendix G limits for cooldowns not exceeding 100 F/hr).
2.3 In-hant Trainino Program The effort of the task force to determine the effectiveness of CP&L training in PTS began by developing training criteria which would be used in evaluating the training material, interviewing Robinson 2 shift personnel; and assessing e
The criteria the evaluation CP&L made after completion of the training.
developed into three general areas:
Training should incidde specific instruction on HDT vessel limits for
"~
^ (1)
~
l NORMAL modes of operation.
I Training should include specific instruction on HDT vessel limits for (2) transients and accidents.
ROBINSON SER INPUT SEC 2 2-7 04/14/82
.-,-----,9-
. (3) Training should particularly emphasize those events known to require operator response to mitigate PTS.
More specific criteria were also developed to aid i'n the review of the training program and in preparation of interviews with, operating personne1.
"P&L was requested to furnish an outline of their training program on PTS and the lesson plan which was used in the training classes.
They were also ques-tioned on the method used to evaluate the effectiveness 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 plan'ts (WCAP-10019), normal and emergency procedures furnished by Robinson 2, th'e 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 in the 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.
9 Y
7*
J 04/14/82 2-8 ROBINSON SER INPUT SEC 2 l
t l
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.
I 3.1.2 Robinson 2 Overcoolina' Events
~
CP&L reviewed the Robinson 2 operating history and presented three events where the cooldown rate exceeded 100'F 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 driteria (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 avercooling event, particularly if the pressuriz,er 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 i
following a severe overcooling event.
CP&L and Westinghouse believe that i
repressurization to 2000 psig will not compromise vessel integrity.
t l
04/09/82 3-1 ROBINSON SER INPUT SEC 3 l
3.1.3 Robinson 2 Teminai. ion Criteria 3.1. 3.1 SI Temination Durinq LOCA The termination criteria for safety, injection during a LOCA are:
RCS pressure greater than 2000 psig and increasing, Pressurizer level at no-load level and responding, Heat sink available (U-tubes covered), and RCS subcooled at least 40*F.
These criteria are weighted to core cooling concerns, and do not explicitly address the pressurized themal shock issue. The licensee h'as indicated that, based on the Vestinghouse anal,yses under review by the staff, no PTS concerns exist during a LOCA.
A.
One of the criteria for termination of SI during a LOCA is that the primary This value provides the coolant system pressure is 2000 psig and increasing.
following information:
The br,eak has been isolated, or the SI flow is equal to or greater than (1) the break flow.
1 Some margin exists to terminate SI before the PORV would be challenged.
(2) e Repressurization to 2000 psig further assures a 40*F subcooling margin, l
l (3) including uncertainties.
At the time the emergency procedure was developed, Robinson did not have the To subcooling meter installed, and core cooling was the dominating issue.
- ~
verify subcooling, and include uncertainties in instrument readings and flow (It is noted' conditions, a primary system pressure of 2000 psig was adopted.
that the Robinson high head safety injection pump cut-off head is 1500 psia.)
i i
~
I ROSINSON SER INPUT SEC 3 3-2 i
'"'*^A**
a SI Termination Criteria During Steam Line Breaks
'3.1.3.2 The termination criteria for safety injection during a steam.line break, as presented in Section 2.1.3.4, address the pressurized thermal shock issue by a The change to the LOCA criteria discussed in the p' receding Section 3.1.3.1.
Therefore, criteria reduces the pressure at which SI termination is allowed.
we conclude that these criteria provide a reasonable balance between core
~
cooling and PTS c'oncerns.
3.1.4 Thermal-Hydraulic Analyses FSAR design bases analyses are not suited to the evaluation of vessel integr,ity.
The events Insufficient carryout in time. exists to perfonn fracture analyses.
.. presented in WCAP-10019 are bounding overcooling events, and are represe of. design bases events (single failure). These analyses are suitable for vessel integrity studies. Analyses performed by Westinghouse, using "best-estimate" assumptions, indicate that considerable conservatism exists in the These best-estimate analyses indicate that:-the WCAP-10019 calculations.
While cooldown would not be less than 350*F for the steam line break spectrum.
some uncertainties exist with regard to mixing for small-b'reak LOCAs, the loss of RCS inventory events appear to be bounded by the steam line break spectrum.
