ML20058K133
| ML20058K133 | |
| Person / Time | |
|---|---|
| Issue date: | 12/05/1990 |
| From: | Jordan E Committee To Review Generic Requirements |
| To: | Arlotto G, Miraglia F, Johari Moore NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), Office of Nuclear Reactor Regulation, NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| References | |
| NUDOCS 9012070243 | |
| Download: ML20058K133 (2) | |
Text
_ _.___
o, UNITED STATES
[\\
g NUCLE AR REGULATORY COMMISSION
.s r li WASWNGTON, D. C. 20bbb
%,,,,,+
December 5, 1990 MEMORANDUM FOR:
Guy A. Arlotto, NMSS Frank J. Miraglia, Jr., NRR Janice E. Moore, OGC Brian W. Sheron, RES Leonard J. Callan, RIV FROM:
Edward L. Jordan, Chairman Committee to Review Generic Requirements
SUBJECT:
CRGR MEETING NO. 196 The Committee to Review Generic Requirements (CRGR) will meet in the afternoon of Wednesday, December 12, 1990, in Room 6507, MNBB. The agenda is as follows:
O 1:00-3:00 p.m.
E. Rossi (NRR) will present a briefing on technical specification activities, includi*q:
(1)
The standard technical specification program, including criteria for removing requirements and a proposal that has been discussed for removing PTS curves.
(2) for recent waiver requests related to technical specifications:
(a)
Plant-specific letters on requirements for BWR
^
scram accunL'. ors.
(This waiver request, along with an associated issue sheet, was distributed I
to members on 11/21/90.)
(b)
Generic letter on requirements for response time testing.
(This waiver request was distributed to members on 11/21/90.)
(c)
Generic letter on reactor vessel surveillance specimen removal schedule.
(This waiver request was distributed to members on 11/21/90.)
l (d)
Generic letter on component lists.
(This waiver request was distributed to members on 12/4/90.)
If a CRGR member cannot attend the meeting, it is his responsibility to assure that an alternate, who is approved by the CRGR Chai, man, attends the meeting.
Persons making presentations to the CRGR are responsible for (1) assu the infonnation required for CRGR review is provided to the Committee (CRGR
]0 Charter - IV.B), (2) coordinating and presenting views of other offices, (3) as appropriate, assuring that other offices are represented during the F. g\\ '
c m _
O D- [
9012070243 901205
/
'~g(g/'g
'p p
PDR REVGP NROC 4GR "ee=m 's Poc 20 RE CWM @W I
e, '
2 presentation, and (4) assuring that agenda modifications are coordinated with the CRGR contact (J. Conran x29855) and others involved with the presentation.
Division Directors or higher management should attend meetings addressing agenda items under their purview.
In accordance with the ED0's March 29, 1984 memorandum to the Commission concerning " Forwarding of CRGR Documents to the Public Document Room (PDR),"
the review packages for items scheduled at this meeting, which contain predecisional information, will not be released to the PDR until the NRC has considered (in a public forum) or decided the matter addressed by the information.
Oiidnel Signed by:
Dcnwood F. Rose, v
(a Edward L. Jordan, Chairman Committee to Review Generic Requirements cc:
SECY J. Taylor Commission (5)
J. Lieberman, OE Regional Administrators W. Parler, 0GC R. Fraley, ACRS P. Norry, ADM T. Murley, NRR Distribution:
CRGR S/F CRGR C/F S. Treby M. Taylor J. Sniezek J. Richardson E. Rossi G. Thomas R. Lobel M. Lopez Otin E. Sullivan A. Thadani P. Kadambi T. Dunning G. Mizuno K. Wichman J. Conran D. Ross..
E entral. File
- D. Allison C
PDR (NRC/CRGR) l (AGN196.DTA 1
l l
CRGR:AE0DA DD:AE00 C/l D
DAll on:sim Dross EJ 12/ /90 12/ /90 12 '
90
s.
