ML19248D621
| ML19248D621 | |
| Person / Time | |
|---|---|
| Site: | Zimmer |
| Issue date: | 07/23/1979 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Borgmann E CINCINNATI GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7908170102 | |
| Download: ML19248D621 (19) | |
Text
.
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i Docket File LRubenstein NRC PDR ISE (3)
Local POR SSheppard JUL 2 31979 LWR #1 File PKreutzer NRR Reading VRooney
- ecket io: LG-350 DEisenhut RBevan WPGamill DClark TJCarter DVerrelli i
It. Earl A.' Ecrpann DBrinkman PPolk WFKane JHannon Vice Fresident - Engineering Cincinnati Cas & Electric Cccipany JHeltemes TXevern P. C. eux 960 DFRoss Attorney, ELD f7 f
a Cir.cinnati, Chio 4o201 Lear.ir borgaann:
RLBaer bcc:
J. R. Buchanan ODParr T. B. Abernathy StiaECT: ACUITICf4AL IfiF0R,ATI0ff REGUIRED FOR LRC STAFF UENCbCRT CR LUI.Lli:G WATER REACTCRS On June 26, 1979, the IRC staff met with representatives from each of the licensees of coiling water reactors (8LR's) as v. ell as the applicants for i
- r. ear-tem operating licenses for E'.iR 's.
At that meeting v;e discussed cur short-term program for reviewinc the implications of the Three File Island Unit 2 accicent on operating EWR 's and near-tern (peratinc License applica-i tions for DR's.
At the r.:eeting he held a discussion of cur general informa-tion needs and noted that our review will concentrate on tuo basic areas, i.e., systems and analysis. -he stated that ve would provide.ycu with cur formal requests for information at a later cate.
,, which consists of three attachrents, contains our recuests for accitional information in the systems area. Enclosure 2 contains our requests for acditional infomation in the analysis area.
In order fcr us to r..aintain cur schecule we request that you provide clear and cornplete respcnses to the enclosed requests by August 27, 1979.
If you cannot reet this schedule or if you require any clarification.cf these r.atters please ccntact i.illiem F. Kane, (201) 492-7745 iar.;edi atel y.
Sir.cerely, Original signed by D
m:ny,stal:
John F. Stolz, Chief D
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_a liivisicn of Froject ~anasement
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Zee next p ce 07/, y /79 07/ 1 3 /79 OATE b NEC PORM 318 (9 76) NRCM 0240 W u.a. e ew s m aas s a' *a'at'a* ** *'c u ieve. see - vee 7908170/o2 g
Mr. Earl A. Borgmann JUL 2 31979 Vice President - Engineering The Cincinnati Gas and Electric Company P. O. Box 960 2
Cincinnati, Ohio 45201 3r cc: Troy B. Conner, Jr., Esq.
David B. Fankhauser, PhD
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Conner, Moore & Corber 3569 Nine Mile Road 1747 Pennsylvania Avenue, N. W.
Cincinnati, Ohio 45230 Washington, D. C.
20006 Thomas A. Luebbers, Esq.
Mr. William J. Moran Cincinnati City Solicitor General Counsel Room 214, City Hall
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The Cincinnati Gas and Electric Cincinnati, Ohio 45202 Company 4.
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P. O. Box 960 Mr. Stephen Schumacher Cincinnati, Ohio 45201 Miami Valley Power Project P. O. Box 252 fir. William G. Porter, Jr.
Dayton, Ohio 45401 Porter, Stanley, Arthur and Platt Ms. Augusta Prince, Chairperson 37 West Broad Street 601 Stanley Avenue Columbus., Ohio 43215 Cincinnati, Ohio 45226 C
Mr. Steven G. Smith, Manaaer Charles Bechhoefer, Esq., Chaiman F-Engineering & Project Control Atomic Safety & Licensing Boa c The Dayton Power and Light Panel Ccmpany U. S. Nucle 3r Regulatory Commiss,.-
P. 0. Box 1247 Washington, D. C.
20555 Dayton, Ohio 45401 J. Robert Newlin, Counsel S
The Dayton Power and Light Company r
>. 0. Box 1247 c.
Dayton, Ohio 45401 h-7 Mr. caes D. Flynn h'
Manager, Licensing Environmental Affairs 7!
The Cir.cinnati Gas and Electric Company L
P. O. Box 960 Cincinnati, Ohio 45201 Mr. J. P. Fenstermaker Senior Vice President-Operations Columbus and Southern Ohio Electric Company 215 North Front Street Columbus, Ohio 43215 "g
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ENCLOSURE 1 REQUESTS FOR ADDITIONAL INFORMATION BULLETINS & ORDERS' SYSTEMS GROUP
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Information on Systems Capable of Providind Post-Accident and Transient
_ Core Cooling
_ Ins tructions
's Table T 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 ccoling, that have not been specifically identified, they should La 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 prcvided where needed.