The NRC independent audit therr.al-hydraulic calculations for the large steam l
line break addr'essed in Section 2'.i.4.4 support' the above observation on the Additional audit calculations to'be performed during Westinghouse analyses.
April are expected to provide further confirmatinn of the Westinghouse t hydraulic analyses.
3.2 Procedures _
"~
3.2.1 Description of the Audit Our audit included a review of procedures selected as discussed in Sectio discussions with licensee and Westinghouse representatives on the instru relating to PTS and their bases, and an audit of the control room copy o Our audit included the procedures to determine their legibility and currency.
ROBINSON SER IHPUT SEC 3 3-3 04/09/82
..,---..--.--s
-e--
y
- d_
' following ~ Emergency In'structions (EIs), 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) Turbi11e and G'enerator Trip (Part B)
AP-19 Malfunction of RCS Pressure Control System AP-24 Loss of Instrument Bus AP-25 Spurious Safeguards Actuation GP-2 Heatup (Cold Solid to Hot Subtritical at No-Load TAVG)
GP-3B Reactor Trip Recovery GP-5 Shutdown (Normal Plant Shutdown From Power' Operations to Hot Shutdown Conditions)
GP-5A Plant Temperature and Pressure Control Using Natural Circulation GP-6 Cooldown (Plant Cooldown From Hot Shutdown to Cold Shutdown Conditions) 3.2.2 Comoarison 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; 3
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.
i the procedural instructions may violate these limits (even though 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 ingtructi'ons for pressure control or operation of the charging pumps until a controll' d cooldown/,
e depressurization is begun using GP-6.
Discussions with Westinghouse representatives indicated that the SI temination criteria are under review as part of the generic procedural guideline development and it is anticipated that thsy will be changed to a lower pressure, at least for,
- ~
the plants having intemediate head SI pu=ps like Robinson 2.
i I
4
~
04/09/82 3-4 ROSINSON.SER INPlIT SEC 3
V i
\\
Procedures sha11 provide guidance on recovering from transient or 1
(2)
See it'em
,I accident conditions without violating NDT or saturation limits.
(1) above for discussion on NDT limits. The procedure for depressuri-
, zation events (EI-1) re(ers the operator to Curve 3.5.and provides
~
If reactor coolait' pumps inst,ructionto maintain at least 40*F subcooling.
i-are tripped, the procedure for natural circulation instruc,ts the operator to maintain at least 50*F subcooling. The procedures do not provide a
,j I l maximum subcooling limit.
Curve 3.5 is a pressure-temperature plot The recovery showing a saturation curve and a 40'F subcocied curve.
instructions for a secondary coolant rupture instruct the operator to 1
l j
establish steam dump from the " good" steam generators to stablize temperatures when temperature and pressure start to increase following I
I dryout of the faulted steam generator.
Procedures should provide guidance on recovering from. PTS conditions.
- (3)
While the procedures provide instructions for maintaining the RCS w'ithin conditions allowed by th6'NDT curves, it is not apparent that the pro-cedures recognize that some transients or accidents may result in PTS conditions at the time that the operator can begin to control plant 3
l There are no explicit instructions to the operator on 'how to j
conditions.
recover from PTS conditions. However, terminating feedwater flow to the faulted steam generator and the SI termination criteria help to limit PTS'
~
following a steam line break.
PTS procedural guidance should have a succorting technical basis (4)
The procedural guidance is generally consistent with that provided by the
~
l These guide-
';1
' Westinghouse Owners' Group emergency procedure guidelines.
The a~ctions lines are based on best-estimate analyses of transients.
o e
'i specified in the guidelines which would ic: pact PTS are also c'onsistent with the bounding analyses, presented in WCAP-10019.
Westinghouse repre-sentatives stated that thg guidelines are also being reviewed against 1
=-
best-estimate fracture mechanics analyses and that this effort will be '
,l See Sections ~ 2.1 and 3.1 for a discussion of the s
j completed in May 1982.
safety injection termination criteria.
4 r.
b a
04/09/82 3-5 ROBINSON SER INPUT SEC 3
_ _ }
l (5) Hioh pressure infection 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 -
t with a cool vessel. There are no specific instructions for operation of l
the charging pumps following the~depressurization transients.
i (6) Feedwater (FW)' and/or auxiliary feedwater (AFW) operating instructions should reflect PTS concerns.
