ML19261E523
| ML19261E523 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/17/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| NUDOCS 7908280580 | |
| Download: ML19261E523 (19) | |
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NUCLEAR REGULATORY COMMISSION g
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WASHINGTON. D. C. 20555 July 17,1979
....s Docket No. 50-d 5 Mr. W. G. Counsil, Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company P. 0. Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
ADDITIONAL INFORMATION h nUIRED FOR NRC STAFF GENERIC REPORT 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-term 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 immediately.
i ncerely, aDW At Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors Encl osures:
1.
Request for Additional Infomation (SystemsArea) 2.
Request for Additional Infomation (AnalysisArea) cc w/ enclosures:
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/0 7908280 ti
3 Mr. W. G. Counsil July 17, 1979 cc W/ enclosures:
William H. Cuddy, Esquire Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, N. W.
Washington, D. C.
20005 Northeast Nuclear Energy Company ATTN:
Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 05385 Mr. James R. Himmelwright Northeaet Utilities Service Company P. O. bt.5 270 Hertford, Connecticut 06101 Nuclear Regulatory Commission, Region I Office of Inspection and Enforcement ATTN: John T. Shedlosky 631 Park Avenue King of Prussia, Pennsylvania 19406 Waterford Public Library Ropa Ferry Road, Route 156 Waterford, Connecticut 06385 2022 023
ENCLOSURE 1 REQUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS SYSTEMS GROUP 2022 024 O
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Infonnation on Systems Capable of Brovidini Post-Accident and Transient Core Coolino Instructions Table I is intended to be an all inclusive list of the systems that are canable 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 infonnation that only requires a yes/no answer his 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 () articular)y 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).
?!0TE: 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.
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Table i
?
Systems for which information is requested 1.
Reactor Core Isola + ion Cooling System (RCIC)
'2.
Isolation Condenser 3.
High Pressure Core Spray System (HPCS)
'4.
High Pressure Coolant Injection System (HPCI)
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S.
Low Pressure Core Spray System (LPCS) 6.
Low Pressure Coolant Injection System (LPCI) 7.
Automatic Depressurization System (ADS) 8.
Safety Relief Yalves 9.
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
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13.
Condensate Storage Tank
~
14.
Main Feedwater System 15.
Recirculation Pump / Motor Cooling Systems 2022 026 S
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Table II Infomation on Systems Caoable of Providing Post-Accident and Transient Core Cooling General System Desico Infomation e
- 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
- Nomal Position of Valves, Indication Location Direct or Indiregt indication 1 l
- Failed State of Each Valve 1
- Normal Power Sources for System Operation 1
- Nomal 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 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 Reg'd, locations, Manpower Reg'd
- Potential Corr:enalities with Control Systems
- System Interiocks & Diversion
- Operator Actions Required for System Operation & Control 2022 027
2-Water Sources
- Safety Classification & Sei,smic Classification
- Primary Water Source, Total & De'djcated Capacity, Time AvMlable
- Secondary and Backup Water Sources, Autematic/ 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)
- Alanns Located in Control Room
- System Indications located in Control Room' (pump, valves, level etc.)
- Rermte Control Panels
- Methods of Detecting Leaking Safety / Relief Valves (i.e., leakicg bellows, unseated valve)
Testing / Technical Soecifications
- Limiting Conditions for Operation
- Frequency of System & Caponent Tests 1
- System Testing Lineups 1
- System Bypass and/or Test Loops
- Method of Verification of Correct Test Lineup and 2022 028 -
Restoration to Normal Condition t
Allowable System dutage 'imes System & Componentional festing Following Mainter.ance Components Not Periodically, Tested.
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Auto Override During Tests Other Components or System Affected by Tests 1/ May be provided by a drawirig 2022 029 e
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Infomation Nceded for Containment Isolation System M
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 containnd; d.
Engineered safety feature system (yes or no);
Figure showing arrangement of containment isolation barriers; e.
f.
Isolation valve number; tecation of valve (inside or outside containment);
g.
h.
Valve type and operation;
- i. Primary mode of valve actuation; S' condary mode of valve actuation; j.
e k.
Nomal valve position; 1.
Shutdown valve position; Postaccident ' valve position; m.
Power failure valve position; n.
Containment isolation signals, including parameters sensed and their o.
set point; p.
Yalve closure time-q.
Power sourcei Valve position indication (direct or indirect) r.
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II.
Discuss the design requi.rements for the containment isolation barriers regarding:
The extent to which the quality standards and seismic desige a.
classification of the containment isolation provisions follow the recommendations of Regulatory Guides 1.26, " Quality Gre ip Classi.ffcations and Standards for Water, Steam, and Radioactive-Water-Containing Components of Nuclear Power Plants," and 1.29. " Seismic Design Classification";
Assurance of the operability of valves and valve operators in the b.
containment atmosphere under normal plant operating conditions and postulated accident conditions.
Qualificati;n of closed systems inside and outside the containment c.
as isolatien barriers; d.
Qualification of a valve as an isolation barrier; Required isolation valve closure times; e.
Mechanical and electrical redundancy to preclude comnon rode f.
failures; Primary and secondary 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 yngineersd sahety feature system or essential line) for the purpose of detemining 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 isola _ tion 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 modes and containant isolation provision and procedures for lines that transfer potentially radioactive fluids out
' of the containment.
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Additional Systems and Operational Infonnation 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 ;.;easurement is inside or outside the core shroud.
2.
Describe the instrument piping arrangements and types of transducers used.
3.
Which levels are conitored in the control room and how are they indicated (i.e., recorders, meters)?
4.
Which measurements provide signals.or 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.
nonnal 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.
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ENCLOSURE 2 REQUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS ANALYSIS GROUP 2022 034 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 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 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 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 LOCA.
Discuss any ways in which natural circulation can be degraded, such as fluid stratification in the lower plenum caused by inoperation of the cleanup system. Assess the possible effects cf non-condensible gases.
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. II. The following questions pertain to your small break LOCA analysis methods:
(3) Demonstrate that your current srall break LCCA analysis methods are appropriate for application to each of the cases identified in itms (7) through (10) below. This demonstration should include an asse:,-
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 ssessment, you nudify 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 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 is different for different classes of plants (jet pump, non-jet pump, BWR 4, SWR 5), provide an example of each kind.
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(7) Provide the results of a sample analysis of each type of small break behavior discussed in the response to item (i) (e.g., depressurization, pressurehangup,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 which does not result in HPCI initiation.
This may require more than one calcu-1 ation.
(9) Provide the results of an analysis for a single stuck open safety / relief valve, and the maximum nuber 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 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 -
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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4 (12)
Provide revised emergency procedures or guidelines for the preparation of operational ecocedures for the recovery of plants following small LOCA's. This shoulo include b%n shor term and long-term situat' ions and follow 'hrough to e, table condition.
The guidelines should include recognitior cf 'he even.
grecautions, 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 not be relied upon during these events. What would be the effect of this instrumentation on automatic protection 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 analysis of the symptoms of inadequate 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 inadeouate core cooling exists. The calculations should be carried out far enough so that all important phencmena and instrument indications are included.
Ea'ch 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 orocedures 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.
Tb assist in the review of these cases, we will require computer output information 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, P Downcomer:
L, X or T SUB
'eak:
SRV, W, X or Break,W,X_,[Wdt Nomenclative: P - Pressure L - Mixture Level X - Quality T - leg erature W - Mass Flow Rate H - Enthalpy 2022 040
.