ML19248D478
| ML19248D478 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 07/23/1979 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Brian Lee COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 7908160309 | |
| Download: ML19248D478 (19) | |
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Docket tos: 50-373 and 50-374 JUL 2 31979 Mr. Byron Lee, Jr.
Vice President Commonwalth Edison Company P. O. Box 767 Chicaco, Illinois 60690 Dea r Mr. Lee:
SUBJECT:
ADDITIONAL INFORMATION REQUIRED FOR NRC STAFF GENERIC REPORT ON BOILING WATER REACTORS On June 28, 1979, the PRC staff met with representatives from each of the licensees of boiling water reactors (BWR's) as well as the applicants for near-term operating licenses for BWR 's.
At that meeting we discussed our short-term program for reviewing the implications of the Three Mile Island Unit 2 accident on operating BWR 's and near-term Operating License applica-tions for BWR's.
At the meeting we held a discussion of c general informa-tion needs and noted that our review will concentrate on two basic areas, i.e., systems and analysis.
We stated that we would provide you with our formal requests for infor.etion at a later date., which consists of three attachments, contains our requests for additional information in the systems area. Enclosure 2 contains our requests for additional information in the analysis area.
In order for us to maintain our schedule we request that you provide clear and complete responses to the enclosed requests by August 27, 1979.
If you cannot meet this schedule or if you require any clarification of these matters please contact William F. Kane, (301) 492-7745 immediately.
Si ncerely, K b. L Olan D. Parr, Chief Light Water Reactors Branch tb. 3 Division of Project Management
Enclosures:
1.
Request for Additional Information (Systems Area) 2.
Request for Additional Inf ormation
( Analysis Area) cc w/ enclosures:
See next page 63012.9 7908160 3 09 e
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JUL 2 31979 Mr. Byron Lee, Jr. m cc: Richard E. Powell, Esq.
! sham, Lincoln & Beale One First National Plaza 2400 Chicago, Illinois 60670 Dean Hansell, Esq.
J-Assistant Attorney General 2
State of Illinois 188 West Randolph Street Suite 2315 Chicago, Illinois 60601 E
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e ENCLOSURE 1 REQUESTS FOR ADDITIONAL INFORMATION BULLETINS & CRDERS SYSTEMS GROUP Y
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Information on Systems Caoable of Providing Post-Accident and Transient Core Cooling Instructions Table I is intended to be an all inclusive list of the systems that are capable of providing post-accident and transient core cooling for all cypes of BWRs.
However, if your plant has additional or alternate systems that provide core cooling, that have not been specifically identifiec. they should be included in your submittal.
Table II contains a list of information that should be provided as applicable, for the systems identified in Table I. 'The information that only requires a yes/no answer has been identified. As noted on the table s:me of the information may be provided by utilizing drawings, however, the drawings must be large enough to be clearly 1 gible, the systems and components ma-ked (particularly if plant,P&ID drawings are used), and drawing legends provided where needed.
If questions arise pertaining to the interpretation at tne type of information requested contact Byron Siegel (301-492-7341) or Wayne Hodges (301-492-75SS).
NOTE:
We are aware that much of the information we are requesting may have already been submitted on your docket.
However, in orcer to excedite our review, we are requesting that you com?ile anc resubmit ne information in this attachment.
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Table I Systems for which information is requested 1.
Reactor Core Isolation Cooling System (RCIC) 2.
Isolation Condenser 3.
High Pressure Core Spray System (HPCS) 4.
High Pressure Coolant Injection System (HPCI) 5.
Low Pressure Core Spray System (LPCS) 6.
Low Pressure Coolant Injection System (LPCI) 7.
Automatic Depressurization System (ADS) 8.
Residual Heat Removal System (RHR) including Shutdown Cooling, Steam Condensing, Suppression Pool Cooling and Containment Spray Modes 10.
Standby Coolant Supply System 11.
Reactor Closed Cooling Water System 12.
Control Rod Drive System 13.
Condensate Storage Tank 14.
Main Feedwater System 15.
