ML20046B104
| ML20046B104 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/30/1993 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-93-003, IEB-93-3, NUDOCS 9308030106 | |
| Download: ML20046B104 (7) | |
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Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesot'a 554011927 Telephone (612) 330-5500 l
July 30, 1993 10 CFR Part 50 Section 50.54(f)
U S Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555 MONTICELLO NUCLEAR GENERATING P1 ANT Docket No. 50-263 License No. DPR-22 Response to NRC Bulletin 93-03: Resolution of Issues Related ta Reactor Water Level Instrumentation in BWRs This letter is being provided to satisfy the reporting requirements contained in NRC Bulletin 93-03, titled " Resolution of Issues Related to Reactor Vater Level Instrumentation in BWRs", dated May 28, 1993.
The bulletin requires that by July 30, 1993, all addressees must submit a written report providing:
(a) the description of the short term compensatory actions taken, and i
i (b) a description of the hardware modifications to be implemented at the next cold shutdown after July 30, 1993.
If the addressee chooses not to make the requested actions specified in the Hardware Modifications section, the report shall contain a description of the proposed alternative course of action, the schedule for completing it, and a justification for any deviations from the requested actions.
Our responses to the above items are contained in Attachment 2 to this letter.
Please contact Terry Coss, Sr Licensing Engineer, at (612) 295-1449 if you require additional information.
i This letter contains the following new NRC commitments:
j 1.
We will complete at least those portions of the installation work that can only be done while shutdown (such a:- the piping tie-ins to the reference legs) at the first cold shutdown after August 31, 1993.
2.
Any modification installation work remaining will be completed within 30 days of restart from the above cold shutdown, which will provide a hard piped manual backfill capability.
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3.
Assuming that all technical concerns can be resolved satisfactorily by the end of this year, we will test the new backfill system and have it fully operable by the completion of the first cold shutdown after December 31, 1993.
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9308030106 930730
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PDR ADOCK 05000263 W O
PDR y
i USNRC-Northem Ststes Power Company j
July 30, 1993 j
Page 2 r
4.
As required by Bulletin 93-03, we will' submit a followup report within 30 days of completion of the requested hardware modifications (i.e., after all necessary installation, testing and evaluation work is completed and the new equipment is declared operable) confirming completion and describing the modification implemented.
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,i Roger O Anderson Director Licensing and Management Issues cc: Regional Administrator-III, NRC l
NRR Project Manager, NRC Resident Inspector, NRC
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State of Minnesota, i
Attn: Kris Sanda i
J Silberg i
Attachments:
(1)
Affidavit to the US Nuclear Regulatory Commission (2)
NSF Response to Reparting Requirements 2(a) and 2(b) l 4
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UNITED STATES NUCLEAR REGULATORY COMMISSION l
NORTHERN STATES POWER COMPANY MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263 Response to NRC Bulletin 93-03: Resolution of Issues Related to Reactor Water Level Instrumentation in BWRs l
Northern States Power Company, a Minnesota corporation, hereby provides the information requested by NRC Bulletin 93-03: Resolution of Issues Related to Reactor Water Level Instrumentation in BWRs.
This letter contains no restricted or other defense information.
NORTHEPJj STATES POWER SOMPANY su h ldD 1 -
By '
Moger 0' Anderson Director Licensing and Management Issues On this 80 ay of 4
/ki 8 before me a notary public in and forsaidCounty,pers/nalljappearedR.oger0 Anderson, Director, Licensing and Management Issues, and being first duly sworn acknowledged that he is authorized to execute this doeuinent on behalf of Northern States Power Company, that he knows the co atents thereof, and that to the best of his l
knowledge, information, and bell?f the statements made in it are true an that i
it is not interposed for delay.
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- F MARCA K. LaCORE NOTARY PUBLIC--MihNESOTA HENNEPIN COUNTY q
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. Attachment 2'
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Jvly 30, 1993 1
Page 1 i
NSP Response to Reporting Requirements 2(a) and 2(b) of NRC Bulletin 93-03 The following describes the actions that have been'taken to date or are i
planned in response to NRC Bulletin 93-03:
NRC Bulletin 93-03 Reportinn Reauirement; 2.