Instructions are provided in the steam generator tube rupture and the loss-of-coolant accident procedures to 4
terminate FW/AFW flow to the faulted steam generator. These and other
~
l J
procedures provide instructions to maintain steam generator levels in the good steam generators within a define' bandi
~
d l
l (7) An NDT curve and a saturation curve should be orovided in the control room..
These curves are provided in the Curve Book lo~cated in the control room and are referenced in the applicable procedures.
Each of these curves is
,.e on a pressure-temperature plot.
Curves 3.3 and 3.4 show the technical specification heatup and cooldown limits.
Curve 3.5 shows'the. saturation curve-and a 40*F subcooled curve.
i.
The control room copy of the procedures and curves that we audited was 1egible
~
and current.'
i 3.3 Training Introductio'n
~
3.3.1 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 50s were Shift Foremen), and two Shift Technical Advisces.
e 9
04/09/82 3-6 ROBINSON SER INPUT SEC 3.
_._]
3'.3.2 Commison of training 'with the Audit Criteria Training should include specific instruction on HDT vessel limits for (1)
All senior operators (50s) and Shift Technical NORMAL modes of operation.
Advisors (STAS) were aware of NDT vessel limits and the bases w
The 50s exhibited a good knowledge plant heatup and cooldown. restrict-ions'.
in the use of plant procedures, control board indications and controls,-
Recent classroom training had re-emphasized 'the and vessel limit curves.'
Both STAS lacked a familiarity with control reason for these limits.
board indications and controls.
. Training should include.soecific instructions on NDT vessel limits for.
(2) transients and accid'ents.
Training was conducted to emphasize concerns of vessel limits during transients and accidents, however, the training'
~ '
The training included discussions -
was limited to classroom'. instruction.
Four of the termination criteria for LOCA and.steamline break accidents of five 50s and'one of the-two STAS were familiar with PTS concerns
~
during accidents.' One of the' STAS had not attended the classroom training.
Training should particularly emchasize those events known to reouire (3)
Classroom training included actions coerator response to mitigate PTS.
required by the operators to mitigate PTS events; however, no traini was conducted in the control room, nor were past events at Robinson 2 In addition, training did not include dis'cussions of reviewed in de' tail.
events in which a steam bubble could develop,,in the RCS (other than pressurizer),.nor the potential for competing concerns'in the ste break procedure between attempting to control RCS temperature and.
pressure while not worsening the cooldown.
Three of the five 50s had recent simulator training and recalled that they could adequate 1y control RCS pressure and temperature du The other two 50s did not recall the details of previous line break.
It was recognized that there was steamline break simulator exercises.
ROBiNSONSERINPLITSEC3 3-7 04/09/82
limited instrumentation (wide range pressure recorder) to alert.the operator of rates of pressure rise during the steam line break recovery 3.3.2 Personnel Interviews The initial interviews with two Senior' Operators (50) indicated an excelle background of vessel pressure / temperature NDT (P/T) limits and basis curves, in addition to a good knowledge of PTS concerns and how pl They exhibited an excellent knowledge of tions could lead to PTS events.
control room instr,uments and equipment controls.' During the PTS eve cussion, which included single-and two phase flow in addition to a rea i
vessel steam bubble, they were able to follow procedures and predict po of the recovery procedure which would challenge P/T liinits.
One of these two 50s was concerned with the operator's ability to an He recognized that the wide-re rapid rate of pressure change using meters.
recorder was the only instrument which could display the past and pr The other-transient, and adequately depict any rapid rate of increase.
He remembec operator had recently trained at the Shearon Harris Simul,ator.
the team's concern on core subcooling limits during steam line bre d rg events and that they could edequately control. safety injection (SI)*an The other 50 di@
RCS temperature and pressure rise by use of steam dumps.
recall specific details of the last time he witnessed an SLB at the Both were concerned that a bubble in the reactor vessel head could control of pressure after termination of SI; however, they believed th control' secondary plant steaming to negate a rapid r, ate of primary temperature or pressure increase.
With regard to the interviews with two Shift Technical Advisors (ST had attended training in PTS and had a good understanding of react He was also aware of PTS concerns during limits during' normal operation.
