ML19257D528
| ML19257D528 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 01/31/1980 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 LAC-6769, NUDOCS 8002040539 | |
| Download: ML19257D528 (20) | |
Text
'
s Ib.tIIt YLANIb I'01%'EIt C001*EIt.t TIVE Ea Grone, Gibconan
$4601 January 31, 1980 In reply, please refer to LAC-6769 DOCKET NO. 50-409 Director of Nuclear Reactor Regulation ATTN:
Mr. Harold Denton, Director Office of Nuclear Reactor Regulation U.
S. Nuclear Regulatory Commission Washington, D.
C.
20555
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 ADDITIONAL INFORMATION - THREE MILE ISLAND SHORT TERM LESSONS LEARNED
Reference:
(1)
NRC Letter, Denton to Linder, dated January 2, 1980.
(2)
NRC Letter, Denton to Linder, dated October 30, 1979.
(3)
DPC Letter, Linder to Denton, LAC-6706, dated December 20, 1979.
(4)
DPC Letter, Linder to Denton, LAC-6680, dated December 6, 1979.
(5)
Licensee's Answer to Order to Show Cause, dated January 22, 1980.
(6)
DPC Letter, Madgett to Giambusso, LAC-2788, dated October 9, 1974.
(7)
DPC Letter, Madgett to Director NRR, LAC-4082, dated July 21, 1976.
(8)
DPC Letter, Linder to Ziemann, LAC-6705, dated December 20, 1979.
Gentlemen:
Your letter (Reference 1) requested a description of the methods used to implement the " Category A" requirements completed by January 1, 1980, which were established in NUREG 0578.
Below are listed the " Category A" items as identified in Table B-1 of NUREG 0578 together with the method of implementation as requested by Reference (2) and the date of implementation as requested by Reference (1).
1866 271
_1_
5 39 8002040
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.1 Emergency Power Supply Requirement for the Pressurizer Heaters, Power-Operated Relief Valves and Block Valvei, and Pressurizer Level Indicators in PWRs.
Not applicable to boiling water reactors.
2.1.2 Performance Testing for BWR and PWR Relief and Safety Valves.
DPC committed to participate in the BWR Owners Group evaluation of the operability of pressure relieving main steam safety valves.
The conclusions arising from the Owners Group evaluation of BWR safety valve operation under single phase and two-phase flow conditions will be factored into the overpressurization analysis as they apply to the LACBWR facility.
This commitment was con-tained in Reference (3) which was dated December 20, 1979.
NRC staff has accepted BWR Owners Group program partici-pation in lieu of individual program description and schedule.
2.1.3.a Direct Indication of Power-Operated Relief Valve and Safety Valve Position for BWRs and PWRs.
Direct valve position indication for each of LACBWR's three (3) safety relief valves (LACBWR does not have any power-operated relief valves) will be provided.
This unambiguous safety grade position indication system will be the primary source of safety valve position status.
The existing safety valve discharge temperature indi-cation system to provide the operators with a backup method of safety valve position status is retained.
This latter system was described in Feference (4) which was submitted on December 6, 1979.
The direct valve position indication system to be in-stalled by LACBWR has been developed and tested by the valve manufacturer (Crosby Valve and Gage Company).
It consists primarily of limit switch subassemblics which mount on a modified the valve cap and directly sense the open-closed position of the relief valve stem (spindle).
Two environmentally qualified limit switches are provided in each subassembly to separately detect the closed and open positions of each safety valve.
The direct valve i866 272
-2_
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.3.a position indication system will be seismically qualified (Cont'd) consistent with the components and system to which it is attached.
Direct position indication will be provided to the plant operator in the form of indicator lights mounted on a contr71 room panel:
a coninuously "on" green indicator light will indicate that the associated relief valve is in its normal closed position; an "on" red light will alert the operator that the relief valve is not closed.
The operator can independently verify the open valve condition indicated by the "on" red light by checking the safety valve discharge temperature indi-cators/ alarms and by checking the containment building radiation monitors for changes in activity.
Power for the direct position indication system will be obtained from the reactor plant 125V DC bus.
Owing to the unavailability of the necessary equipment prior to January 31, 1980, installation of this modifica-tion will be initiated following plant shutdown on March 11, 1980.
