ML19249A216
| ML19249A216 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/13/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Andoginini G BOSTON EDISON CO. |
| References | |
| NUDOCS 7908210358 | |
| Download: ML19249A216 (20) | |
Text
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f, UNITED STATES y y)e(j h NUCLEAR REGULATORY COMMISSION 3g
/,. E WASHINGTON, D. C. 20555
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.gv JU:.Y 1 3 1373 Docket No. 50-293 Mr. G. Carl Andognini Boston Edison Ccmpany M/C NUCLEAR 800 Boylsten Street Boston, Massachusetts 02199
Dear Mr. Andegnini:
SUBJECT:
ADDITIONAL INFORMATION REQUIRED FCR NRC STAFF GENERIC REPORT ON BOILING WATER REACTORS On June 23, 1979 the NRC staff met with representatives frot 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-term program for reviewing the implications of the Three Mile Island Unit 2 accident on operating SWRs and near-tem Operating License applica-tions for SWRs.
At the meeting we held a discussion of our general infonnation needs and noted that cur review will concentrate on two basic areas, i.e., systems and analysts. We stated that we would provide you with our femal requests for information at a later date. which consists of three attachments contains our requ6sts for additional infomation in the systems area. 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 canplete responses to the enclosed requests by August 17, 1979.
If you cannot meet this schedule or if you require any clarification of thes e matters please contact William F. Kane, (301) 492-7745 immediately.
Si ncerely,
-7 W-Themas Ippolito, Chief Opera *ing Reactors Branch #3 Divi.an of Operating Reactors Encicsures:
1.
Reauest for Additional Infomation (Systems Area)
~
2.
Recuest for Additional Information D-( Analysis Area)
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cc w/enciesures:
See next page O
7908210-l
Mr. G. Carl Andegnini CC Mr. Paul J. McGui re Pilgrim Station Acting Manager 30ston Edison Company RF0
- l, Rocky Hill Road Plymouth, Massachusetts 02360 Anthony Z. Roisman Natural Resources Cefense Council 917 15th Street, N. W.
Washingten, D. C.
20005 Henry Herrmann. Esqui re Massachusetts Wildlife Federation 151 Tremont Street Boston, Massachusetts 02111 Plymouth Public Library North Street Plymouth, Massacnusetts 02360 e
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Cccket Nos. 50-249 Cocket No. 50-263 50-254 and 50-265 Mr. L. O. Mayer, anager Nuclear Sucport Services Mr. C. Reed Northern States Power Company Assistant Vice President 414 Nicollet Mall - 8tr. Flcor Commonwealth Edison Company Minneapolis, Minnesota 55401 P. O. Box 767 Chicago, Illinois 60690 Cocket No. 50-220 Docket Nos. 50-325 Nr. Donald P. Dise and 50-324 Vice President - Engineering Niagara Mohawk Power Corporation Mr. J. A. Jones 300 Erie Coulevard West Executive Vice President Syracuse, New York 13202 Carolina Power and Light Company 326 Fayetteville Street Docket No. 50-298 Raleigh, North Carolina 27602 Mr. J. M. P11 ant, Director Occket No. 50-271 Licensing a Quality Assurance Nebraska Public Pcwer District Mr. Robert H. Groce P. C. Gox 499 Licensing Engineer Columbus, Nebraska 68601 Yankee Atomic Electric Company 20 Turnpike Road Occket No. 50-331 Westboro, Massachusetts 01581 Mr. Quane Arnold Docket Nos. 50-259 President 50-250 Iowa Electric Light & Power Company and 50-296 P. O. Sox 351 Cedar Rapids, Iowa 52406 Mr. Hugh G. Paris Manager of Power Docket Nos. 50-321 Tennessee Valley Authority and 50-366 500 A Chestnut Street Tower II Chattanooga, Tennessee 37401 Mr. Charles F. Whitm.er Vice President - Engineering Docket No. 50-333 Georgia Pcwer Company P. O. Box 4545 Mr. George T. Berry Atlanta, Georgia 30302 General Manager and Chief Engineer Pcwer Authority of the State of New York 10 Columbus Circle New York, New York 1C019 Occket Nos. 50-277 941 254
- and 50-278 unt.54 atd G. Sauer,.Jr., Escuirel l
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ENCLOSURE 1 2EQUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS SYSTEMS GROUP 255 p1
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Information on Systems Capable of Broviding 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 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 pro',ided 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 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 inforration 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.
Safety Relief Yalves 9.
