ML20058G659
| ML20058G659 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 11/30/1993 |
| From: | Borchardt R Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9312100030 | |
| Download: ML20058G659 (2) | |
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e-November 30, 1993' Docket No.52-001 L
Mr. Patrick W. Marriott, Manager i
Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125
Dear Mr. Marriott:
SUBJECT:
REVISED COMMENTS DUE DATE RE: ADVANCE COPY OF THE DRAFT COMMISSION PAPER, " DIVERSITY IN THE METHOD OF MEASURING REACTOR PRESSURE VESSEL LEVEL IN THE ADVANCED BOILING WATER REACTOR (ABWR) AND SIMPLIFIED BOILING WATER REACTOR" This letter supersedes my letter transmitting the draft Commission paper dated November 22, 1993.
It corrects the date of the ACRS meeting and the due date i
for GE comments from December 15 to December 9, 1993.
i Enclosed is a copy of the subject Commission paper for distribution to appropriate GE staff.
The Office of Nuclear Reactor Regulation staff is planning to present a discussion of the contents of this paper during the December 9, 1993, Advisory Committee on Reactor Safeguards (ACRS) ABWR full i
committee meeting.
Because of our tight schedule, we request that GE provide any written comments it might have on the staff's proposals included in the paper before December 9,1993, so that we may include them in our final version of the paper. We also expect that GE will provide comments on the paper during the ACRS meeting.
In addition it is expected that the full com-mittee will provide comments from the members to include in our final version.
If you need additional information about this issue please contact Chet Poslusny on extension 504-1132.
Sincerely, OMM%n R. W. Borchardt,7 Director Standardization Project Directorate i
Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation l
Enclosure:
DISTRIBUTION:
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Mr. Patrick W. Marriott Docket No.52-001 General Electric Company cc:
Mr. Joseph Quirk Mr. Raymond Ng GE Nuclear Energy 1776 Eye Street, N.W.
General Electric Company Suite 300 175 Curtner Avenue, Mail Code 782 Washington, D.C.
20086 San Jose, California 95125 Mr. Victor G. Snell, Director Mr. L. Gifford, Program Manager Safety and Licensing Regulatory Programs AECL Technologies GE Nuclear Energy 9210 Corporate Boulevard 12300 Twinbrook Parkway Suite 410 Suite 315 Rockville, Maryland 20850 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.
I Washington, D.C.
20460 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004 Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 Mr. Steve Goldberg Budget Examiner i
725 17th Street, N.W.
Room 8002 Washington, D.C.
20503 1
Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874
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NUCLEAR REGULATORY COMMISSION e
j WASHINGTON, D C. 20h0001 November 15, 1993 MEMDRANDUM FOR: The Chairman Comissioner Rogers Comissioner Remick Comissioner de Planque FROM:
James M. Taylor Executive Director for Operations
SUBJECT:
ADVANCE COPY OF THE DRAFT COMISSION PAPER, ' DIVERSITY IN THE METHOD OF MEASURING REACTOR PRESSURE YESSEL LEVEL IN THE-ADVANCED BOILING WATER REACTOR AND SIMPLIFIED BOILING WATER REACTOR
- To keep the Comission informed about key policy and technical issues bearing on future light-water reactor designs, I an enclosing a draft copy of a Comission paper proposed on the above subject. The staff believes that it is appropriate to have public and industry comments on the issue discussed in this paper before requesting final Comission approval of the staff's posi-tion.
The staff intends to issue the final safety evaluation report (FSER) for the-advanced boiling water reactor (ABWR) at the end of this year. Therefore, it may not be possible to issue this draft paper as a SECY paper before forward-ing the ABWR FSER.to the Comission.
In that event, the staff will request-the Comission's guidance on this matter through the FSER.
CONTACTS:
George Thomas, NRR 504-1814 Chet Poslusny, NRR 504-1132
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Enclosure
The Comissioners,
The staff intends to place the enclosed draft paper into the public document room after three (3) working days from the date of this memorandum.
The staff vill provide the Advisory Comittee on Reactor Safeguards (ACRS) and GE Nuclear Energy with a copy of the enclosed paper in preparation for discussion at the December ACRS meeting.
Origi..sl signed by James M. Taylor James M. Taylor Executive Director for Operations
Enclosure:
As stated cc:
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James M. Taylor j
Executive Director for Operations
SUBJECT:
DIVERSITY IN THE METHOD OF MEASURING REACTOR PRESSURE VESSEL LEVEL IN THE ADVANCED BOILING WATER REACTOR AND SIMPLIFIED BOIL!NG WATER REACTOR PURPOSE:
i To request Commission approval of the staff recommendation that diversity of reactor pressure vessel (RPV) water level measurement be required in the advanced boiling water reactor (ABWR) and simplified boiling water reactor (SBWR).
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BACKGROUND:
During the TMI-2 accident, low water level in the reactor vessel and inadequate core cooling (ICC) were not recognized for a considerable. time.
Insufficient instrumentation was one of the causes of this problem.
