ML19248D499
| ML19248D499 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 07/17/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 7908160335 | |
| Download: ML19248D499 (19) | |
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,0, UNITED STATES y 7) e r ' h NUCLEAR REGULATORY COMMISSION WASidlNGTON, D. C. 20555
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July 17,1979 Docket No. 50-155 Mr. David Bixel Nuclear Licensing Administrator Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
Dear Mr. Bixel:
SUBJECT:
ADDITIONAL INFORMATION REQUIRED FOR NRC STAFF GENERIC REPCRT ON BOILING WATER REACTORS On June 28, 1979 the NRC staff met with representatives from each of the licensees of boiling water reactors (BWRs) as well as the applicants for near-tem operating licenses for BWRs. At that meeting we discussed our short-tem program for reviewing the implications of the Three Mile Island Unit 2 accident on operating BWRs and near-tem Operating License applica-tions for BWRs. At the meeting we discussed our general infomation needs and noted that our review will concentrate on two basic areas, i.e., systems and analysis. We stated that fomal requests for infomation would be made at a later date. which consists of three attachments contains our request for additional infomation in the systems area. Enclosure 2 contains our request for additional infomation in the analysis area. To maintain our schedule we request that you provide clear and complete responses to the enclosed requests by August 17, 1979.
If you cannot meet this schedule or if you require any clarification of these matters please contact William F. Kane, (301) 492-7745 immedi ately.
Si ncerely, d
Sy Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors Encl osures:
1.
Request for Additional Infomation (Systems Area) 2.
Request for Additional Infomation (Analysis Area) cc w/ enclosures:
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See next page 7 9 0 816 0 33S~f
Mr. David Bixel July 17,1979 cc w/ enclosures:
Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michilan 49201 Hunton & Williams George C. Freeman, Jr., Esquire
~ ~ P. O. Box 1635 Richmond, Virginia 23212 Peter W. Steketee, Esquire 505 Peoples Building Grand Rapids, Michigan 49503 Charlevoix Public Library 107 Clinton Street Charlevoix, Michigan 49720 te s ? n.,.. -d A-
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ENCLOSURE 1 REOUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS SYSTEMS GROUP L.90CCb
e.
Information on Systems Capable of Providind Post-Accident and Transient Core Coolino Instructions Table I is intended tc be an all inclusive list of the systems that are capable of providing post-accident and transient core cooling for all types of BWRs. However, if your plant has additional or alternate systems that provide core cooling, that have not been specifically identified, they should be included in your submittal.
Table II contains a list of infomation that should be pruvided 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 some of the information may be provided by utilizing drawings, however, the drawings must be large enough to be clearly legible, the systems and components marked (particularly if plant.P&ID drawings are used), and drawing legends provided where needed.
If questions arise pertaining to the interpretation of the type of information requested contact Byron Siegel (301-492-7341) or Wayne Hodges (301-492-7588).
fl0TE: We are aware that much of the information we are requesting may have already been submitted on your docket.
However, in order to expedite our review, we are requesting that you compile and resubmit the information in this attachment.
L33;257
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 G331 S
Table II Infortnation on Systems Capable of Providing Post-Accident and Transient Core Cooling General System Design Information d
- 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 Separation of Trains
- Normal Position of Valves, Indication Location Direct 1
or Indireqt Indication l
- Failed State of Each Valve 1
- Normal Power Sources for System Operation
- Normal Power Sources for Support System Cperation, e.g., lube oil, lube oil cooling, ventilation
- Systems and Components Shared Between Units
- Air Sources for Pneumatic Valves, Cycling Capacity & Alternate 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, Time Req'dl Locations, Manpower Reg'd
- Potential Coninonalities with Control Systems
- System Interlocks & Diversion
- Operator Actions Required for System Operation & Control c332 s
2-Water Sources
- Safety Classification & Seismic Classification
- Primary Water Source, Total & Dedjcated Capacity, Tire Available
- Secondary and Backup Water Sources, Automatic / Manual, Procedure, Time, Reg'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 Action, Time Req'd
- Loss of On-si.te AC Power - System Response Operator Action, Time Req'd
- Loss of All AC Power - System Response, Operator Action, Time Req'd Instrumentation & Control
- Safety Classification & Seismic Category
- Automatic and Manual Control from Control Room (Yes/No)
- Alarms Located in Control Room
- System Indications Located in Control Room (pump, valves, level etc.)
- Remote Control Panels
- Methods of Detecting Leaking Safety / Relief Valves (i.e., leaking bellows, unseated valve)
Testinc/ Technical Specifications
- Limicing Conditions for Operation
- Frequency of System & Component Tests 1
- System Testing Lineups 1
- System by7 ass and/or Test Loops'
- Method of Verification of Correct Test Lineup and Rostoration to Normal Condition b hn cs.:w0 f
-3
- Allowable System butsge Times
- System & Componentional Testing Following Maintenance
- Components Not Periodically, Tested
- Auto Override During Tests Other Components or System Affected by Tests 1/ May be provided by a drawing
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Information Needed for Containment Isolation System I.
For each fluid line and fluid instrument lines penetrating the containment, provide a table of design information regarding the containment isolation provisions which include the following information:
a.
Containment Penetration number; b.
System name; c.
Fluid contained; d.
Engineered safety feature system (yes or no);
Figure showing arrangement of containment isolation barriers; e.
i f.
Isolation valve number; I
Locaticn of valve (inside or outside containment);
g.
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h.
Valve type and operation; 1
i
- i. Primary mode of valve actuation; J.
Secondary mode of valve actuation; k.
Normal valve position; 1.
Shutdown valve position; m.