4 WQLF CREEK
~
NUCLEAR OPERATING CORPORATION
- d*.7' November 30, 1990 che t=ww oe.,
WM 90-0194 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk Mail Station P1-137 Washington, D. C. 20555
Reference:
Letter dated July 24, 1990 from D. V. Pickett, NRC, to B. D. Withers, WCNOC Subjects Docket No. 50-482: Response to Request for Additional Information Concerning Steam Generator Tube Rupture Operator Action Times for the Wolf Creek Generating Station Gentlemen:
Attached is Wolf Creek Nuclear Operating Corporation's (WCNOC) response to the Reference which requested additional information regarding Steam Generator Tube Rupture (SGTR) operator action times for Wolf Creek Generating Station (WCGS).
The Attachment to this letter provides information on operator. response times from five simulated SGTR scenarios.
Three of the SGTR scenarios deal with steam generator overfill and two scenarios with a stuck-open atmospheric relief valve (ARV).
The stuck-open M V is the limiting SGTR accident scenario for WCGS since it has the largest oose consequence.to the general public.
The operator action times obtained from the simulated SGTR scenarios have demonstrated that the action times assumed in the analysis are realistic and are representative of the current operator population at WCGS.
WCNOC believes that the information provided in the Attachment is sufficient for closure of License Condition 2.C.(11).
We would be happy to meet with you to discuss this submittal if necessary.
Very truly yours,
{
_p L
art D. Withers President and Chief Executive Officer BDW/jra Attachment cci A. T. Howell (NRC), w/a R. D. Martin (NRC), w/a l
D. V. Pickett (NRC), w/a l
M. E. Skow (NRC), w/a l
P.O. Fkm 411/ Burhngton, KS 66639 / Phone:(316) 364 8831 Dk M An Equal opportunny Employer Mf/HC' VET
///
Attachm:nt to WM 90-0194 Pago 1 of 13
+
RESPONSE TO NRC REQUEST FOR ADDITICIIAL INFORI ATION manAnnING STEAM GENERATOR TURE RUPTURE OPERATOR ACTION TIMES FOR WOLF ca m GENERATING STATION QUESTION:
The licensee should provide further information that demonstrates that the operator response times assumed in the Wolf Creek analysis are realistic:
that is, are representative of the current operator population at the plant, and that the maximum response times fall within the bounds of the analysis.
This information should address both control room actions and actions performed outside the control room such as manual steam generator liquid sampling.
The licensee should provide response times for the following operator actions Identify ruptured steam generator.
Isolate ruptured steam generator.
Initiate cooldown of reactor cor.iant system.
Complete cooldown of reactor coolant system.
Initiate depressurization of reactor coolant system.
Complete depressurization of reactor coolant system.
Initiate safety injection.
Terminate safety injection.
Equalize pressure.
Assuming design basis conditions, the licensee should provide demonstrated operator action times for (1) the worst case dose scenario and (2) the worst case overfill scenario.
RESPONSE
1.0 Introduction In response to Reference 1. Wolf Creek Nuclear Operating Corporation (WCNOC) performed simulated steam generator tube rupture (SGTR) scenarios on the Wolf Creek Generating Station (WCGS) simulator during the period,9/05/90 to 9/12/90.
Additionally, information from a scenario performed on 8/29/86 was utilized to support the
~
request for additional information.
The scenarios simulated included three steam generator overfill events and two stuck-open atmospheric relief valve (ARV) events.
l L
Attachm:nt to WM 90-0194 8
Page 2 cf 13 2.0 Scenario Descriotion 2.1 Stuck-Open ARV Cases The design basis SGTR with the maximum potential offsite dose is one with the failure of the ARV on the faulted steam generator in the wide-open position.
This failure releases radioactive steam directly to the atmospheres if the ARV is not isolated, it has the potential to release the entire contents of the steam generator to the atmosphere.
Isolation of the faulted steam generator is accomplished when an equipment operator (EO) is dispatched by the control room operators to manually close the faulted ARV's block valve.
2.1.1 Initial conditions
(
The initial conditions assumed for the stuck-open ARV case are detailed in Table 3-2 of Reference 2.
In the analysis of the design basis SGTR, initial values of plant parameters are determined by adding or subtracting parameter uncertainties as appropriate to maximize the resultant offsite doses.