Residual Heat Removal System i.RHR) including Shutdown Cooling, Steam Condcnsing, 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 841 2"37
Table II Infomation on Systems Caoable of Providing Post-Accident and Transient Core Cooling General System Design Infomation J
- 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 Yalves, Indication Location Direct 1
or Indireqt Indication l
- Failed State of Each Valve 1
- Normal Power Sources for System Operation I
- 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 Sources
- Number of Safety & Relief Valves & Relieving Capacity
- Relief & Safety Valve Setpoints
- System Trips
- Methods of Cooling System, Components (i.e., pumps, valves)
Syrtem 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 Conronalities with Control Systas
- System Interlocks & Diversion
- Operator Actions Required for System Operation & Control p\\
. ~. _ ~
2 Water Sources
- Safety Classification & Seismic Classification
- Primary Water Source, Total & Dedjcated Capacity, Time Available
- Secondary and Backup Water Sources, Automatic / Manual, Procedura, Time, Req'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-site 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.)
- Remote Control Panels
- Methods of Detecting Leaking Safety / Relief Valves (i.e., leaking bellows, unseated valve)
Testing / Technical Specifications
- Limiting Conditions for Operation
- Frequency of System & Coinponent Tests 1
- System Testing Lineups 1
~
- System Bypass and/or Test Loops
- Method of Verification of Correct Test Lineup and Restoration to Normal Condition
-.. -AllowableSystembutageTimes
- 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 i
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t i
Infomation 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 infonnation:
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.
f.
Isolation valve number; g.
Location of valve (inside or outside containment);
h.
Valve type and operation;
- i. Primary mode of valve actuation;
- j. Secondary mode of valve actuation; k.
Nomal valve position; 1.
Shutdown valve position; m.
Postaccident valve position; n.
Power failure valve position; Containment isolation signals, including parameters sensed and their o.
set point; D.
Yalve closure time; q.
Power sourcei r.
Valve position indication (direct or indirect) h B U-
b.
. II.
Discuss the design requirements for the containment isolation barriers regarding:
The extent to which the quality standards and seismic design a.
classification of the containment isolation provisions follow the recomendations of Regulatory Guides 1.26. " Quality Group Classi.ff catici s and Standards for Water, Steam, and Radioactive-Water-Containinn Components of Nuclear Power Plants," and 1.29, " Seismic Design Classification";
b.
Assurance of the operability of valves and valve operators in the containment atnosphere under normal plant operating conditions and postulated accident conditions, Qualification of closed systems inside and outside the containment c.
as isolation b'arriers; d.
Qualification of a valve as an isolation barrier; Required isolation valve closure times; e.
f.
Mechanical and electrical redundancy to preclude connon mode failures; g.
Primary and secondary modes of valve actuation 7'61
,- III. Discuss the provisions for detecting leakage from a remote manually controlled system (such'as 'an vmgineered safety feature system or essential line) for the purpose of detemining wher to isolate the affected system or system train. Specify the part. meters sensed, their set ;11nt, and procedure for initiation of ccntainment isolation.
IV.
Discuss the design provisions for testing the operability of the isolation 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 containment isolation provision and procedures for 11nes that transfer potentially radioactive fluids out of the containment.
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A_ttachment 3 Additional Systems and Operational Infomation Reauired I.
Provide copies of the procedures for loss of feedwater and small break
- OCA.
L II. Discuss the reactor vater 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 af the level meaeurement 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.
nomal 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 levtl 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
@A\\
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 % 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 ror
(:) a range of postulated small break LOCA's, including tN zero break ca.se, 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) aaintain the reactor coolant system at some intennediate pressure and (c) repressurize the primary systan to the safety / relief valve setpoint pressure. Various break locatnns 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 whicn natural circulation can be degraded, such as fluid stratification in the lower plenum caused by inoperation of the cleanup system. Assess the possible effects of non-condensible gases.
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841 266
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 i: ems (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 encicsure.
(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 systum response would be affected by control systems.
(iv)
If the scenario is different for different classes o' plants (jet pump, non-jet pump, BWR 4, BWR 5), provide an example of i
}Q each kind.
i (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).
(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-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 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 transients. Provide justification for your assumptions, including the time at which two-phase discharge,if it is calculated to cccur, would be experienced.
841 268
, (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 not be relied upon during these events. What would be the effect of this instrumenuation on automatic protection actions?
IV. In addition to the short tenn 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 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 nNcedures or guidelines for the preparation of emergency procedures for riant recovery from inadequate core cooling.
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. l15) Frovide evised emergency crocedures cc ;uidelines for the = dating of emergency procedures for accidents and transients considered in Section 15 of plant SAR's.
(16) *he TIRC is planning to perfom audit calculations of the BWR small break LOCA. The necessary computer program input informa' tion and comparative calculations should be provided to fricilitate this study.
To assist in the review of these cases, we will require comouter output infomation in excess of that specified in Table 1.
n *I Q I-gA\\
se TABLE 1 Plotted Output Parameters Core:
L, XAVG., 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_,jWdt Nomenclative: P - Pressure L - Mixture Level X - Quality T - Temperature W - Mass Flow Rate H - Enthalpy p\\