Recirculation Pump / Motor Cooling Systems 9
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Table II Infomation on Systems Capable of Providing Post-Accidenc and Transient Core Cooling General System Design Infor: nation
- Safety Classification & Seismic Category
- Plant Steam By-Pass Capacity
- Potential of Systems & Component Flooding (i.e., injection of water from CST in excess of Technical Specification min.) and Separction of Trains
- Nomal Position of Valves, Indication Location Direct 1
or Indirect Indication l
- Failed State of Each Valve 1
- Normal Power Sources for System Operation 1
- Normal Power Sources for Support System Operation, e.g., lube oil, lube oil cooling, ventilation
- Systems and Components Shared Setween Units
- Air Sources for Pneumatic Valves, Cycling Capacity & Aicernate Sources
- Number of Safety & Relief Valves & Relieving Capacity
- Relief & Safety Valve Setpoints
- System Trips
- Methods of Cooling System, Components (i.e., pumps, valves)
System Activation
- Automatic Startup Logic (initiation signals) & Power Source
- Automatic Sequencing Back onto Diesel Following Reset (Yes/No)
- Auto Initiation Overriding Capability
- Auto Initiation Built in Time Delay
- Manual Initiation Capability, Procedure. Tin >a Req'd, Locations, Manpower Reg'd
- Potential Cormonalities with Control Systems
- System Interlocks & Diversion
- Operator Actions Required for System Operation & Control o
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2 Water Sources
- Safety Classification & Seismic Classification
- Primary Water 50urce, Total & Dedjcated Capacity, Time Available
- Seconoary and Backup Water Sources, Automatic / Manual, Procedure, Time, Req'd
- Strainers in System and Location Power Source
- Number of Trains
- Pumps Connected to Diesel Generators
- AC & DC Sus Arrangement for Trains
- Loss of Offsite Power - System Response Operator Acticn, Time Req'd
- Loss of On-site AC Fower - System Response Operator Action, Time Reg'd
- Loss of All AC Power - System Response, Operator Action, Time Reg'd Instrumentation & Control
- Safety Classification & Seismic Category
- Automatic and Manual Control from Control Room (Yes/N)
- Alarms Lt.2ted in Control Room
- System Indications Located in Control Room' (pump, valves, level etc.)
- Retmte Control Panels
- Methods of Detecting Leaking Safety / Relief Yalves (i.e., leaking bellows, unseated valve)
Testing /Techr:ical Specifications
- Limiting Conditions for Operation
- Frequency of System & Component Tests 1
- System Testing Lineups 1
- System Bypass and/or Test Loops
- Bethod of Verification of Correct Test Lineup and Restoration to Norml Condition
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- Allowable 5ystem Outage Times
- System & Jomponentional Testing Following Mair.;enance
- Componen ts ibt Periodically, Tested.
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- Auto Override During Tests
- Other Components or System Affected by Tests If May be provided by a drawing b30.b.'A6
Information Needed for Containment Isolation System I.
For each fluid line and fluid instrument lines penetrating tr.e containment, provide a table of design information regarding the containment isolation provisions which include the following infon:ation:
a.
Containment Penetration number; b.
System name; c.
Fluid contained; d.
Engineered safety feature system (yes or no);
e.
Figure showing arrangement of containment isolation barriers; f.
Isolation valve number; g.
Location of valve (inside or outside containment);
h.
Yalve type and operation; i.
Primary made of valve actuation; j.
Secondary mode of valve actuation; k.
Normal valve position; 1.
Shutdown valve position; m.
Postaccident valys position; n.
Power failure valve position; Containment isolation signals, including parameters sensed and their o.
set point; l'
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Yalve closure tire; Power sourcei q.
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Valve position indication (direc' or indirect) a f
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2-II.
Discuss the design requirements _ for the contain ent isciation barriers regarding:
a.
The extent to which the quality standards and seismic design classification of the containment isolation orovisions folicw the recomendations of Regulatory Guides 1.25, " Quality Grcup Classi.fications and Standards for Water, Steam, and Radioactive-Water-Containing Components of Nuclear Power Plants," and 1.29, " Seismic Design Classification";
b.