He had difficulty identifying whic[
accidents and events leading to PTS.
in Si terminat temperatures to monitor for PTS (procedures identify T i
H0T is more of interest than THOV f
but concluded after discussions that T COLD d
l did have some problems identifying meters on the console, but
,~
He did not c'onsider possibility of steam bubble foms general location.
ROBINSON SER INPf ha
v Ld fter RC the reactor vessek head and the possibility of two phase conditions a He did not know the manual actions required for any reactor pumps are tripped.
iliary trip nor'did he find the procedural manual actions to terminate aux feedwater in the affected loop for a steam line break.
explicit.)
~
f He did not. appreciate that two steps (2.9 and 2.12) in the SLB pro ld concerning control of RCS temperature and pressure using steam d SI involve another cooling transient on the vessel, and could com Some difficulty locating SI flow and pump controls was e
~
termination criteria.He feels that his duty is to warn the Shift Forema does not believe he demonstrated.
may be violating procedure steps or exceeding limits, but in str'ategies is ready to contribute to any discussion of deviations or cha when conditi'ons do not match procedures.He did not recall simulato and may apply in January 1983.
2 events which may approached vessel P/T limits nor recall significM Rob have challenged P/T limits.
0019).
however, he has reviewed the Sumary Report on Vessel Inte h t he He indicated a basic understanding of P/T limits; however,
~
He had consider _
needs more knowledge in PTS background and possible events.
t s on the able difficulty in locating equipment, specific controls, and me er He also had difficulty with interpreting the RC,5 wide-rang ure-loop temperature indications, and in determining de control board.
to-saturation on the saturation curve.
dumps and SI pump head / flow' values and also needed assistance in He also had no appreciation of possible auxiliary feedwater controls.
ture and competing steps of termination of SI and controlling RCS temp d ksystemto pressure increase during an SLB, nor how to co achieve these goals.
t that challenged could help him in PTS events, nor previous Robinson 2 even s reactor vessel P/T, limits.
ths.
Althougi One 50 was interviewed who has not been on sh an increase he had received PTS training, he believed that the PTS con He stated in RCS tecperature after idecrease in RCS temperature and that the pressurizer surge line is on the cold leg and ROBINSON SER INPUT S 3-9 04/09/82 d
mt
_.- a difficulty) to reevaluate his statement. During discussions on the. steam line break, he attempted to use the steam generator tube rupture (SGTR) pr i
He took almost 2 minute. to determine his.
s lieu of the SLB recovery procedure.
He did not appreciate possible competing steps concerning control of error.
RCS temperature and pressure increase coupled with terminatioscriteria fo In It was obvious that he has not "Galked thru"- the procedure for some time.
addition, fie did not recall specifics of the SLB when he last had simulator He did recall two Robinson 2' events (safety valve failure and large training.
He believes 1eak in an RC pump) that challenged reactor vessel P/T limits.
they could have been helpful in reviewing PTS' history.
The final two 50s were recently licensed and had received addi.tional simula
~
Both were very knowledgable about reactor vesse1 training in February 1982.
P/T limits and the PTS issue; however, both stated that the PTS training wa conducted after the simulator training. They had worked as a team with ot 50 candidates and did con' sider reactor vesselP/T. limits in many of their Although they considered that PTS classroom training was good, exercises.
did not receive prepared training material.
(They apparently were not aware of the PTS reference material which had been recently placed in the co Both 50s were' exceptionally knowledgeable in predicting SLB response of possible repressurization with and without steam bubbles in th They recognized that the SLB model at the Shearon Harris Simulato respond to the same event at Robinson 2.
The Robinson 2 PTS training outline was reviewed prior to the site visit and found to be acceptable with the.gener&1 criteria as we April 5-7,1982 The CP&L training was conducted 'over a 2-month
- most of the specific criteria.
All licensed personnel were period and consisted of six classroom sessions.
required to attend the training sessions; however, STA attendance l
No forgal evaluation of the effectiveness of the training was mandatory..
conducted; however, the instructor did question individuals during th r-room sessions.
ROBINSON SER INPtJT SEC' 3-10 est0W)
- J.a P
3.4 Sumary On the positive side, it was clear that operator training, specifically on.the i.
PTS issue, had been conducted by CP&L.
