ML20056E712
| ML20056E712 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/19/1993 |
| From: | Hampton J DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 9308250097 | |
| Download: ML20056E712 (6) | |
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Duke Pouer Com;cny J W Hrsen*
Oconee kdearSite nce(Vendent FG Eka H19 i803.M3499 G7tce Sene'ca SC29G79 (Mt3%3%f fax l
DUKEPOWER August 19, 1993 U.
S.
Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 i
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Suoject:
Oconee Nuclear Station Docket Mos. 50-269, -270, -287 Supplement response to IE Bulletin 80-04 l
1 By letter dated May 27, 1993 Duke Power Company informed the NRC that it had reanalyzed the FSAR chapter 15 Main Steam Line Break (MSLB) transient and found that containment design pressure would be exceeded without operator action to isolate main feedwater.
As discussed in the letter, Oconee sti11 met its design basis acceptance criteria for the MSLB, however, as recommended by IEB 80-04, action is r.equired to limit containment pressurization. Duke described that it was relying on Integrated Control System (ICS) and operator action while evaluating and implementing long term solutions and found this acceptable based on the following:
1 The offsite dose consequences of the MSLB inside containment are bounded by the current FSAR analysis.
The most probable ecenario for MSLB 2ncide containment limits the reactor building pressure to within design l
limits.
The results of a recent analysis indi cat e the Reactor Buildings are capable of withstanding a mean maximum pressure of 144 psig.
Further, it has been determined that no major failure of internal civil structures or components is predictad.
The equipment required to mitigate the consequences of the MSLB is environmentally qualified and would perform its safety function.
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l 9308250097 930819 3
ADOCK05000269{u PDR G
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p Document Control Desh Page 2 August 19, 1993 The estimated frequency of 2E-07 per year for a non-core damage sequence with potential for containment failure is not considered to represent a significant risk to the public.
t Duke also described that a supplement to the original IEB 80-04 response would be submitted within the original Bulletin request of 90 days.
This letter and its attachment are considered to be this supplement.
Very Truly Yours, A
m_
Ham,pton J.
W.
(
xc:
Mr.
L. A. Wiens Office of Nuclear Reactor Regulation U.
S.
Nuclear Regulatory Commission Washington, DC 20555 Mr.
S.
D.
Ebneter Regional Administrator, Region II U.
S.
Nuclear Regulatory Commission 101 Marietta Street, NW.,
Suite 2900 Atlanta, GA 30323 I
Mr.
P.
E.
Harmon Senior NRC Resident Inspector Oconee Nuclear Station l
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j Oconee Nuclear Station Supplemental Response to IE Bulletin 80-04 Analysis of a FWR Main Steam Line Break (MSLB) i Oriainal Actions to be talten by the Licensee; 1.
Review of the containment pressure response analysis to determine if the potential for containment overpressure for a MSLB inside containment included the impact of runout flow from the auxiliary feedwater system and the impact of other energy sources, such as continuation of feedwater or condensate flow.
2.
Review of the reactivity increase which results from a MSLB inside or outside containment.
s 3.
If the potential for containment overpressure exists or i
the reactor-return-to-power worsens, provide a proposed corrective action and schedule for completion of the l
corrective action.
Provide a description of any interim action that will be taken until the proposed corrective action is completed.
4.
Within 90 days of the date of the Bulletin, complete the l
review and evaluation required and provide a written i
response describing the review and actions taken in I
response to each item.
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- 1. Review of Contaimment Pressure Response ADMyps_is:
l A reanalysis identified that the continued addition of main i
feedwater or condensate to the faulted steam generator would result in exceeding the design pressure without Integrated j
Control System (ICS) and operator action. The impact of runout
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flow from emergency feedwater was also evaluated and found not.
to be a
significant contributor to the potential for overpressurization of containment.
The following describes the different cases for main feedwater and condensate isolation with and without ICS and operator actions along with their probabilities of occurrence:
With Credit For Automatic Main Feedyater Controli The station is-normally operated with main feedwater in automatic
- control, therefore this defines the most probable steam line break scenario.
Following the steam line break, the reactor and turbine trip in 2 seconds.
The Integrated Control System (ICS) throttles back 1
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i feedwater flow to zero and then re-initiates feedwater to 4
maintain the minimum post-trip steam generator level.
1 Provided that the operator manually isolates main j
feedwater by closing the main feedwater control valves (20 second stroke time) beginning at 170 seconds, the peak 4
containment pressure will remain below the design pressure of 59 psig.
The estimated probability of this scenario occurring and the operator failing to respond in time to maintain the containment pressure below 59 psig is 1.4E-7 per reactor year.
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6 Without Credit For Automatic Main _Feedvater Co_DiroL o
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With main feedwater flow in manual control, an unlikely operating mode, a high flowrate continues after reactor' trip until feedwater isolation is manually performed by closing the main feedwater control valves.
These valves have a 20 second stroke time.
This action must be initiated at 25 - 30 seconds in order to maintain the peak j
containment pressure below 59 psig.
The estimated probability of this scenario occurring and the operator l
l failing to respond in time to maintain the containment i
pressure below 59 psig is 4.0E-9 per reactor year.