However,-
since the goal-of the simulator scenario was only to validate operator response times assumed in the analyses, the initial' simulated plant conditions corresponded to nominal plant conditions with the exception that initial break flow was manually
" dialed-in' to correspond to a value of 480 gpm which was calculated by analysis.
g 2.1.2 Availability of Offsite Power l
For the stuck-open ARV case, offsite power is assumed lost coincident with reactor trip. This results in increased releases to l'
the atmosphere and higher offsite doses than would otherwise occur E
if offsite power were available.
In addition to the release through the stuck-open ARV, a loss of offsite power results in radioactivity being released through the main steam line safety valves without l
being filtered through the condenser air removal system.
L 2.1.3 Operator Response to Stuck-Open ARV l
Steam release through a stuck-open ARV is terminated by manual closure of the associated block valve.
The longer it takes the i
control room operators to isolate the faulted steam benerator, the greater the offsite dose release. A conservatively long estimate of the time to manually close the block valve is 20 minutes after the ARV fails open.
The ARV is assumed to fail open coincident with its first initial lift immediately following reactor trip.
Reference 2 initially discussed the 20 minute failed-open period and Reference 3 presented the times required by an E0 to physically walk to the Steam Tunnel and close the ARV block valve.
Attachm:nt to WM 90-0194 Page 3 cf 13 During an emergency situa*.Aon, control room operators dispatched to close the ARV block val',e would walk from the Control Room through the Secondary Alarm Station and into the Control Room Filtration Room in the Auxaliary Building.
Once in the Auxiliary Building, the operator would have a direct route into the Steam Tunnel.
The total distance is approximately 180 ft.
The expected travel time is 1 to 3 minutes.
Once in the Steam Tunnel, it wr ld req"tre 3 to 5 minutes to identify and operate the ARV block valve.
As a result, the total isolation time is estimated to be between 4 and 8 minutet.
For the simulated ARV cases, once the control room operators have identified the stuck-open ARV, they contact an E0 over the plant paging system and direct him to manually close the ARV block valve.
Because it requires a maximum of 8 minutes to isolate the ARV block
- valve, the control room operator must complete his call to the EO within 12 minutes after the ARV opens (at reactor trip) in order that the ARV, and thus the steam generator, be isolated within the required 20 minutes.
2.2 Steam Generator Overfill Cases The worst case single failure with respect to steam generator overfill is a failure in the open position of the auxiliary feedwater ( AW) control valve on the discharge side of the motor driven AW pump feeding the faulted steam generator.
Failure of the control valve coupled with the flow contribution from the turbine driven A W pump can supply initial A W flow to the faulted 9 team generator to values near 723 gpm.
In addition, realizing that AW flow is delivered as a function of steam generator pressure and that pressure decreases as a result of relief valve actuation after trip.
AW flow can increase until AW flow to the faulted steam generator is terminated.
Isolation of the faulted steam generator is accomplished when AW ' flow is terminated to the faulted steam t.
generator.
l l
2.2.1
-Initial Conditions The initial conditions assumed for the steam generator overfill g
case are detailed in Table 3-1 of Reference 2.
In the analysis of I
the desica basis overfill scenario, initial values of plant parameters are determined by adding or subtracting parameter uncertainties as appropriate to maximize the resultant overfill l..
potential.
h
- However, since the goal of the simulator scenarios was only to validate operator response times assumed in the analyses, the initial simulated plant conditions corresponded to nominal plant conditions with the exception that initial break flow was manually i
- dialed-in' to correspond to a value of 445 gpm which was calculated by analysis.
Att2 chm:nt to WM 90-0194 P0ce 4 cf 13 2.2.2 Availability of Offsite Power The potential for steam generator overfill is not strongly dependent on the availability of offsite power.
- However, the potential for overfill is slightly greater if offsite power is assumed lost at reactor trip La AFW flow is initiated esrlier.
P.tr t he rmor e.
if overfill should occur, then subsequent offsite doses are greater if offsite power is assumed lost.
2.2.3 Operater Response To Steam Gens lor Overfill The potential for overfill of the faulted steam generator is largely ne6ated when AFW flow to the faulted steam generator is terminated.