TMI-2 Action Item II.F.2, " Instrumentation for Detection of. Inadequate Core Cooling," required licensees to consider additional instrumentation in order to provide an unambiguous, direct method to interpret indication of-ICC. The BWR Owners' Group submitted the S. Levy reports, SLI-8211 and SLI-8218, in l
response to the action item.
In these reports, the BWR Owners' Group identified and considered methods for measuring RPV water level other than the differential pressure (dp) measurement currently and exclusively used in i
operating BWRs and proposed for the ASWR and SBWR. The evaluation was l
. performed based on backfit considerations for the operating plants.
It was determined, by both staff and industry, that backfitting diverse level l
CONTACTS:
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George Thomas, SRXB/DSSA/NRR 504-1814 i
Amy Cubbage, SRXB/DSSA/NRR
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The Comissioners.
instrumentation on operating BWRs would not provide sufficient incremental safety benefit to justify the cost. The ABWR and SBWR were not considered in the study since their design had not yet been developed.
It should be noted that the discussion which follows is based primarily upon the ABWR design.
This is because the staff has not completed a detailed review of the SBWR design. However, the basic logic regarding level instrumentation importance, ccmon cause failure potential, and the desirability of diverse level instrus.entation applies to the SBWR as well as the ABWR.
DISCUSSION:
The reactor vessel water level instrumentation in the ABWR will be used for both non-safety-related normal functions (controlling feedwater) and safety-elated functions (scram, containment isolation, and emergency core cooling systems (ECCS) actuation).
It also provides operator's with information necessary for act"ons to ensure adequate core cooling in accordance with the emergency operating procedures. Many of the actions in the emergency procedures are keyed to reactor water level and many safety-related operator actions are taken to ensure adequate reactor water level.
As in operating BWRs, the level instruments for the ABWR RPV are all dp instruments.
Each instrument uses a reference leg, which is maintained full by a condensing chamber connected directly to the steam space in the RPV, and uses a variable leg which is connected to the RPV water space. All the dp level instruments operate on the same physical principle. Therefore, comon-cause failures caused by a design deficiency or saintenance error could result in inaccurate indication of reactor vessel water level.
The staff's concern is based on experience with potential comon-mode failure mechanisms in the reactor water level instruments.
For example, during the a
past two years, anomalies have been observed in reactor vessel water level indication at several BWRs (Millstone-1, Pilgrim, LaSalle, and WNP-2) during controlled depressurization to comence plant outages. These anomalies consisted of " spiking' or " notching" of level indication, and in one instance, a sustained error in level indication.
The effect of non-condensible gas in the condensate chamber and reference leg of the cold-leg type of water level instruw nts has been determined to be the root cause of these level indication anomalies. Testing has shown that under depressurization conditions, non-condensible gases can cause significant errors in the level indication. To ensure high functional reliability of the instrumentation, the staff issued Generic Letter 92-04, ' Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)," and Bulletin 93-03, " Resolution of Issues Related to Reactor Yessel Water Level Instrumentation in BWRs,' requesting hardware modifications for operating reactors.
In response to these generic comunications, all BWR licensees have comitted to implement hardware modifications to their level instrumentation systems.
Similar cold-leg instruments are used in the ABWR design.
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Eefere tFe noncondensible gas level inaccuracies, there were other problems
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uith dp level instrumentation used in BWRs.
In 1984, the staff issued GL 84-23, " Reactor Vessel Water Level Instrumentation in SWRs," to address concerns related to high entain: ment temperature during a depressurization event. High containment temperature combined with reactor depressurization can lead to false water level readings as a result of f' lashing or boiling in the reference leg within the containment.
l These known corpon-mode deficiencies in BVR level instrumentation systems have 4
been addressed at operating BWRs and by GE in the ABVR design.
It should also be noted that these particular deficiencies would not have compromised the automatic protective functions of the level instrumentation for accident scenarios initiated while at power, and that no previous incidents at BWRs of l
inaccurate level indication have been misinterpreted by plant operators so as to lead to unsafe actions.
The staff concludes that the ABWR level l
instrumentation system without the proposed level diversity meets the minimum requirements of all applicable General Design Criteria (GDC). However, in view of the importance of level instrumentation for safety in BWRs, and the experience discussed above where the potential existed to fail redundant level instruments due to a conrnon cause, the staff believes that the addition of level instrumentation which operates on a diverse physical principle is desirable and prudent for the purpose of guiding operator emergency actions.
l Since the ABWR and SBWR are currently in the design stage, c tiderable flexibility exists in addressing this issue. Additionally, since these are advanced plants, a more robust solution to potential vessel level inaccuracies is appropriate. Other evolutionary designs, such as ABB-Combustion 1
Engineering (CE) System 80+, provide diverse methods of RPV level measurement.
The inadequate core cooling instrumentation package in CE System 80+ includes reactor vessel level monitoring system probes employing both dp sensors and the heated junction thermocouple concept. The staff is aware of a diverse
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method of level monitoring that is currently in use in at least one nuclear power plant in Germany employing ultrasonic measurement technioues.
