ML19242A452
| ML19242A452 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/17/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-06-04, TASK-RR NUDOCS 7908020034 | |
| Download: ML19242A452 (19) | |
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UNITED STATES
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WASHINGTON. O. C. 20555 o.
'%' 'i '+.... f' July 17,1979 Docket No. 50-219 Mr. I. R. Finfrock, Jr.
Vice President - Generation Jersey Central Power & Light Company Madison Avenue at Punch Bowl Road Morristcwn, New Jersey 07960
Dear Mr. Finfrock:
SUBJECT:
ADDITIONAL INFORMATION REQUIRED FCR NRC STAFF GENERIC REPORT ON BOIL:NG WATER REACTORS On June 28, 1979 the NRC staff met with representatives frmi each of the licensees of boiling water reactors (BWRs) as we'l as the applicants for near-tem operatino licenses for BWRs.
At that meeting we discussed our short-tem prc'
- for review;ng the implications of the Three Mile Island Unit 2 accider.. 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. 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.
Si ncerely, M
be Dennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosures:
1.
Recuest for Additional Infomation (Systems Area) 2.
Request for Additional Infomation
( Analysis Area) cc w/ enclosures:
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(')() 4 See next page l /
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Mr. I. R. Finfrock, Jr. July 17,1979 cc w/ enclosures:
G. F. Trowbridge, Esquire Shaw, Pittman, Fotts and Trowbridge 1800 M Street, N. W.
Washington, D. C.
20036 GPU Service Corporation ATTN:
Mr. E. G. Wallace Licensing Manager 260 Cherry Hill Road Parsippany, New Jersey 07054 t
Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, N. W.
Washington, D. C.
20005 Steven P. Russo, Esquire 248 Washington Street P. O. Box 1060 Toms River, New Jersey 08753 Joseph W. Ferraro, Jr., Esquire Deputy Attorney General State of New Jersey Department of Law and Public Safety 1100 Raymond Boulevard Newark, New Jersey 07012 Ocean County Library Brick Township Branch 401 Chambers Bridge Road Brick Town, New Jersey 08723 4 E3 k U U ')
ENCLOSURE 1 REOUESTS FOR ADDITIONAL INFORMATION SULLETINS & ORDERS SYSTEMS GROUP 4 !j '
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Infomation on Systems Capable of Providind Post-Accident and Transient Core Coolino
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Instructions Table I is intended to be _ a all inclusive list of the systems that are capable of providing post-accident and transier.t core cooling for all types of BWRs. However, if your plant has additional or alternate systems that provide core coo]ing, that have not been soecifically 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 (i:articularjy 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).
t:0TE: We are aware that much of the infomation we are requestir.g may have already been submitted on your docket.
However, in order to expedite our review, we are requesting that you ccmpile and resubmit the information in this attachment.
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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)
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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
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12.
Control Rod Drive System
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13.
Condensate Storage Ta'nk
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15 Main Feedwater System 15.
Recirculation Pump / Motor Cooling Systems e
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Table II infomation on Systems Capable of 'roviding Post-Accident and Transient Core Cooling General System Design Infor.ation d
- Safety Classification & Seismic Category
- Plant Steam By-Pass Capacity
- Potential of Systems & Component Flooding (i.3., injection of water from CST in excess of Technical Specification min.) and Separation of Trains
- Nomal Position of Valves, Indication Location Direct 1
or Indirect Indication l
- Failed State of Each Valve 1
- Nomal Power Source: tur 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 Sou.