"By July 30, 1993, all addressees musc submit a written report i
providing:
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(a) the description of the short term com,pensatory ections N
taken, and"
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-i NSP Response: The following is a brief summary of the short term i
compensatory actions taken to satisfy the speci c requirements of the
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bullet;n:
j Bulletin Ecouirement 1.fa)(1): " Establish enhanced monitoring of all RPV le vol instruments to provide early detection of level '
i anomalies associated with degassing from the reference legs."
Action Taken: The plant process computer monitoring of Reactor Pressure l
Vesssi 'P.PV) level has been enhanced. The process computer will
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corcinuoacly monitor and compare the RPV level indication signals from tho five (.;) reference legs. A computer alarm signal will be activated j
if a kPV safeguards reference leg (i.e., one of the two reference legs I
that provide level indication for the scram, isolation, ECCS initiation, j
and ATWS logic) compen c'ed level signal deviates from the validated reactor level indicat order to alert control room operators of I
This will also serve tc alert operators potential RPV level ar.
t of any level indication ta.smalies that may be the result of degassing so that appropriate actions can be taken.
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Bulletin Recuirement 1. (a ) (2 ) : " Develop enhanced procedures or
-i additional restrictions and controls for valve alignments and l
maintenance that have a potential to drain the RPV during wode 3."
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6,ction Taken: Automatic valve interlocks to prevent RPV drain-down are -
already in place at Monticello. In addition, a precaution in the j
shutdown cooling mode startup procedure alerts the operator that valve l
misalignment can result in a rapid loss of RPV coolant. Additional restrictions have been added to the applicable procedures on shutdown cooling. mode, torus cooling mode, and RHR to radwaste mode to further s.,tial to drain the RPV.
limit activities that have s.
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j Epiletin Recuirement 1. la) (3 ): " Alert operators to potentially l
confusing or misleading level indication that may occur during j
accidents or transients initiating from mode 3 (hot shutdown).
i For example, a drain down event could lead to automatic initiation of ECCS without autumatic system isolation or low pressure system actuation".
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Action Taken:
The operations group at Monticello is sensitized and
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knowledgeable concerning the effect of degassing during rapid depressurization and during normal cooldown.
The'small notching j
observed at Monticello during the cooldown for the '1993 refueling outage has been discussed with operations. A summary of the WNP-2 event and how it affects Monticello has been provided to the operations group.
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Bulletin Reauirem~nt 1.(b): "By July 30, 1993 each licensee is l
requested to complete augmented operator training, on loss of RPV l
inventory scenarios during mode 3, including RPV drain-down events i
and cracks in piping. "
I Action Taken: Augmented operator training satisfying the above
.l requirements has been completed.
The effect of notching on RPV loss of inventory scenarios in mode 3 (hot shutdown) was discussed in detail during this training. A1 abnormal operating procedure has been prepared l
to give guidance to the operator in the event of notching of the l
3 safeguards reference legs. A Safety Evaluation (SR1 93-017) has been l
generated to address the safety significance of the guidance.
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"(b) a description of the hardware modifications to be implemented at the next cold shutdown after July 30. 1993.
If the addressee chooses not to make the requested actions specified in 'the i
Hardware Modifications section, the report shall contain a 1
description of the proposed alternative cource of action, the
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schedule fa: completing it, and a justification for any deviations
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from the requested actions."
i NSP Response: Although we have been aggressively evaluating possible hardware modifications to address the concerns raised by NRC Bulletin 93-03, a final design and firm schedule for this modification have not yet been developed and several options remain under consideration.
At.
i present, we plan to install a modification that will provide continuous backfill capability, supplied by the control rod drive hydraulic system, l
for the two safeguards reference legs and the two fuel zone reference-legs. However, concerns over possible adverse system interactions, such
.j as temperature or pressure perturbations of the safeguards reference legs that may cause spurious trips or otherwise impact instrument i
setpoints, remain to be resolved before we can fully implement this change. We are concerned that premature selection of a specific design 1
to resolve the issues of NRC Bulletin 93-03 could increase the risk of 1
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July 30, 1993 Page 3
.l adverse effects on RPV level indication due to inadequate review.