Control room operators are aware of AFW control valve malfunction
)
when the faulted steam generator's narrow range level is increasing significantly coupled with the indication that the AFW control valve is wide-open.
Isolation of the AFW flow, and thus the steam generator, is accomplished by the operator's action to deenergize the appropriate motor driven AFW pump.
This action is accomplished from the control board by the balance of plant (BOP) operator when i
he places the control of the motor driven AFW pump switch into the
" pull-to-lock' position.
3.0 Operator Responses to Mitinate an SGTR As identified in Reference 1, there are nine operator responses which must be performed in a timely manner to mitigate the consequences of an SGTR.
These responses are irrespective of the scenario assumed and are:
1)
Identify the ruptured steam generator, 2)
Isolate the ruptured steam generator,
- 3) Initiate RCS cooldown, 4)
Terminate RCS cooldown, l'
5)
Initiate RCS depressurization,
- 6) Terminate RCS depressurization, 7)
Initiate safety injection,
- 8) Terminate safety injection, and 9)
Equalize primary and secondary pressures.
These individual operator responses have previously been extensively described in References 2 through 5.
Discussion of the responses will however be reiterated below.
Attachment to Mi 90 0194 Page 5 of 13 i
3.1 Identification of the Ruptured Steam Generator Reference 5 details the numerous indications available to the control room operators to alert them to the occurrence of a SGTR.
In the simulated SGTR scanarios, identification of the particular steam generator ruptured can occur at any time the operators can state unequivocally that tube rupture is in progress.
This determination can be made either before or while in Emergency Operating Procedure, EMG E-3, 'bteam Generator Tube Rupture'.
Given. a reactor trip or safety injection signal as a result of an
- SGTR, the control room operators would enter Emergency Operating Procedure, EMG E-0,
" Safety Injection" which governs their actions to verify the proper response of the automatic protection system following manual or automatic actuation of safety injection.
Thrcugh symptom-based diagnosis, the operator is directed to transfer to EMG E-5 when indications are such that a tube rupture is in progress.
Once the operator is in EMG E-3, procedural guidance in Step 2 requires identification of the ruptured steam generator.
This is accomplished by observing one of the following:
t.
- 1) Unexpected rise in any steam generator's narrew range level, or 2)- High turbine driven AFW pump exhaust radiation, or
- 3) High radiation from any steam generator steamline radiation monitor, or by
- 4) Steam generator blowdown samples.
Items 1 through 3 above can be observed in the con.rol room.
Item 4 allows a manual sampling of the suspected faulted rteam generator as well as a sampling of.the intact steam generator blowdown lines for verification of the faulted steam generator.
Reference 5 provided in, formation regarding the capabilities for manual sampling of the steam generator.
Tables 1 and 2 provides both the time of identification ano isolation of the faulted steam generator on the same line.
This is due to the fact that it is not possible to precisely document when in fact the control room operators identify that a SGTR has occurred.
For example, for all of the scenarios presented, the operators were aware that a potential SGTR was in progress early in l
the transient..
Statements made by the operators ranged from 'Looks like a rupture in SG A' to 'We've got a rupture in SG A' and to 'SG E
A is showing signs of a tube rupture'.
Rather than attempt to l
assign an observed time value to statements like these, the identification and isolation of the faulted steam generator was lumped and then compared to the assumed response time as recorded in References 5 and 6 for the stuck-open ARV case and the steam generator overfill case, respectively.
l
=
Attachm:nt to WH 90-0194
]
Page 6 cf 13 3.2 Isolation of the Ruptured Steam Generator Steps 3 and 4 of EMG E-3 requires isolation of the faulted steam generator by performance of the following:
- 1) Adjusting the ruptured steam generator's ARV controller to a high setpoint or 1125 psig and verify it closed.
2)
Close steam supply valves from ruptured steam generator to the inlet of the turbine driven AFW pump, 3)
Ensure blowdown lines have isolated.
4)
Clase faulted steam generator's main st eamline isolation, bypass, and drain valves, and 5)
Stopping AFW flow to the faulted steam genorator when narrow t
rar.ge level has reached 42.
Specific to *.he stuck-open ARV case, the faultet. sttam generator is assumed to be isolated within 20 minutes following retetor trip, the point at which 'it is first opened to relieve pressure.