In addition, a diverse level measurement system which uses heated junction i
thermocouples has been in use for the past five years at a Swedish BWR, and another Swedish BWR uses float switches for diverse level indication and i
automatic systems actuation. Other Swedish BYRs have decided in principle to l
install diverse level measurement systems.
GE does not agree with the staff recocnendation and has presented its position in a letter dated October 26, 1993.
As part of the letter, GE presented the follcwing sumary:
'AEWR water level instrumentation is rugged, simple and highly redundant for failure tolerance. All known operating problems have been addressed in this design and it is incredible to postulate simultaneous comon-l mode failures which would yield identical errors in all the dp instrumentation. Alternate technologies are unqualified for this application; further, there is no need to add this complexity, since the i
plant operating staff has ample additional indications of an impending
The Commissioners '
I problem without relying solely on water level.
The EPGs direct the operator to use all information available to him and make. conservative (safe) decisions.'
s In the attachment to the letter, GE also provided a list of indications of l
inadequate RPY water level which are independent of the dp RPV water level instrumentation.
The staff recognizes that other parameters could aid the I
operator in assessing the adequacy of core cooling under accident conditions.
These include instrumentation for indication of reactor power, core neutron flux, the recirculation flow control systes response, and feedwater flow and steam flow mismatch. However, the staff believes that these indications cculd be easily misinterpreted or could be insufficient because they are only indirect methods of inferring reactor water level or core cooling.
The diverse method of level measurement is recommended for indication in the control room only (there is diverse instrumentation, namely high drywell pressure, in both the operating BWRs and the ABWR design which provides diverse signals for automatic safety systems actuation for many event scenarios). This would provide a direct and back-up means for the operator to i
identify inadequate core cooling and to take appropriate manual actions to initiate and control safety systems as identified in the plant emergency operating procedures. The staff reconnends that the diverse level measurement device be reliable, redundant, and capable of being powered by on-site power sources.
As noted above, GE does not agree with the staff that diverse level indication should be required.
However, during a meeting between the staff and GE on October 14, 1993, design requirements for diverse level indication, should it a
be required, were briefly discussed. Table 1, enclosed, provides a preliminary list of design requirements developed from this meeting. This list requires significant development, and the staff is pursuing this further with GE. However, the design requirements do illustrate the overall approach the staff believes to be appropriate for this diverse instrumentation.
SUMM M V:
For advanced reactors like ABWR and SBWR, two independent and diverse methods i
of measuring the RPV level should be required because of the importance of RPV i
level instrumentation to BWRs and because ope ating experience has shown the potential for failure of redundant level instruments due to connon cause.
t COCRDINATION:
The Office of the General Counsel (OGC) has reviewed this paper and has no legal objection. OGC notes that Coanission approval would be tentative, subject to further review in design certification rulemakings, and that comunications with GE regarding this Comission position should state this fact.
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EECDM"ENDATION:
The staff recommends that the Commission (1)-
Egli that this draft paper will be placed into the public document room.
j (2)
H211 that the staff plans to discuss this position with the ACRS.
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(3)
H211 that final Commission approval of the staff's position may be requested through the ABWR FSER review process.-
James M. Taylor Executive Director for Operations
Enclosure:
As stated l
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TABLE I PRELIMINARY ABWR DIVERSE RPV LEVEL INSTRUMENTATION DESIGN REQUIREMENTS 1)
The system shall be highly reliable, t>ut is not required to be safety grade or seismically qualified.
2)
The instrumentation will provide operator information only, including control room indication and alams. The control room indication and alarms could be provided on the operator console visual display units.
No automatic actuation functions will be provided, and indication would not be provided on the remote shutdown panel.
3)
The system shall be dual redundant including separation of power supplies. The instrumentation shall be capable of being powered by on-site power upon loss of offsite power, but it is not required that the power source be Class IE.
i 4)
The instrumentation shall be capable of remaining functional during and i
following Design Basis Accident (LOCA) environments, but the i
requirements of the EQ rule (10 CFR 50.49) would not apply. Generic i
tetter 85-06, " Quality Assurance Guidance for ATVS Equipment that is not Safety-Related," should be used for guidance on survivability and quality assurance of the instrumentation.
5)
The level instrumentation shall be capable of measuring RPV water level from the top of active fuel to the emergency operating procedures control band limit, and could be either full range continuous indication or discrete position indication.
If the design involves discrete i
position indication, 3 or more level indications shall be provided with l
locations ranging from top of active fuel to the emergency operating proceh res control band.
i G)
Accuracy on the order of +/- 10% or +/- I ft and response time on the order of minutes may be acceptable depending upon the location and type of level measuring device used.
l 7)
The instrumentation is not required to be hardwired; the use of the plant multiplerer equipment would be acceptable. To minimize additional I
vessel penetrations, existing instrumentation taps could be used where pessible. Measurement of water level outside the core shroud would be
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acceptable.
8)
The system would not be required to be included in the Technical Specifications, and the operability of the system would be controlled by the operational reliability assurance program (ORAP).
ENCLOSURE
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