ML20029E112
| ML20029E112 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/12/1994 |
| From: | Feigenbaum T NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NYN-94056, NUDOCS 9405160275 | |
| Download: ML20029E112 (7) | |
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O }I North Atlantic Energy Service Corporation g
P.O. Box 300
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. Atlantic Seabro Ic, NH 03874 gp (603) 474-9521, Fax (603) 474 2987 The Northeast Utilities System Ted C. Feigenbaum Senior Vice President &
Chief Nuclear Ofhcer NYN 94056 May 12,1994 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Document Control Desk
References:
(a)
Facility Operating 1.icense NPF-86, Docket No. 50-443 (b)
North Atlantic Letter NYN-92021, dated February 21,1992, " Steam Generator Tube Rupture Operator Response Times", T.C. Feigenbaum to USNRC (c)
North Atlantic I etter NYN-91061, dated April 16, 1991, " Analysis of a Postulated Design Basis Steam Generator Tube Rupture for Seabrook Station",
T.C. Feigenbaum to USNRC
Subject:
Steam Generator Tube Rupture Operator Response Times Gentlemem North Atlantic Energy Senice Corporation (North Atlantic) submitted a report entitled " Analysis of a Postulated Design Basis Steam Generator Tube Rupture for Seabrook Station" on April 16,1991
[ Reference (c)]. The operator response times assumed in the Steam Generator Tube Rupture Analysis (SGTR) were detised from times observed during SGTR simulation runs on the Seabrook Station plant-specific training simulator and by plant walkdowns. The simulations were perfbrmed by two operating crews using plant-specific emergency operating procedures and design basis SGTR scenario assumptions.
The NRC subsequently requested that North Atlantic provide further validation of the operator response times assumed in the plant specific SGTR analysis by conducting additional operating crew design basis SG'lR simulations to ensure that at least live of the six operating crews have participated.
North Atlantic committed to the NRC on February 21,1992, [ Reference (b)], that at least three additional operating crews would perfbrm a design basis SGIR s:mulation during their March 1993 requalification training. North Atlantic also committed [ Reference (b)] that in the event that the observed operator action times do not support the times assumed in the current design basis SGTR analysis, that the analysis would be revised and resubmitted to the NRC.
9405160275 940512
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United States Nuclear Regulatory Commission May 12,1994 Attention:
Ikscument Control Desk Page two Three operating crews participated in SGTR simulator scenarios in March 1993 The response times of the crews in mitigating the SGTR were evaluated and compared to the response times assumed in the design basis SGTR analysis, Some operator response times were within the SGTR analysis assumptions while others were not. Overall, the operator response time required to isolate primary to secondary leakage was observed to be less expeditious than observed in the same scenario when performed in late 1988/carly 1989. The design basis SGTR analysis submitted on April 16,1991, [ Reference (c)],
credits operator response times observed in the late 1988/early 1989 simulations The safety significance of the observed March 1993 SGTR operator response times was evaluated by North Atlantic to determine their effect on the conclusions in Reference (c) regarding margin to Steam :
Generator overfill and offsite radiological doses. The Reference (c) conclusion that Steam Generator overfill will not occur for a postulated design basis SGTR at Seabrook Station remains valid.
Additionally, the March 1993 SGTR operator response times were determined to cause no additional steam releases lbr the case oflimiting offsite doses, therefore the limiting offsite doses in Reference (c) are not affected. The evaluation of the March 1993 SG~l R operator response times and their effect on the design basis SG I R analysis is enclosed.
North Atlantic has reviewed changes which have been made to Emergency Operating Procedure li-3 " Steam Generator Tube Rupture" and has determined that there have been no changes that can be attributed to the increased operator response times. North Atlantic believes that'lhe increased operator response times are due to an increased emphasis placed on communications and self checking.
Should you have any questions regarding this letter, please contact Mr - James M. Peschel, Regulatory Compliance Manager at (603) 474-9521 extension 3772.
Very truly yours,
/
Ted C. Feigenbaum TCF:MDO Enclosure i
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(Jnited States Nuclear Regulatory Commission May 12,1994 Atterition:
Document Control Desk Page three cc:
Mr. Thomas T. Martin Regional Administrator U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Unissia, PA 19406 Mr. Albert W. De Agazio, Sr. Project Manager Project Directorate 1-4 Division of Reactor Projects 1T.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. Antone C. Cerne NRC Senior Resident inspector P.O. Ilox 1149 Seabrook, Nil 03874
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North Atlantic May 12,1994 1
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ENCI.OSliRE I EVALUATION OF STl!AM GliNERATOR TUBE RUPTURE OPERATOR RESPONSE TIMES B ACKGROllND INFOR MNilON Three operating crews participated in Steam Generator Tube Rupture (SGTR) simulator scenarios in March j
1993. The response times of the crews in mitigating the SGTR were evaluated and compared to the response times assumed in the design basis SGTR analysis. Some operator response times were within the SGTR analysis assumptions while others were not. Overall, the operator response time required to isolate primary to secondary leakage was observed to be less expeditious than observed in the same scenario when performed in late 1988/carly 1989. The design basis SGTR analysis submitted on April 16,1991,[ Reference (c)), credits operator response times obser ed in the late 1988/early 1989 simulationc The safety signilicance of the observed March 1993 SGTR operator response ti,nes was evaluated by North Atlantic to determine the effect on the conclusions in Reference (c) regrrding margin to steam generator (SG) overfill and offsite radiological doses. The results of the evaluat:on are enclosed herein.