The 20 minute time period includes 8 minutes for manual isolacion of the valve.
Once the ARV block valve is closed, the f.iulted steam generator'is isolated in that the direct path to the atmosphere for
[
offsite release !s terminated. As is indicated in Table 1 the steam generator was isolated, on averate, about 4.5 minutes prior to the assumed response time of 28.4 minutes.
In both cases, the EO was dispatch to isolate the ARV within 12 minutes after the valve stuck it
'open position.
Assuming an 8 minute period for the EO to effi
.osure of the block valve, the stuck-open ARV was isola
. thin 20 minutes of its initial opening.
Specific to the steam generator overfill scenario, the faulted steam generator is considered isolated when its AFW flow is terminated.
b The overfill concern is negated at this point as no other feedwater enters the steam generator from this point on.
As is indicated in Table 2 the steam generator is isolated, on average, about 4.5 l
-minutes prior to the assumed response time of 16 minutes.
3.3 Initiation of RCS Cooldown Cooldown is begun when the intact steam generator's ARVs are opened i
to allow steam dump to the atmosphere.
This assumes that offsite power is lost..
Should offsite power be available, then steam could be dumped to the main condenser.
Table 1
details that.for the stuck-open ARV case, the control room operators initiate RCS cooldown well before the assumed response time of 40 minutes.
In the scenario of 9/05/90, the operators actually began cooldorn before the faulted ARV was isolated.
This action is acceptable per EMG E-3.
The 9/12/90 scenario saw the SO wait until the ARV was isolated until he allowed cooldown.
In either case, the operator response times were on the order of 19 minutes earlier than required by analysis.
Attcchment to WM 90-0194 P ge 7 cf 13 For the overfill scenario.
Table 2 shows that cooldown was initiated, on the average, about 12 minutes prior to the assumed
- esponse time.
Note also that the crew of 8/29/86 elected to begin cooldown prior to isolation of AFW flow.
- Again, this is merely a judgement call by the So and not a violation of procedure.
3.4 Termination of RCS Cooldown Cooldown is terminated upon closure of the intact ARVs and is done when the appropriate core exit thermocouples/RCS vide range temperatures are reached.
t Table i shows that for the stuck-open ARV scenario, cooldown was E
terminated, on average, about 20 minutes prior to the assumed response time.
Table 2 indicates that for the overfill scenario cooldown was terminated, on average, about 16 minutes prior to the assumed response time.
3.5 Initiation of RCS Depressurization Depressurization of the RCS is initiated shortly after cooldown has terminated.
EMG E-3 specifies the use of one pressurizer pilot operated relief valve (PORV) to depressurize should normal spray not be available, as is the case when offsite power is lost.
Table 1
details that for the stuck-open ARV scenario, depressurization was initiated well before the assumed response time.
The average response time was about 16 minutes prior to tho assumed response time.
Table 2 indicates that for the overfill scenario depressurization was initiated, on average, about 15 minutes ahead of the assumed response time.
3.6 Termination of RCS Depressurization The decision to terminate depressurization iw based on either the difference between RCS and faulted steam generator pressure, pressurizer
- level, or the amount of RCS subcooling.
Depressurization is terminated upon closure of the pressurizer PORV.
S.
Table 1
indicates that for the stuck-open ARV scenario, l.
depressurization was terminated well before the assumed response time.
The average operator response time was about 16 minutes prior
?'
to the assumed response time.
Table 2 details that for the overfill scenario, depressurization was terminated, on average, about 13 5,
minutes prior to the assumed response time.
3.7 Initiation of Safety Injection Safety injection (SI) is not an operator action for the two analyzed SGTR scenarios but is an automatic actuation as the result of losing offsito power.
i Att0chment to WM 90-0194 Page 8 of 13 3.8 Termination of Safety Injection Following completion of RCS depressurization, EMG E-3 requires several conditions to be met prior to termination of SI.
These ares a minimum amount of RCS subcooling, a secordary heat sink via at least one intact steam generator, a minimum RCS pressure, and a stable or increasing RCS pressure.