Assurance of the operability of valves and valve operators in the containment atmosphere under nomal plant operatin; conditions and postulated accident conditions.
Qualification of closed systems inside and cutside the containment c.
as isolation barriers; d.
Qualification of a valve as an isolation barrier;
- e.
Required isolation valve closure times; f.
Mechanical and electrical redundancy to preclude cocoon rcce failures; g.
Primary and secondary nodes of valve actuation e
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3 III.
Discuss the provisions for detecting leakage from a rerote ranually controlled system (such as 'an yngineered sa'fety feature system or essential line) for the purpose of detemining wher to isolate the affecte: system or system train.
Specify the paramters sensed, their set point, and procedure for initiation of containment isolation.
IV.
Discuss the design provisions for testing the operability of the isolaMon,_
valves.
V.
Identify the codes, standards, and guides applied in the design of the containment isolation system and system components.
VI.
Discuss the normal operating nodes and containment isolation provision and procedures for lines that transfer potentially radioactive fluids out of the coritainment.
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Additional Systems and Ocerational Information Re:uired o
I.
Provide copies of the procedures for loss of feecwater and smali cr ar.
LOCA.
II.
Discuss the reactor water level measurement system.
In particular:
1.
Provide a diagram showing location of pressure taps used in reasuring level. The diagram should be detailed enough to show whether the measurement is inside or outside the core shroud.
2.
Describe the instrument piping arrangements and types of transducers used.
3.
Which levels are nonitored in the control room and how are they indicated (i.e., recorders, meters)?
4.
Which measurements provide signals for safety systecs, which for control systems, which for other systems?
5.
Describe the dynamic response of each of the level reasurement and indicating instruments for conditions typical of a small break LOCA.
6.
What are the level measurement uncertainties?
7.
What level difference is expected between core and measurement location for:
a.
normal operations, b.
reactor shutdown with decay heat and with recirculation pumps running, c.
reactor shutdown with decay heat and recirculation pumps not running, and d.
moderate level transient as for a small break LOCA or stuck open SRV.
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E!iCLOSURE 2 REOUESTS FOR ADDITI0tiAL ItiFO MATI0ti BULLETItiS & ORDERS AfiALYSIS GRC'JP e
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REQUEST FOR ADDITIONAL INFORMATION REGARDING SMALL BREAK LOCA ANALYSIS I.
The response of the reactor system of a given plant to a small break i
LOCA will differ greatly depending upon the break size, tr.e location of.the break, mode of operation of the recirculation pumps, number of
%CS systems functioning, and the availability of isolation condensers or RCIC.
In addition, this response may differ for different plants designed by the same NSSS vendor because of differences in the recircu-lation loop configuration or different ECCS designs.
In order for the staff to complete its evaluation of the response of currently operating BWR designs to postulated small break LOCA's, the following information is needed:
(1)
Provide a qualitative description of expected system behavior for (a) a range of postulated small break LOCA's, including the zero break case, and (b) feedwater-related limiting transients combined with a stuck-open safety / relief valve.
These cases should include situations where HPCI and RCIC (or isolation condenser) are assumed available and not available.
The cases considered should also include breaks large enough to (a) depressurize the reactor coolant system, (b) maintain the reactor coolant system at soma intermediate pressure and (c) repressurize the primary system to the safety / relief valve setpoint pressure.
Various break locations in the reactor coolant system should be considered.
(2)
Provide a qualitativc description of the various natural circulation modes of expected system behavior following a small break LOCA.
Discus any ways in which natural circulation can be degraded, such as fluid stratification in the lower plenum caused by inoperation of the cleanur, system.
Assess the possible effects of non-condensible gases.
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II. The following questions pertain to your small break LOCA analysis methods:
(3)
Demonstrate that your current small break LOCA analysis methods are appropriate for applicaticn to each of the cases identified in items
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(7) through (10) below.