I Without Credit For Automatic Main Fegsly_ater Control and Main Feedwater Control Valve Sticks Onen:
With main feedwater flow in manual control, and assuming i
that the main feedwater control valve sticks open, a high feedwater flowrate will continue af ter reactor trip.
The mi a feedwater block valve (stroke time 120 seconds) must be relied on for isolating feedwater flow.
This
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additional assumption that the main feedwater - control-
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valve c*icks open significantly. reduces the probability of this scenario, which is 4.0E-10 per reactor year.
Even though the probability of this scenario is extremely low, operator response to initiate closure of the main feedwater block valve within 120 seconds will limit the peak containment pressure to approximately 140 psig.
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i With a Loss of Offsite Powel If a loss of offsite power occurs concurrent with the steam line break, the Feedwater and Condensate System pumps trip due to the loss of power.
The operator must isolate feedwater via the main feedwater control valves within 170 seconds in order to maintain the peak containment pressure below 59 psig.
The estimated probability of the operator failing to respond and 2
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maintain containment pressure less than 59 psig is 1.4E-i 10 per reactor year.
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2.
Review o_f React;ivity Increap31 l
Reanalysis of the core response during a
MSLB yielded acceptable results.
3.
Potential for Containment Overpr_efsurizatiomi l
Since the reanalysis identified that the potential for t
overpressurization did exist without ICS and operator actions, i
corrective actions are necessary. What follows is a descrf ption of the proposed corrective actions and schedule for completion l
of the corrective actions. Additionally, justification for the schedule for completion is also provided-.
Corrective Action:
In order to alleviate the reliance on operator action, l
feedwater isolation will be initiated by. an automatic signal during a main steam line break (MSLB). This signal uill actuate feedwater isolation whether the M3LB is inside or outside of containment.
A combination of the l
following signals will be utilized to initiate feedwater isolation:
1)
Reactor trip signal from 4 RPS channels.
2)
Steam generator pressure signals (2
signal per generator).
l The reactor trip signal will be generated by 2 out of the 4 RPS channels and combined with 1 of the 2 steam generator pressure signals (setpoint and rate) to initiate i
the isolation of all main feedwater control valves and the startup feedwater control and block valves along with the i
main feedwater block valve on the affected sta m generator. Both the reactor trip and the steam genere %r pressure signals are generated by safety related i
l instruments. Redundant instruments provide protection from i
failure of any single instrument.
Additionally, the
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failure of an instrument will not cause an inadvertent actuation.
The depressurization rate will be time dependent and the steam pressure setpoint will be armed and disarmed automatically at a
predetermined RCS j
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temperature. In effect, the system will actuate when a reactor trip signal is present, the RCS temperature arming setpoint i ra satisfied, ar.d the etcam generator setpoint and rate permissives are satisfied on the affected steam generator.
The isolation of the steam generators is accomplished by the closing of main feedwater control valves which are air operated valves capable of closing against the full i
discharge pressure of the feedwater pumps.
However, the
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main feedwater pumps will more than likely be tripped from a false high level indication from the faulted steam I
generator. Valve stroke time is 20 seconds and.the. valves are included in the Section XI IST program. The startup i
control valves on the affected steam generator will also l'
isolate and an interlock from the startup valve position will initiate cloeure of the main block valves. These i
valves are also included in the Section XI IST program.
Inclusion of the steam generator pressure signals in the f
instrument surveillance Technical Specification will be submitted r on completion of _ the modifications on all three units.
In the
- interim, a
Selected Licensee j
Commitment (SLC) or a proposed Technical Specification l
will cover each completed unit until all' three are completed.
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Schedule for Corrective Action:
l The following is the schedule for implementation of the i
modifications for the automatic feedwater isolation following a MSLB:
1 Unit 3 -
EOC-15 Refueling Outage - Scheduled to start May
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1995 l
Unit 1 -
EOC-16 Refueling Cutage - Scheduled to start i
October 1995 Unit 2 -
EOC-15 Refueling Outage - Scheduled to start March 1996.
Lustification for Schedule:
The most probable scenario for a MSLB inside containment limits the Reactor Building pressure to within the design limit of 59 psig.
The estimated fregt.ency of those steam line break scenarios which exceed the design limit is 2E-4
07/yr, which is very low.
Furthermore, these accident sequences do not involve core da nage.
The Reactor Buildings are capable of withstanding a mean maximum pressure of 144 psig.
No major failure of' internal civil structures or components is predicted.
The equipment required to mitigate the consequences of the MSLB is environmentally qualified and would perform its safety function.
Based on the above, offsite dose consequences of the MSLB inside containment are bounded by the current FSAR analysis and the event is not considered to represent l
a significant risk to the public.
i In consideration of cost and scheduled outage length a limited nun.ber of modifications can be implemented during each refueling outage. Typically, modification packages l
are prepared well in advance of implement.' tion to allow adequate time for design,
- review, procedural changes associated with operating with the modification, any associated Technical Specifications and equipment procurement. Acceleration-of this schedule requires the delay of current work and the increased costs associated with shortening the lead time.
t Modifications that are required to alleviate a safety I
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- concern, significantly enhance safety, or to meet a license condition can be accelerated through the urgent modification process. Duke Power feels that the costs cf accelerating the schedule for the implementation of these modifications are not justified by the estimated frequency of 2E-07 per year for a non-core damage sequence with the
-i potential for containment failure.
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