Table 1 shows that for the stuck-open ARV scenario, SI is terminated, on average, about 13 minutes prior to the assumed response time.
Table 2 indicates that for the overfill scenario control room operator response terminates SI, on the average, about 12 minutes earlier than the assumed response time.
3.9 Equalization of Primary and Secondary Pressures The immediate situation after termination of SI is that RCS pressure 1
is a few hundred psi greater than the faulted steam generator and break flow, though reduced, still continues.
EMG E-3 then requires j
that the operator equalize pressures between the RCS and the faulted steam generator.
Continued steam release from the intact steam j
generators' ARVs is necessary to remove decay heat which reduces j
secondary pressure which, in turn, reduces RCS pressure.
To aid in the pressure equalization, the operators may elect to open the pressurizer PORV following termination of SI to obtain equalization.
Table 1 details control room operator response in the stuck-open ARV scenario.
As is indicated for the 9/05/90
- run, pressure equalization / backfill did not occur within the required 63 minute time period.
This was due to the placement of a tag on main control board RL001 which identified the incorrect charging train in service.
EMG E-3 requires that 60 gpm charging flow be established immediately following termination of SI.
Due to the incorrect placement of the tag, the operator spent significant time attempting 1,
to. establish the required charging ' flow.
It was not until the L
simulator controller realized the incorrect tag placement that steps were taken to allow the proper charging flow.
Once charging flow
(.
was established, the operaters immediately took steps-to equalize pressures. Considering that the operators were able to terminate SI almost.14 minutes earlier than the assumed response time, it is reasonably assumed that pressure equalization would have been g
achieved within the required time limit, as it was in the 9/12/90 L
run, had the simulator been configured correctly.
Furthermore, it should also be noted that failure to achieve pressure equalization
.within the assumed response time is insignificant as far as the
~
contribution to offsite releases are concerned.
The stuck-open ARV scenario is essentially terminated from a radiological release perspective when the faulted ARV le isolated twenty minutes following its initial opening at reactor trip.
Following isolation of the ARV,
-the only releases that occur are a result of the Technical Specification assumed leakage limit from primary to secondary of 1 gpm.
This leakage and subsequent partitioned release (as a result of controlled steam releases from the intact steam generators for the purposes of RCS cooldown) is insignificant when compared to those releases occurring when the ARV is stuck-open.
L
Attachm:nt to WM 90-0194 P g3 9 of 13 l
Table 2 details operator response times for the overfill scenario and indicates that, on average, the operators were able to reverse break flow and equalize pres =ures within the assumed response time.
4.0 Simulator Crew Composition Tne simulator crews for the five simulator scenarios were composed of-at least three licensed individuals filling the positions of
)
supervising operator, reactor operator and BOP operator.
For the scenarios performed on 9/05/90, an additional licensed individual was available to fill the position of shift supervisor.
Three different simulator crews performed the five simulator scena'ios.
These crews were made up of licensed individuals from the Operations organization, the Training organization and other licensed SPJs in plant management.
The crews provided a representative cross,ection of licensed operators and the results shown in Tables 1 and 2 indicate the ability' to respond to a SGTR scenario in a timely manner.
As part of the Licensed Operator Requalification Training Program licensed operators are required to review the Emergency Operating Procedures on an annual basis.
The review methods of these procedures include drills using the simulator.
As identified in Reference 5, procedure EHG E-3 is almost identical to the procedure used at the Callaway Plant with only minor variances in step numbering and cautionary. notes.
Union Electric provided information in Reference 7 that verified the demonstrated times are bounded by the assumed operator action times and that the demonstrated times are reasonable and conservative. WCNOC considers the Union Electric demonstrated operator response times as supportive in demonstrating operator-response time at-WCGS since plant design, procedures and simulators are very similar.
Based upon the five simulator scenarios performed at WCGS and those I
performed at the Callaway Plant, Wolf Creek Nuclear Operating Corporation believes that it has sufficiently demonstrated that. the operator response times assumed in the analysis are realistic and are representative of the current operator population at WCGS.
l L
1 L
l
AttachmOnt te WM 9000194 Pago 10 sf 13 5.0 Plant-Specific Criteria The NRC staff's evaluation of the Westinghouse Owners Group WCAP-10698 stipulates plant specific criteria for assessing operator action times in the event of a SGTR.