SUMMARY
Some operator response times were within the assumptions of Yankee Atomic Electdc company (YAEC)
Report 1648, " Analysis of a Postulated Design Basis Steam Generator Tube Rupture For The ScaNook Nuclear Power Plant"(YAEC-1698) while others were not. Overall, progress in Emergency Procedure E-3, " Steam Generator Tube Rupture", towards stopping the primary to secondary leakage was observed to be less expeditious than observed in videotapes of the same scenario taken in late 1988/early 1989.
Analysis of the design basis SGTR documented in YAEC 1698 credits operator actions using response times based on 1988/1989 videotapes.
The safety significance of the observed operator response times has been evaluated and the conclusions of YAEC-1698 regarding no steam generator overfill and acceptable offsite doses remain valid.
DISCUSSION In March 1993, three operating crews were videotaped responding to the same SGTR scenario:
a)
A double-ended tube rupture in the "C" steam generator; b) initiated from full power; c)
With loss of olisite power occurring at the time of reactor trip; i
d)
The Radiation Data Monitoring System is inoperative; and;
.I c)
One control rod remains stuck out.
'I he videotaped scenarios are nearly identical to the limiting design basis SGTR fbr SG overfill identilied in YAEC-1698. 'Ibe scenarios did not include the limiting single failure, i.e., one intact SG Atmospheric Steam Dump Valve (ASDV) fails to open and plant cooldown proceeds with only two ASDVs. The
omission of this single failure did not affect the value of the scenarios for evaluating operator response times.
Each crew succeeded in early identification of a rupture in the "C" SG. Transition to Emergency Operating Procedure E-3 was prompt. Once in E-3, progress towards stopping the primary to secondary leakags was to be less expeditious than obsersed in videotapes of the same scenario taken in late 1988/early 1989. Analysis of the design basis SG1R documented in YAEC-1698 credits operator actions using response times based on the 1988/1989 videotapes. The operator response times taken from the data plots and videotapes made in March 1993 are shown in the table below:
Scenario No/Date YAEC-1698 Actionilnterval Description +
Analysis l
2 3
Assumption 3/11/93 3/18/93 3/25/93 maw..
Ruptured SG Narrow Range
~25%
-24%
~29%
33 %
Level Q time of isolation Interval between MSIV closure 12:07 5:02 9:26 5:00 and start of RCS cooldown Ruptured SG NR level at the start
~76%
.46%
~49%
~64%
of RCS cooldown Interval between end of RCS 2:26 2:15 2:44 2:00 cooldown and start of RCS depressurization Interval between end of 2:18 2:11 2:47 5:00 depressurization and Safety injection termination Interval between end of RCS
.5 :07 3:53 15:00**
depressurization and a second attempt if required Maximum ruptured SG NR level
~98%
~68%
-83%
Off-scale during the scenario I
Time mtervals are in minutes and seconds.
A second depressurization was not required.
These results conllrm :he observation that progress towards stopping the primary to secondary leakage was obsened to be less expeditious than observed in videotapes of the same scenario taken in late 1988/carly 1989, Two operator action inten als are seen to be inconsistent with the YAEC-1698 analysis assumption.
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sal:ETY EVAL.11ATION 1
The safety significance of the observed response times on the SGTR analysis in YAEC-1698 was evaluated for both margin to SG overfill and limiting offsite radiological doses. The bounding evaluation is documented in Calculation SilC-619. " Verification of Operator Response Times During a SGTR
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Scenario" and assumes the slowest response times between the three crews and YAEC-1698 analysis value for each response intenal shown in the table above. YAEC-1698 and SilC-619 are available for NRC review at Seabrook Station.
These changes to the assumed opsrator response times in YAEC-1698 cause a reduction in the calculated margin to SG overfill. The YAEC-1698 conclusion that SG overfill will not occur for a postulated design basis SGTR at Seabrook Station remains valid. These changes to the assumed operator response times in YAEC-1698 cause no additional steam releases for the case of limiting offsite doses. The limiting offsite doses in YAEC-1698 are not affected.
The conclusion of YAEC-1698 regarding no SG overfill and acceptable offsite doses remains valid with the observed operator response times.
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