ML19248D474
| ML19248D474 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 07/17/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-15-19, TASK-RR NUDOCS 7908160303 | |
| Download: ML19248D474 (19) | |
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UNITED STATES
[ y s ;.. c ,j NUCLEAR REGULATORY COMMISSION
. y WASHINGTON D. C. 20555
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July 17,1979 Docket No. 50-409
- ,r. Frank Linder General tfanager Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601 Cear Mr. Linder:
SUBJECT:
ADDITIONAL INFORMATIuN REQUIRED FOR NRC STAFF GENERIC REPORT ON B0ILING 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 SWRs. 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 information would be made at a later date.
Enclosure I which consists of three attachments contains our request for additional infomation in the systems area. contains our request for additional infomation in the analysis area.
To maintain our schedule we request that you provide clear and cor.plete responses to the enclosed requ'sts 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) '92-7745 immediately, ci.cerely, 4r-Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors Encl osures:
1.
Request for Additional Infomatio:
(Systems Area) 2.
Request for Additional Infomation
( Analysis Area) cc w/enclasures:
See next page 790816 4
Mr. Frank Linder July 17,1979 cc w/ enclosures:
Fritz Schubert, Esquire Staff Attorney Dairyland Power Cooperative 2615 East Avenue South La Crosse, 'sisconsin 54601
- 0. S. Heistand, Jr., Esquire Morgan, Lewis & Bockius 1800 M Street, N. W.
Washington, D. C.
20036 Mr. R. E. Shimshak La Crosse Boiling Water Reactor Dairyland Power Cooperative P. 0. Bcx 135 Genoa, Wisconsin 54632 Coulee Region Energy Coalition ATTN:
George R, Nygaard P. O. Box 1583 La Crosse, Wisconsin 54601 La Crosse Public Library 800 Mair Street La Cross?, Wisconsin 54601 I
e S
ENCLOSURE 1 REOUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS SYSTEMS GROUP
- 9'JOCdh'
Infomation on Systems Capable of Providing Post-Accident and Transient Core Coolino 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 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 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 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).
NOTE: We are aware that much of the infomation 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.
L,3OCOO
Table I Systems for which information is r quested 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.
R5 actor Closed Cooling Water System
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12.
Control Rod Drive System
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13.
Condensate Storage Ta~nk
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14.
Main Feedwater System 15.
Recirculation Pump / Motor Cooling Systems
Table II Infomation on Systems Capable of Providing Post-Accident and Transient Core Cooling General System Design Infomation 4
- 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 Indirei;t 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 Between Units
- Air Sources for Pneumatic Valves, Cycling Capacity & Alternate Sc,2rces
- 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) & Pcwer Source
- A"tomatic Sequencing Back onto Diesel Following Reset (Yes/No)
- Auto Initiation Overriding Capability
- Auto Initiation Built in Time Delay
- Manual Initiation Capability, Procedure, Time Reg'd, Locations, Manpower Reg'd
- Potential Coctmnalities with Control Systems
- System Interlocks & Diversion
- Operator Actions Required for System Operation & Control 633060
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 Bus 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 Reg'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 Soecifications
- Limiting Conditions for Operation
- Frequency of System & Component Tests 1
- System Testing Lineups 1
- System Bypass and/or Test Loops
- Method of Verification of Correct Test Lineup and uSOO63.
Restoration to Normal Condition T
. - Allowable System dutage Times
- System & Componentional Testing Following Maintenance
- Components tbt Periodically, Tested
- Auto Override During Tests
- Other Components or System Affected by Tests 1/ May be provided by a drawing u3.3Cli2
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 arrangemcnt of containment isolation barriers; a.
f.
Isolation valve number; Location of valve (inside or outside containment);
g.
h.
Valve type and operation;
- i. Primary mode of valve actuation;
- j. Secondary mode of valve actuation; k.
Normal valve position:
1.
Shutdown valve position; Postaccident valve position;
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m.
Power failure valve position; n.
Containment isolation signals, including parameters sensed and their o.
set point;
/
p.
Yalve closure tire; q.
Power sourcei Valve position indication (direct or indirect) r.
U330G3
. II.
Discuss the des #gn requirements for the containment isolation barriers e
reoarding:
The extent to which the quality standards and seismic design a.
classification of the containment isolation provisions follow the recormendations of Regulatory Guides 1.26, " Quality Group 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 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 coninon mode failures; Primary and secondary rmdes of valve actuation g.
U300G4 L
. III.
Discuss the provisions for detecting leakage from a remote ranually controlled system (such as #an yngin'eer4d sa'fety 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 coder, standards, and guides applied in the design of the containment isoir. tion system and system components.
VI.
Discuss the norral operating modes and containment isolation provision and procedures for lines that transfer potentially radioactive fluids out of the contz.inment.
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g)
Additional Systems and Doerational Information Recuired I.
Provide copies 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 shroud.
2.
Describe the instrument piping arrangements and types of transducers used.
3.
Which levels are monitored in the control room 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 dynamic response of each of the level measurement 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, reactor shutdown with decay heat and recirculation pumps not c.
running, and d.
moderate level transient as for a small break LOCA or stuck open SRV.
b39066
ENCLOSURE 2 REQUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS ANALYSIS GROUP e
REQUEST 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 SWR designs to postulated small break LOCA's, the following information is needed:
(1)
Provide a qualitat.ive 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 /reli. f 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 reacter coolant system, (b) maintain the reactor coolant system at some intermediate pressure and (c) repr.ssurize 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 LOCA.
Discuss any ways in which natural circulation can be degraded, such as fluid stratification in the lower pienum caused by inoperaticn of the cleanup system.
Assess the possible effects of non-condensible gases.
i,3CC(18
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 cf 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 ncding), 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 informaticn 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 each kind.
G33CG9
(7)
Provide the results of a sample analysis of each type of small break behavior discussed in the response to item (1) (e.g., depressurization, pressure hangup, repressurization).
(3)
Provide the results of an analysis of the worst small break size and location in terrs 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 ti on.
(9) Provide the results of an analysis for a single stuck cpen safety / relief valve, 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 wnich affords the least amount 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 the SRVs; identify the assumed steam and two-phase flow rates through the valves for these
_transi ents.
Provide justification for your assumptions, including the time at which two-phase discharge,if it is calculated to occur,' would be experienced.
b33070
-4 (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 situations and follow through to a stable condition. The guidelines should include recognition of the event, precautions, actions, 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
- r. )t be relied upon during these events. What would be the effect of this instrumentation on automatic protection actions?
IV. In addi(ion 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 inac' equate core cooling and required operator actions to restore core cooling. These analyses should include cases assuming the recirculation pumps are both operating and not operating.
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 phencmena 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.
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. (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.
(16) The NRC is planning to perform 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 information in excess of that specified in Table 1.
4 63D0'?."1
TABLE 1 Plotted Outout Parameters Core:
L, Xgyg,, W, Tclad Reactor Vessel:
Lower Plenum:
L, X - or TSUB, P 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 Ficw Rate H - Enthalpy
.