If questions arise pertaining to the interpretation of the type of information requested cor. tact Byron Siegel (301-492-7341) or Wayne Hodges (301-492-7588).
fl0TE:
We are aware that r.uch of the infonnation 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 Isolation Coolino System (RCIC) 2.
Isolation Condenser 3.
High Pressure Core Spray System (HPCS) 4.
High Pressure Ccolant Injection System (HPCI)
S.
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
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Table II Infomation on Systems Capable of Providino Post-Accident and Transient Core Cooling General System Desion Information,
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 1
or Indirei;t Indication l
- Failed State of Each Valve 1
Nomal 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 Sources
- Number of Safety & Relief Valves & Relieving Capacity
- Relief & Safety Valve Setpoints
- System Trips
- riethods of Cooling System,C. nponents (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'd, Locations, Manpower Reg'd
- Potential Comunalities with Control Systems
- System Interlocks & Divsrsion
- Operator Actions Required for System Operation L Control g,
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Water Sources
- Safety Classification & Sei,smic Classification
- Primary Water Source, Total & Dedjcated Capacity. Time Available
- Secondary and Backup Water Sources, Automatic / Manual, Procedure, Time, Reg'd
- Strainers in spem 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-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)
- Alams Located in Controi Room
- System Indications Located in Control Room' (pump, valves, level etc.)
- Renute 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 & Co:nponent Tests UV 1
- System Testing Li.7ps 1
- System Bypass and/or Test Loops
- Method of Verification of Correct Test Lineup and Restoration to Nortral Condition
. - Allowable System dutage Times
- System & Componentional Testing Following Maintenance
- Components ?bt Periodically, Tested.
- Auto Override During Tests
- Other Components or System Affected by Tests J/ May.be provided by a drawing
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; 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.
Nomal valve position; 1.
Shutdown valve pcsition; m.
Postaccident valve position; n.
Power failure valve position; Containment isolation signals, including pararreters sensed and their o.
set point; p.
Yalve closure time; q.
Power sourcei Valve position indication (direct or indirect) r.
d:-
8
II.
Discuss the design requi,rements_for the containment isolation barriers
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regarding:
-l a.
The extent to which the quality standards and seismic design 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 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 nortaal 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; e.
Required isolation valve closure times; f.
Mechanical and electrical redundancy to preclude cocrnon roode failures; J'
g.
Primary and secondary modes of valve actuation t
' III.
Discuss the provisions for detecting leakage from a remote manually controlled system (such as 'an yngineer d sa'fety feature system or essential line) for the purpose of determining when 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 operab'ility of the isola _ti_on valves.
Identify the codes, standards, and guides applied in the design of the V.
containment isolation system and system components.
VI.
Discuss the normal operating modes and containment isolation provision and procedures for lines that transfer potentially radioactive fluids out
' of the containment.
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Additional Systems and Operational Infomation Recuired I.
Provit 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 ronitored 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.
Wha.t 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.
nuderate 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 Er
REQUEST FOR ADDITIONAL INFC9MATION 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 o.-
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-7-
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, 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) 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 of non-condensible gases.
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II. The following questions pertain to your small break LOCA analysis methods:
(3) Demenstrate 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 assessmen'., you modify your analysis methods (e.g., system r.oding), 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 T'ble 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, BWR 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, 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 regult in HPCI initiation. This may require more than one,calcu-l ation.
(9) Provide the results of an analysis for a single stuck open sa'ety/ 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 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 situations and follow through to a stable ccndition.
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 tnet might not be relied upon during these events. What would be the effect of this instrumentation on automatic protection actions?
IV. In addition 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 actier.s to restore core cooling. These analyses should include cases assuming the recirculation pumps are both o" rating 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 procedu es or guidelines for the preparation of emergency procedures for plant recovery from inadequate core cooling.
-S-(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.
7 (16) The NRC is planning to perform audit calculations of the BWR small break LOCA. The necessary computer program input information and comparative calculations shou'd 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 1\\ s
TABLE 1 Plotted Output Parameters Core:
L, XAVG., ". Tclad Reactor Vessel:
Lower Plenum:
L, X - or T3gg, P Downcomer:
L,.'X or T SUB Leak:
SRV, W, X of Break, W, X,[Wdt Homenclative: P - Pressure L - Mixture Level X - Quality T - Temperature W - Mass Flow Rate H - Enthalpy
.t,:
),\\ L
...