Provided below is WCGS specific information associated with each criteria.
Criterion 1.
Provide simulator and emergency operating procedure training related to a potential SGTR.
Response to Criterion 1.
Tables 1 and 2 provides the results of WCGS simulator runs conducted on 8/29/86, 9/05/90 and 9/12/90.
Actions to respond to a SGTR are carried-out in accordance with procedure.
EMG E-3,
' Steam Generator Tube Rupture'.
Licensed Operators perform a SGTR scenario on the simulator as part of the Licensed Operator Requalification Training Program.
Criterion 2.
Utilizing typical control room staff as participants in demonstration runs, show that the operator action times assumed in the~SGTR analysis are realistic and achievable.
Response to Criterion 2.
Tables 1 and 2 document the results of five simulated SGTR scenarios performed at WCGS. With the exception of.one identified response, operator responses have shown that the assumed response times in the design basis analyses are realistic and achievable.
Discussion in Section 3.9 has shown that the only failure to meet the required response time was due to an incorrectly configured simulator and not due to operator response, i
criterion 3.
Complete demonstration runs to show that the postulated SGTR accident can be mitigated within a period of time compatible with overfill prevention, using design basis assumptions regarding available equipment and its impact on operator response times.
l Response to Criterion. 3.
Table 2 documents results of the steam generator overfill scenarios conducted at WCGS.
Operator response times are within the bounding values assumed in the design basis l
analysis.
Furthermore, the simulator was configured to be as near to design basis as possible (i.e., initial AFW flow delivered to the i
faulted steam generator was 723 gpm, loss of offsite power i.
coincident with reactor trip, and initial break flow was ' dialed-in' to correspond to' a value of 445 gpm which was calculated by
-analysis).
l Criterion 4.
If the emergency operating procedures specify L
steam generator sampling as a means of identifying the steam I
generator with the ruptured tube, provide the expected time period for obtaining the sample results and discuss the effect on the duration of the accident.
Response to Criterion 4.
Step 2 of EMG E-3 specifies steam generator sampling as one of four methods of identifying a tube rupture.
Reference 5 indicates that manual sampling would require 15-20 minutes, allowing sufficient time to complete steam generator isolation activities within the required response times, j
k **
- Attachment to WM 90 0194 Page 11 cf 13 6.0 References i
1.
Letter dated July 24, 1990 from D.
V.
Pickett, NRC, to i
B. D. Withers, WCNOC, dated July 24,
- 1990,
' Request For Additional Information Concerning Steam Generator Tube Rupture Operator Action Times For The Wolf Creek Generating Station'.
2.
SLNRC 86-01, January 8,
- 1986,
' Steam Generator Tube Rupture Analysis - SNUPPS*.
3.
SLNRC 86-05, April 1,
- 1986,
' Steam Generator Tube Rupture Analysis - SNUPPS'.
(
4.
SLNRC 86-08 September 4,
- 1986,
' Steam Generator Tube Rupture Analysis - SNUPPS'.
5.
WH 87-0029, February 4,
- 1987,
' Response to RAI Regarding SGTR Analysis' 6.
WM 87-0145, May 15,
- 1987,
" Response to RAI Regarding the SGTR Analysis'.
7.
ULNRC-2145, January 29,
- 1990,
' Docket Number 50-483/Callaway Plant / Steam Generator tube Rupture ~0perator Action Times *
.p.
q'4
. ia, \\
o
'I
{:
y YlI' i
1 i
i i
t>
l 4
l n
1
'AttCchm:nt to W 90 0194 Paga 12 of 13 l
TAB 1,E 1 STUCK-OPEN ARV SCENARIO ASSUMED RESPONSE TIME SIMULATOR RUNS (min)
(Response times in minutes)
ACTION (1) 9/05/90 9/12/90 AVERAGE t
Tube Rupture Begins 0.0 0.0 0.0 Reactor Trip 08.40 05.08 02.65 03.87 Identify / Isolate Faulted SG 28.40 25.38(2) 22.51(2) 23.95 Initiate Cooldown 40.00 18.08 23.51 20.80 Terminate Cooldown 50.40 24.47 37.34 30.91 Initiate Depress 53.40 26.55 38.84 32.70 Terminate Depress 55.00 34.33 44.17 39,25 5
Terminate SI 58.00 44.33 45.76 45.05 Equalize Pressure /
I 61.67(
64.92 fackfill Fitd SG 63.00 68.17 (1) The assumed response times are documented in submittal, W 87-0029, dated February 4, 1987 Attachment 2.