A general awareness of brittle fracture.
concerns existed, and some personnel interviewed were very' good on procedural walk-thru's and control board knowledge (indications, co,ntrols, etc). The procedures used in the control room frequently reference curv'e's of NDT limits,
~
l-particularly those procedures us'ed for nonnal heatup or cooldown evolutions..
Some accid nt procedures address the PTS issue, specifically the modified SI e
termination criteria in EI-1, Appendix B, " Loss of Secondary Co'olant."
On the negative side, our audit of seven plant personnel in the control room 1
produced a varied response from very good to poor.
Knowledge of the PTS issue, l
location of key control room indicators and controls, and procedural walk-thru's 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 fracture concerns in the LOCA instructicns, and a relatively high We -
pressure (2000 psig) remains as one of the four SI tennination criteria.
also noted that no emergency procedures addressed strategies'en khat to do ones the operator found himself in a severe PTS condition (specifically, trying to reduce pressure or. minimize repressurizat. ion).
In addition; step 2.9 of EI-1, Appendix B, provides minimal guidance to the operator on using stgam dump valves to stabilize' temperatures following a steam line or feedwater Tine Excessive dumping of steam could extend the cooldown transient. ' With break.
regard to the PTS classroom training, STAS were not required to attend the sessions and.the absen'ce of CP&L validation of the learning process were large reasons for the variation in PTS knowledge.
The previous overcooling history i
l of Robinson 2 provides a particularly valuable training tool which was not emphasized sufficiently.
c While The existing procedures remain weighted toward core cooling concerns.
calculations performed conservatively to bound PTS concerns (WCAP-10019) have merit (anal,ogous to Appendix K core cooling calculations), the use of only l
=-
conservative analyses is not necessarily a sound approach in writing operator As has been endorsed by the industry since the THI-2 accident.in guidelines.
1979, more rigorous " bet.ter estimate" analyses are needed to supplement and Such an objective (currently underway as c
support such procedural guidance.
i.
3-11 ROBIHSON SER INPIIT SEC 04/09/82
.~ -
a m-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 that current procedural inadequacies will be corrected w'ithin approximately one year undeFTMI Action' Item I.C.1 (both from a technical and a human factors standpoint), we conclude that with two exceptions, pro-cedural changes should await completion of this program. Those,exceptioni are, reducir.g 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, "Recomendations").
G G
e e
8 e
9 e
W 6
A
.: ~
z 04/09/82 3-12 ROBINSON SER INPlJT SEC 3
- .~L.
~
4 FRACTURE MECHANICS 4.1 General Aside from the primary mission to audit pr'ocedures and training, the Task.
Force also included in the following sections a discussiori of an overview fracture mechanics and a summary of Robinson 2 reactor vessel properties
' Fracture mechanics analyses and thermal shock experiments have confirm relatively, shallow pre-existing cracks can initiate, that is they can grow.
deeper into a cylindrical metal wall if the inner surface of the cylinde subjected to a thermal shock by rapidly decreasing its temperature to th This i
temperature or icwer.
.., region of the metals nil'-ductility transit on transition region between ductile to more brittle material is referenced f the caterial, which increases in magnitude with neutron irradiaticn.
RTNDT In addition to the tiiertral shock which cculd occur due to a rapid co i t if.
the beltline region of a react.or' vessel., pressure stresses can also ex s d
the primary coolant ocessure is maintained and/or the' system is For vessels with a relatively high RT'NDT'
- after an initial drop in pressure.
i temper-particular cooldown transient is more likely to approach the transit
. herefore, PTS '
ature than if the same transient were to occur in a new v T
i considerations prescribe that repressurization should be avoided to lt the potential.~for jeopardizing' vessel integrity.. This consideratio to an overall objective of minimizing the RCS cooldown and subseque surization while s,till ensuring that the core, remains cool.'
Robinson 2 Fracture Mechanics 4.2 In the fracture analyses of pressurized thermal shock, the fract l
function of of the material is obtained from curves given in the ASME Code as a l
It is the sum of two temperature relative to the reference temperature, RTme
=-
NDT.
quantities, the initial RT caused by radi,ation damage and measured as require HDT and the ARTHDT 10 CFR Part 50.
ROBINSON SER SEC 4 INPU 4-1 04/14/82
..M
_,..,_-_,__...g.
_ ~..
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 Although the longitudinal welds have low nickel content (less copper content.
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.