Postaccident valve position; Power failure valve position; n.
Containmer.t isolation signals, including parameters sensed and their o.
set point; p.
Yalve closure tire; i
q.
Power sources Valve position indication (direct or indirect) r.
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. II.
Discuss the design requirerents for the containrent isolation barriers recarding:
g The extent to which the quality standards and seismic design a.
classification of the containment isolation provisions follow the recommendations of Regulatory Guides 1.26, " Quality Group Classi.fications and Standards for Water, Steam, and Radioactive-Water-Containing B
Components of Nuclear Power Plants," and 1.29, " Seismic Design Classification";
b.
Assurance of the operability of valves and valve operators in the co.itainment atmosphere under normal plant operating conditions and postulated accident conditions.
Qualification of closed systems inside and outside the containment c.
as isolation barriers; d.
Qualification of a valve as an isolation barrier; Required isolation valve closure times; e.
f.
Mechanical and electrical redundancy to preclude common mode failures; Primary and secor.dary modes of valve actuation g.
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. III.
Discuss the provisions for detecting leakage from a remote manually controlled system (such as 'an engineered safety feature system or essential line) for the purpose of determining wher to isolate the affected system or system train. Specify the parameters sensed, their set point, and procedure for initiation of containment isolation.
IV.
Discuss the design provisions for testing the operability of the isolation,
valves.
V.
Identify the codes, strndards, and guides applied in the design of the containment isolatica system arid system components.
VI.
Discuss the nornal operating modes and containment isolation provision ar.d procedures for lines that transfer potentially radioactive fluids out of the containment.
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Additional Systems and Operational Information Recuired I.
Provide copics of the procedures for loss of feedwater and small break LOCA.
II. Discuss the reactor water level measurement system.
In particular:
1.
Provide a diagram showing location of pressure taps used in measuring level. The diagram should be detailed enough to show whether the measurement is inside or outside the core sh roud.
2.
Describe tr.c 'nstrument piping arrangements and types of transducers used.
3.
Which levels are monitored in the c:ntrol 1 om and how are they indicated (i.e., recorders, meters)?
4.
Which measurements provide signals for safety systems, which for control systems, which for other systems?
5.
Describe the dynar..ic response of each of the level measurement ard 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 icvel transient as for a small break LOCA or stuck open SRV.
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ENCLOSURE 2 REQUESTS FOR ADDITIONAL INFORMATION
_B_ULLETINS & ORDERS ANALYSIS GROUP
RE00EST FOR ADDITIONAL INFORMATION REGARDING SMALL BREAK LOCA ANALYSIS I.
The response of the reactor system of a given plant to a small break LOCA will differ greatly depending upon the break size, the location of the break, mode of operation of the recirculation pumps, number of ECCS 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, a.nd (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 av0ilable 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 some 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 qualitative description of the various natural circulation modes of expected system behavior following a small break LCCA.
Discuss any ways in which natural circulation can be degraded, such as fluid stratification in the lower pienum caused by inoperation of the cleanup system. Assess the possible effects of non-condensible gases.
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2 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 application to each of the cases identified in items (7) through (10) below. This demonstration should include an assess-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 modification.
(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 LOFT 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 1 of this enclosure.
(ii)
Indicate when the System safety / relief valve would open.
(iii)
Include appropriate information about 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 i.s different for different classes of plants (jet pump, non-jet pump, BWR 4, BWR 5), provide an example of i
each kind.
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. (7) Provide the results of a sample analysis of each type of small break behavior discussed in.he response to item (1) (e.g., depressurization, pressure hangup, repressurization).
(8) Provide the results of an analysis of the worst sm.all creak si:e and location in terms of core uncovering assuming a failure in the ECCS and the RCIC (or isolation condenser).
This may be a break which does not result in HPCI initiation. This may require more than one calcu-l a tion.
(9) Provide the results of an analysis for a single stuck open safety / relief val.ve, and the maximum number of valves that could open following 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 annunt of time for operator action should be analyzed. A single failure in the ECCS and failure of the RCIC (or isolation condenser) should be considered.
(11) Provide. a list of transients expected to lift tne SRVs; identify the assumed steam and two-phase flow rates through 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 the recovery of plants following small LOCA's. This should include both short-term and long-term situation:
and follow through to a stable condition.
The guidelines should include recognition of the event, precautions, actions, and prohibited actions.
If recirculation pump operaticn is assumed under two-phase conditions, a justification of pump operabili',y should be provided.
Discuss instru-mentation available to the operator and any instrumentation that might not be relied upon during these events. What would be the effect of this instrumentation on automatic p rtection actions?
IV. In addit' ion to the short term requirement identified above, it is requested that the following information be provided by November 1,1979.
(13)
Provide an ana'ysis of the symptoms of inadequite core cooling and required operator actions to restore core cooling. These analyses should include cases assuming the recirculation pumps are both operating and not cperating. The calculation should include the period of time during which 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 enough 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 from inadequate core cooling.
(15) Provide revised emergency procedures or guidelines for the updating of emergency procedures for accidents and transients considered in Section 15 of plant SAR's.
n (16) The NP.C is planning to perfom audit calculations of the BWR small break LOCA. 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 infomation in excess of that specified in Table 1.
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TABLE 1 Plotted Outout Parameters Core:
L_, XAVG., " ' clad Reactor Vessel:
Lower Plenum:
L, X - or TSUB' Downcomer:
L, X or T SUB Leak:
SRV, W, X or Break,W,X_,[Wdt Nomenclative: P - Pressure L - Mixture Level X - Quality T - Temperature W - Mass Flow Llate H - Enthalpy
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