This demonstration should inclad an usess-ment of the adequacy of system noding potential counter current flow limitations, and water accumulation above the core.
If, as a result of the above assessment, you modify your analysis methods (e.g., system noding), provide justification for any such nodification.
(4)
Verify the break flow model used for each break flow location analyzed in the response to item (7) below.
(5)
Verify the analytical calculation of fluid level in the reactor vessel for small break LOCA's and feedwater transients.
(6)
Provide integral verification of your small break loss-of-accident, method through comparison with experimental data.
TLTA and LCFT small break tests are possible examples.
III. For each of the analyses requested in Items (7) through (10) below.
(i)
Provide plots of the output parameters specified in Table I of this enclosure.
(ii)
Indicate when the System safety / relief valve would open.
(iii)
Include appropriate information aoout the role of control systems in the course of the transient.
Describe how the system response would be affected by control systems.
(iv)
If the scenario is different for different classes of plants (jet pump, non-jet pump, SWR 4, BWR 5), provide an exampie of each kind.
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. (7)
Provide the results of a sampTe analysis of each type of scall break e.
behavior discussed in the response to item (1) (e.g., cepressurization, pressure hangup, repressurization).
(8)
Provide the results of an analysis of the worst small break size and location in terms of core uncovering assuming a failure in the ECCS and the RCIC (or isolation condenser).
This may be a break wnich does not result in HPCI initiation.
This may require more than cae calcu-l a ti on.
(9)
Provide the results of an analysis for a single stuck open safety / relief valve, and the maximum number of valves that could open folicwing the worst single failure.
(10)
Provide the results of a small break LOCA analysis assuming loss of feedwater.
The case with the worst break location which affords the least amount of tire for operator action should be analyzed.
A single failure in the ECCS and failure of the RCIC (or isolation condenser) shculd be considered.
(11)
Provide a list of transients expected to lift the SRVs; identify the assumed steam and two-phase flow rates througn the valves for these transients.
Provide justification for your assumptions, including the time at which two-phase discharge,if it is calculated to occur, would be experienced.
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(12)
Provide revised emergency procedures or guidelines for the preparation of operational procedures for '.he recovery of plants following small LOCA's.
This should include both short-term and long-tarm situations and follow through to a stable condition.
The suicelines sn;uid include recognition of the event, precautions, acticas, and prohibited actions.
If recirculation pump operation is assumed under two-phase conditions, a justification of pump operability should be provided.
Discuss instru-mentation available to the operator and any instrumentation that might not be reliec upon during these events. What would te the effect of this instrumentation on autor,atic orotection actions?
IV. In addition to the short term requirement identified above, it is requested that the following information be provided by November 1,1979.
(13)
Provide an analysis of the symptoms of inacequate core cooling and required operator actions to restore core cooling. These analyses should include cases assuming the recirculation pum?s are boch operating and not operating.
The calculation snould include the period of time during wh.ich inadequate core cooling is approached as well as the period of time during which inadequate core cooling exists.
The calculations should be carried out far encugn so that all important phenomena and instrument indications are included.
Each case should then be repeated taking credit for correct operator action.
(14)
Provide emergency procedures or guidelines for the preparation of emergency procedures for plant recovery frcm inadequate core cooling.
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. (15) Provide revised emergency procedures or guidelines for the updating of emergency procedures for accidents and transients ccr.sicerad in Section 15 of plant SAR's.
(16) The NRC is planning to perform audit calculations of the SWR small break LCCA.
The necessary computer program input information and comparative calculations should be provided to facilitate this study.
To assist in the review of these cases, we will require computer output in1ormation in excess of that specified in Ta b l e I.
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TABLE 1 Plotted Outout Parameters a
Core:
L, Xyyg,, W, Tclad Reactor Vessel-Lower Plenum:
L, X - or T3gg, P Downcomer:
L, X or TSUo, Leak:
SRV, W, X or Break,W,X_,[Wdt Nomenclative: P - Pressure L - Mixture Level X - Quality T - Temperature W - Mass Flow Rate H - Enthalpy
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