NDT values were not measured for Robinson 2 becaus Initial RT For the circumferential fabricated before the ASME Code rules were in place.
welds, there were three Charpy tests at +10*F.
From these results, a conserva-was 6btained by using the methods tive estimate of 0*F for their initial RTNDT given in 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 For the longitudin'al be estimated; hence, the latter is used as a best estimate.
welds, th'er6 are no records available, except that they were made with ARCOS From a limited amount of information obtained from cther B-5 weld flux.
values were assumed by.us to be the same as those for plants, the initial RTNDT the circumferential welds--O F for the conservative estimate 'and -2 4
best estimate.
4
'The only measurement of copper content for Robinson 2 welds is a value o for the surveillance weld, which matched the circumferential weld near the top of the core, but not the weld where fluence was greates't.
Consequently, for our prediction of RTNDT, the copper content of the longitudinal welds w mated to be 0.30% best estimate and 0.35% conservative estimate.
analysis of the circumferential weld, 0.34% copper was used for the best For the conservative estimate, the calculated value of shift using estimate.
0.34% copper exceeded the upper' limit of Regulatory Guide 1.99, Revision 1 which tiounds all known surveillance and test data in this fluence region; Nickel hence, the Regulatory Guide prediction was followed, as given below.
content was taken to be 0.1% and 0.75%, respectively, for the longitudinal circumferential welds (best estimate valuec) and 0.2% and 1.2%
ative estimates.
aO b8 x
ROBINSON SER SEC 4 INPUT 4-2 04/14/82
Fluence values foi the various weld 1ccations ara given in the "150 day" report to D. G. Eisenhut from CP&L dated January 25',1982 (7.2 EFPY).
Fo'r the longi-9 tudinal weld, the fluence as of December 31, 1981 was estimated to be (E > 1 MeV) at the inside surface of the weld.
For the 1.30 x 1018.n/cm2 circumferential weld the value was 1.24 x lb18 n[cm
~
2 (E ) 1-MeV).
(The criti-cal weld is below the peak axial fluence ~1ocation.)
was developed from analysis of The trend curve used by us to calculate ARTHDT 136 PWR surveillance data points by G. Guthrie of HEDL.
His mean curve formula, which has terms for percent copper, "Cu," nickel, "Ni," and fluence, "f" is:
NDT = E5 + W Cu.+ 270 CuNG,(f/10")W ART The standard deviation was 22*F.
The mean curve was used by us to complete the "best estimates" and the mean plus 2-sigma was calculated for the " conservative' estimates."
t Substituting the appropriate values in the Guthrie fomula, our current values of RT f r the Robinson 2 welds are:
NDT Best Estimate
. Conservative Estimate.
Longitudinal 140 F 240 F Circunferential 220 F 290 F These values were reported by us in a Commission meeting on March 9, 1982 and were compared with the licensee's conservative estimat'es for the longitudinal and circumferential welds'of 183 F and 290 F, respectively.
[
Current pressure-temperature Appendix G limits being used by Robinson 2 were submitted by letter of January 4, 1977 and were previously accepted by the NRC in a letter dated January 25, 1977. The curves are intended to apply for
. = - -
A recheck of these limits against the 20 EFPY, or about 13 EFPY beyond today.
has information available today regarding fluence accumulation and RTHDT confirmed our acceptance of the pressure-terperature limits.
(An LER dated January 11, 1982 alerted the NRC to a possible 5 F error in the P/T limits, but 04/14/82 4-3 ROBINSON SER 5EC 4 INPUT
resolution of this is' sue is not expected to change the general conclusion.)
These limits do not apply to cooldown rates exceeding 100*F per hour. At th'at-cooling rate, the thermal stresses produce values of K -thermal that are only a y
fraction of K pressure, whereas in more severe (postulated) thermal shock g
transients the reverse is true.
Since definitive cooldown rate-dependent brittle fracture criteria beyond the
~
Appendix G limits have yet to be decided, it is therefore of, interest to
~
minimize any severe RCS cooidown and subsequen( repressurization, while stU1 ensuring that the core remains cool.
The preceding Sectio'n 3 addresses our audit of the operations staff at Robinson to determine their level of awareness l
of this concern, and the procedural guidance available in the corytrol 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
.c l,
minimize the probability of crack initiation.