J Under an expedited modification schedule, we will be prepared to 1
2 complete at least those portions of the installation work that can only l
be done while shutdown (such as the piping tie-ins to the reference l
1egs) at the first cold shutdown after August 31, 1993. Any modification installation work remaining would then be completed with the plant at power within 30 days after restart. Once the modification installation work is complete, the new equipment will provide an l
improved method of quickly backfilling the reference legs on a manual l
basis should the need arise, however the system may not be declared fully operational at that point.
-l We remain concerned that undue haste to select and implement a modification before all of the system interactions are fully understood could inadvertently result in the performance and reliability of the
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existing level instrumentation being degraded rather than enhanced.
Thus, even though installation of the modification may be complete, the
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new equipment will not be declared fully operable and placed into continuous service until a Safety Evaluation is completed confirming it
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is safe to do so.
Pre operational testing requirements have not yet been fully defined, but we expect that special testing requiring a cold shutdown will be needed to verify that the new backfill system can be placed in continuous operation with no adverse impact on the Reactor
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Protection System.
This issue is being considered as part of the BUROG effort and, assuming that all technical concerns can be resolved satisfactorily by the end of this year, we will test the new backfill j
system and have it fully operable by the completion of the first cold shutdown after December 31, 1993.
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There are a number of factors that lead us to believe that the existing i
l reactor water level indication instrumentation installed at Monticello l
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is less susceptible to the phenomenon of notching than comparable instrumentation at other facilities, and that this issue is not as f
significant and therefore less time sensitive at Monticello. These factors can be briefly described as follows:
j (1)
The reference legs that provide level indication for the scram, isolation, ECCS initiation, and ATWS logic (referred to as the l
" safeguards" reference legs) utilize 1" piping as opposed to smaller diameter pipe or tubing used at some other facilities.
BWROC sponsored testing has demonstrated that smaller diameter l
piping or tubing allows void hang-ups and is therefore more l
susceptible to large level errors.
(2)
The scram, isolation, ECCS initiation, and ATUS logic at Monticello is designed such that a failure of one reference leg doe.s not prevent the desired automatic response from occurring.
I In order fcr the notching phenomenon to defeat this logic, both reference legs would have to be affected at the same time by large f
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July 30. 1993 Page 4 and sustained level errors. Although we cannot quantify the l
probability of such a postulated scenario, va do not consider such an event to be credible based on the fact that the two safeguards' reference legs have significantly differing piping lengths and geomecries.
i (3)
During the shutdown /cooldown evolution preceding the 1993
-i refueling outage, special monitoring of all 5 reactor vessel water
-l level reference legs was performed to determine if notching was evident.
The cooldown, which was conducted at the completion of l
an extended run (275 days), was conducted as rapidly as practical without exceeding the Technical Specification cooldown rate.
limitations (100 degree F/hr mavimum) in order to provide the i
conditions under which notchirs was most likely to occur (i.e.,
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rapid depressurization event). Although some level anomalies were observed that could be characterized as notching, none appeared until reactor pressure was below 23 psig, all~were of short duration, and no significant notching occurred on the safeguards reference legs.
(4)
Automatic valve interlocks to prevent draindown events are already in place at Monticello, and relevant procedures have been reviewed and enhanced as appropriate to further control valve manipulations and maintenance activities that have the potential to drain the RPV while in Mode 3.
(5)
Enhancements to operator training and the plant computer. alarm system (which monitors all 5 reference legs and now alerts operators to any level indication anomalies resulting from notching) have been implemented to assure a rapid and appropriate response to a significant safeguards reference leg notch, were one to occur.
In view of tne above, we believe that the existing Monticello reactor water level instrumentation, in conjunction with the compensatory measures described above in the response to item 2(a), is adequate to ensure the continued health and safety of the public and that there is no immediate need for any hardware modifications at Monticello to ensure continued safe operation of the plant. Nonetheless, we will continue to aggressively pursue hardware modifications as an enhancement to the existing reactor water level instrumentation in accordance with the schedule outlined in this submittal.
As required by Bulletin 93-03, we will submit a followup report within 30 days of completion of the requested hardware modifications. (i.e.,
after all necessary installation, testi.ng and evaluation work is completed and the new equipment is declared operable) confirming completion and describing the modification implemented.
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