- 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, Manpcwer Reg'd
- Potential Cormonalities with Control Systems
- System Interlocks & Diversion
- Operator Actions Required for System Operation & Control 484 009
2 Water Sources
- Safety Classification & Seismic Classification
- Primary Water Source, Total & Dedjcated Capacity, Time Available
- Secondary and Backup Water Sources, Automatic / Manual, Procedure, Tirae, 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 Reg'd
- Loss of On-si.te AC Power - 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/No)
- Alarr.s 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 Soecifications
- Limiting Conditions for Operation
- Frequency of System & Component Tests 1
- System Testing Lineups 1
- System Bypass and/or Test Loops
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- Methed of Yerification of Correct Test Lineup and Restoration to Nor al Condition
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. -AllowableSystembuthgeTimes
- System & Componentional Testing Following Maintenance
- Components ?bt Feriodically, Tested
- Auto Override During Tests
- Other Components or System Affected by Tests 1/ May be provided by a drawing
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Information N eded for Containment Isolation System I.
For each fluid line and fluid instrument lines penetrating the containment, provice a table of design infonr.ation 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 snowing arrangement of containment isolation barriers; e.
f.
Isolation valve number;
. Location of valve (inside or outside containmer.t);
g.
h.
Valve type and operation;
- i. Primary mode of valve actuation;
- j. Secondary mode of valve actuation; k.
Nornal valve position; 1.
Shttdown valve position;
'ostaccident valve position; m.
Pcwer failure valve position; n.
Cantainment isolation signals, including parameters sensed and their o.
set point; p.
/alve closure time; q.
Power sourcei Valve position indication (direct or indirect) r.
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_2-II.
Discuss the desig" requirements for the containment isolation barriers recarding:
The extent to which the quality standards and seismic design a.
classification of the containment isolation provisions follow the
.-ecomer.dations 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.
Ass -ance 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.
Mechanical and electrical redundancy to preclude cccrnon mode f.
failures; Primary and secondary modes of valve actuation g.
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3-III.
Discuss the provisions for detecting leakage from a remote manually controlled system (such'as 'an rengineerid sa'fety feature system or essential line) for the purpose of determining wher to isolate the a'fected 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, standards, and guides applied in the design of the containrent isolation 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 containment.
'ttacNnent 3 Additional Systems and Operational Information Reauired I.
Provide copies of the procedures for loss of feedwater and small break LOCA.
II. Discuss the reactor water level measuremen. System.
In particular:
1.
Provide a diagram showing location of pressure taps used in aeasuring level. The diagram should be detailed enough to show whether the measurement is inside or outside the core sh roud.
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 measurerent 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 trare ~ ent as for a small break LOCA or stuck open SRV.
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ENCLOSURE 2 REQUESTS FOR ADDITIONAL INFORMATION SULLETINS & ORDERS ANALYSIS GROUP 4 8 /:
Ili 6
REOUEST 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 locaticn of the break, mode of operation of the re'.irculation pumps, number of ECCS systems functio;ang, 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 res,onse 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 situ.
ons 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 intemediate 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)
Provice a qualitative description of the various natural circulation modes of expected system behavior following a mall break LOCA.
Discuss any ways in which natu al circulation can be degraded, such as fluid stratification in the lower plenum caused by inoparation 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) Demonstrc e that your current smal'. 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 af 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 rodifi cation.
(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 levei 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 wi'h 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 systam response would be affected by control systems.
(iv)
If the scenario i.s different for different classes of plants (jet pump, non-jet pump, BWF 4, BWP. 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 (1) (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-l a ti on.
(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 L~CA 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 1.ft 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 occur, would be experienced k Ij i.
{II
-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 not ue relied upon du:ing these events. What would be the effect of this instrumentation on autoratic protection actions?
IV. In addit' ion to the short tem requirement identified above, it is requested that the following infomation 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 ther 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 perfonn audit calculations of the BWR small break LOCA. The necessary computer program input infomation 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 spec;fied in Table 1.
0<il I82 t
6 TABLE 1 Plotted Onaut Paramters Core:
L_, X;yg,, W, Tclad Reactor Vessel:
Lower Plenum:
L, X - or TSUB' Downcomer:
L, X or T SUB Leek:
SRV, W, X or Break,W,X_,[Wdt homenclative: P - Pressure L - Mixture Level X - Quality T - Temperature W - Mass Flow Rate H - Enthalpy 484 022