J i
i (3) If (1) or (2), above, is not.possible (severe, rapid overcooling accident),
limiting RCS pressure to minimize the probability of through-wall crack propagation.
The licensee has indicated that for the conservative overcooling scenarios j
analyzed in WCAP-10019, at least 31 EFPY remain for the _ Robinson 2 reactor vessel.
However, key technical questions on assumptions for these analyses are not yet resolved.
An example is when to allow credit for warin pre-stress (WPS) which is dependent on defining the events which create PTS risk.
Current experimental information suggests that the beneficial effects of WP5 could be precluded after a cooldown and subsequent repressurization later in the transient.
As addressed at the March 9 Cc:maission meeting, the above question I
and uncertainties are b'eing pursued intensively, but final resolution will not l
be available for the June 1982 reassessment.
ROBINSON SER SEC 4 INPUT 04/14/82 4-4
7 j
Aside.from the prim'ary mission of the Robinson 2 Task Force to audit procedur.es 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 for Robinson 2 pending resolution in,1983.,,While conservative worst-case PTS scenarios are being sought and analyzed, our attention focused on the
~
more probable overcooling scenarios (antic} pated operational occurrences). -
Previous staff evaluation has benchmarked the Ranch'o 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 evaluation's using conservative Robinson vessel properties support a period of at least one yehr 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 that the more likely scenarios (such as a stuck ~
open PORY 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.
O 3
e e
4 e
04/14/82 4-5 ROBINSON SER SEC 4 INPUT
.. - - L. :. -
...e___
5 RECOMMENDATIONS Based on the summary of findings in Section 3.4, which includes the key procedural and training shortcomings, the Robinson 2 PTS Task Force concludes that additional action by CP&L is warranted, particularly in thTtraining area.
The following recommendations are provided; Prior to restart, and pending longer tern generic resolution of the PTS issue, all Robinson 2 operators and STAS should be retrained in the following areas:
This includes all (1), Review of previous ' overcooling events at Robinson 2.
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 P/T limits and sketch the typical progress of key parameters until recovery is achieved.
This exercise 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 mitigate,the transient.-
This includes instrumentationland controls during the recovery phase, with
. a complete walk-thru until c.onditions stabilize.
Emphasis should focus on discussing alternatives for recovering from a PTS condition, and'alterna-tives for minimizing RCS overcooling and subsequent repressurization, while still ensuring that the core remains cool. The shift should provide feedback of any questions or comments arising from these drills to plant management.
Resolution to these questions or comments should then fol. low, with revised procedures and additional training as necessary.
A CP&L audit of the shift's ability to cope with a PTS event should be (3) made after the above,is completed. This includes a short quiz and a drill or demonstration at,the console.
=-
- In the longer term, an independent audit of the ability to cope with PTS using the new I.C.1 procedures st ould be cade to verify an acceptable level of L
5-1 ROBINSON SER SEC 5 INPLIT 04/14/82
.-,e.
e-g * *
- e e
e e
._m-o
~~ -training.
Also, CP&L should review the Shearon Harris Simulator response for PTS events to verify that the models are reasonable and can demonst' rate stead bubble (s) in the reactor coolant system (i.e., vessel head) during forced flow and natural circulation.
Identified anolmalies between the simulator and
'~
Robsinson2responsesshouldbediscussedduiingthetraininfprocess.
l With regard to the current emergency procedures for safety injection termination:
(1) We recomend that prior to restart the SI termination criteria of 2000 psig be modified to lower the pressure at which the operator can secure SI, while still observing adequate subcoolirg, heat sink, and pressurizer.
l 1evel.
Discussions with the licensee and Westinghouse indicate that this j
value could be the safety injection ptsap cut-off head, plus uncertainties (about 1600 psig).
l i
(2) We recomend that prior to restart step 2.9 of EI-1, Appendix B, l
" Detailed Recovery Procedure Steam Line or Feed Line Rupture," be revised l
to provide clear instructions for controlling temperature and pressiire 5
following dryout of the faulted steam generater.
Such instructions should include recognition of the potential for extending the overcooling transient.
(3)
In the longer term, we rec'omend more consideration be given to 3
l j
lowering the RCS pressure SI temination. criterion further than j
(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 safety grade subcooling meter, should reduce the need to accomodate uncertainties A
with as high a pressure reference in the LOCA SI temination criteria.'