(2)
In actuality, the control room operators contacted the E0 at 11.27 minutes for the 9/05/90 run and at 12.0 minutes for the 9/12/90 run to instruct the EO to close the block valve. Allowing 8 minutes to effect closure of the block valve, the stuck open ARV.could have been isolated I,
as early as 19.27 minutes for the 9/05/90 run and 20.0 minutes for the L
9/12/90 run.
In order to make the simulator runs as close to design basio as possible.
Isolation of the faulted ARV was not allowed until 20 minutes following its initial opening at reactor trip.
l (5) DtG E-3 requires that charging flow be established immediately following termination of SI.
Due to the pincement of a tag on main control board RI.001 which identified the incorrect charging train in
- service, the operator spent significant time attempting to establish the required charging flow (see Section 3.9).
It was not until the i
simulator controller realized the incorrect tag placement that steps were taken to allow the proper charging flow.
The time spent j
correcting this problem resulted in the failure to equalize pressures within the assumed 63 minutes.
{
(4)' Following initiation of RCS depressurization, the S0 saw that the
-Safety Parameter Display System was indicating that control should be i
tranuferred to procedure FR P-1,
' Response to Imminent Pressurized Thermt.1 Shock Conditions *.
Performance of this procedure accomplished termination of break flow and backfill by opening the intact steam genetstor ARVs and by initiation of a final depressurization which began at 51.01 minutes and terminated at 56.34 minutes.
_I
Attachm:nt to WM 90-0194 P ge 13 of 13 a
TABLE 2 OVERFILL SCENARIO ASSUMED SIMULATOR RUNS RESPONSE TIME (Response times in minutes)
(min)
ACTION (1) 8/29/86 9/05/90 9/12/90 AVERAGE Tube Rupture Begins 0.0 0.0 0.0 0.0 Identify / Isolate Faulted SG 16.00 09.63 10.38 11.63 10.55 Initiate Cooldown 24.00 09.17 12.05 15.63 12.28 Terminate Cooldown 32.90 22.25 17.80 20.71 16.92 l^
l Initiate Depress 33.90 13.13 18.55 24.63 18.77 Terminate Depress 35.20 18.25 20.38(2) 27.21(3) 21.95 Terminate SI 36.20 19.50 23.93 28.63 24.02 Equalize Pressure /
I}
29.96(
12x17 Backfill Fitd SG 39.90 30.83 30.32 (1) The assumed response times are documented in submittal WM 87-0145, dated May 15, 1987, Table 2.1.
(2)
Two pressurizer PORV openlclose cycles wore experienced during this depressurization.
Upon completion of the first cycle, the RO was immediately aware that he had-terminated the depressurization too i.
soon.
Fifteen seconds following the termination of the first cycle, I
the PORV was opened a second time.
This cycle terminated at the indicated time.-
(3)
This crew completed the depressurization with one open/close cycle as compared to the 9/05/90 crew.
I
-(4)
The value indicated is an observed value-and is noted such as the i
parameter was not recorded on computer printout. All values on 8/29/86 were observed.
( 5.\\
Following SI termination, EMG E-3 allows for further RCS depressurization using pressurizer PORVs in order to equalize pressures and stop break flow.
This final depressurization was started at 29.4' minutes and stopped at 30.57 minutes.
During this depressurization, flow reversed and pressure equilibrated.
(6)
Following SI termination.
EMG E-3 allows for further RCS depressurization using pressurizer PORVs in order to equalize pressures and stop break flow.
This final depressurization was started at 29.54 minutes and stopped at 30.96 minutes.
During this depressurization, flow reversed and pressure equilibrated.
1