Criteria similar to the steam line break procedure (suitably weighted for bothcorecoolingandPTSconcerns)couldthenbeadoptedintheother
[~
accident procedurei.
t j
04/14/82 5-2 ROBINSON SER SEC 5 INPUT
~
APPLICABILITY TO REMAINING SEVEN PWRS 6
tiv^e of the The remaining seven PWRs which have been i.dentified as representa plants having a relatively high RT are:
HDT
'Ft. ' Calhoun (CE)
Oconee (B&W)
San Onofre (W)
' Turkey Point (W)
Calvert Cliffs (CE).
Since it is likely that San Onofre and Turkey Point emergency l
like Robinson, based on simi,lar initial Westinghou conclusions would probably equally apply.
b of main analyses (WCAP-10019) may Sot apply to San On steam line isolation valves.
to cope with to increase the importance of adequate piocedu secondary side breaks.
directly applied to operations staff audits are plant specific and cannot.be the Turkey Point and San Onofre plants.
be The general procedural and training criteria identified applied to each of the plants to be audited.
to plant and accident analys'es is warranted to verify applicability
' configuration.
Based on the problems disclosed during the Robinson revie i h worst vessel necessary to audit six of the remaining seven plants w t be excluded as properties prior to the Commission briefing in Ju they are not operating).
ii for Northwest Laboratory (PNL) personnel audit the procedur d Maine San Onofre 1, Ft. Calhoun, Turkey Point, Oconee, Calver The team (s. ) should consist of, as a minimum:
Yankee.
ROBINSON SER SEC 6-1 04/14/82 3
.y
- - ~ '
= -..
o
,e plant operat1ons specialist (preferably an, operator licensing examiner), a reactor systems specialist for analysis evaluation, and a fracture mechanics
~
specialist.
The team members (as necessary) should visit each site to expedite the audits, to' interview operations personnel,, and,to discuss questions with the licensees.
It may not be necessary for all team membe5s (e.g., the fracture mechanics specialist) to visit each site.
..The team (s),will, conduct an evaluation of each plant's training program for PTS, and conduct a technical and human engineering review of each plant's procedures used during possible PTS events. These reviews will use criteria developed from the Robinson 2 evaluation conducted April.5-7, 1982.
It is anticipated ilhat the site visits will require 3-5 days each..Therefore.
to complete the audits in early June, the site visits should be conducted at a rate of one a week, beginning April 19, 1982.
A draft evaluation should be provided at the end of the veek following each evaluation.
It appears th'at tv1 or more teams will be needed to meet this schedule.
Because of questions q
raised during the SEP review of San Onofre 1, we recommend that it be the fi.rs8
[
plant to be audited.
The OR project manager for each plant should attend the plant visits to provide liaison between the review team and the plant, since hG i
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 n'eces'sary
" documentation and personnel are available at the site,.to. ensure. an efficient evaluation.
i l
The reports will be submitted to the Generic Issues Task Manager, who may, depending on the findings, request additional evaluation by PTRB, LQB, RSB, or, MTEB.
The final evaluation will be sumarized by t'he Generic Issues T,ask j
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 at i
least one per vendor complete by June.
This would leave Ft. Calhoun, Oconee, and San Onofre as the next three candidates.
Assuming a team effort is utilized (PNL), the enclosed schedule outline is proposed.
1 04/14/82 6-2 ROBINSON SER SEC 6 INPlff
j J,
5 Prior to further site audits,.however, copies of.this Robinson 2 report should
- be made available to the six p',lants.
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 simiiar prob 1 ems 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.
e~
9,_
W a
ROBINSON.SER SEC 6 INPUT 04/14/82..
6-3,.
. fl
~
ov
.. =..
April May June 15 3
M -
T
/
/C 2]
l 1.
Robinson Review Complete q
2.
San Onofre i
m Review
^
l 3.
San Onofre Site Visit 4.
San Onofre g
Report 5.
ft. Calhoun a
m Review
^
6.
ft. Calhoun n
Site Visit 7.
Ft. Calhoun O'
Report 8.
Oconee a
Review I
9.
Oconee Site Visit
- 10. Oconee g
Report Y
i Summarv
-- About 3 weeks each plant (total) l
-- 3 day site visit
-- About I week writing report s-i
-