2.1.3.b Instrumentation for Detection of Inadequate Core Cooling In PhJs and BWRs.
A detailed analysis was performed to establish the behavior of reactor water level instrumentation during various plant trans-ients (e.g.,
loss of feedwater, loss of recirculation flow, LOCA events, etc.).
The primary purpose of this analysis was to verify that the reactor water level safety channels would initiate the LACBWR Emergency Core Cooling Systems as required for any transient which might adversely affect core cooling.
This analysis was included in DPC's appli-cation for the full-term operating license.
(See Refer-ence (6) ).
Additional analysis was performed during the ECCS qualifi-cation to the Interim Acceptance Criteria to verify that the current reactor water level safety channels would also properly initiate the emergency core cooling even with LOCA locations directly affecting some of the reactor water level channels (e.g.,
the reactor water level standpipe).
(See Reference (7)).
Both analyses demonstrated that all conditions which could result in the reactor water level approaching the top of the core or could produce divergent uater level indications would also initiate ECCS operation to prevent core damage.
Consequently, it has been concluded 1866 273
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.3.b that the present water level instrumentation is adequate (Cont'd) for core protection and, when supplemented by other existing plant instrumentation, can be used by the oper-ator to recognize conditions which could result in inad-equate core cooling.
DPC has revised the Primary System Leak Procedure (revision was issued December 21, 1979) to guide the operator in recognizing conditions which could result in inadequate core cooling with the existing plant instrumen-tation.
The revised procedure places first priority on verifying the presence of a leak using plant instrumen-tation (reactor water level instruments, feedwater flow, continuous activity air monitors, humidity detecting equipment).
Following leak verification, the procedure then places priority on operator actions which will main-tain the reactor water level above the ECCS initiation water level (-12 inches).
If the water level cannot be maintained by feedwater injection and/or leak isolation, or if the operator is uncertain as to the water level status (e.g.,
divergent water level indication due to standpipe LOCA), the procedure then requires the operator to verify that the High Pressure Core Spray System is operating properly.
If the operator cannot verify that the High Pressure Core Spray System is operating properly within the specified time limit, the procedure requires the operator to place the Alternate Core Spray System into operation.
This sequence of actions combined with the use of existing plant instrumentation will provide verifi-cation to permit the operator to reccgnize and correct, in a timely manner, conditions which could result in inade-quate core cooling.
Detailed descriptions of the existing reactor water level instrumentation for operators use is presented in the LACBWR Operating Manual.
Installation of a primary coolant saturation meter is not required for boiling water reactors.
Additional instrumentation is not required to recognize conditions of inadequate core cooling.
Implementation of this item was regarded complete on December 21, 1979. 1866 274
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.4 Containment Isolation Provision for BWRs and PWRs.
Identification of essential and nonessential systems is listed below.
The basis for selection of each essential system is provided.
Fluid system lines penetrating containment are classified as essential if they meet one or more of the following criteria:
a)
The system is required for energency reactor shutdown; b)
The system is required for emergency core cooling; c)
The system is used to maintain the integrity of the containment structure, thus minimizing off-site releases of radioactive materials.
d)
The system is required to plice and maintain the reactor in a safe, stable condition.
Based on this criteria, the following systems are classified essential:
Overhead Storage Tank Main Steam Feedwater High Pressure Service Water
. Component Cooling Water Demineralized Water
. Control Air Containment Building Pressure and Water Level Shutdown Condenser Vent
. Vacuum Breakers Containment Building Spray Water Handwheel Shutdown Condenser Shell Side Vent
. Alternate Core Spray Fluid system lines penetrating containment classified as nonessential are:
. Station Air
. Decay Heat Removal Startup Line
. Condensate to Seal Injection Reservoir
. Containment Ventilation
. Retention Tank Discharge to Waste Water Storage Tanks 1866 275
-s-
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.4
. Purification Resin Sluice Line (Cont'd)
. Sump Pump Discharge to Retention Tanks Heating Steam and Condensate Return
. Retention Tank Discharge to Evaporator Feed Tank Reactor Cavity Purge Containment Vessel Off-Gas Ver.t Header to Stack Containment Building Drain Line Automatic isolation or other isolation provisions are provided as follows:
System Present Status Station Air---------Check Valve Decay Heat Removal--Automatic Valve Closure on low reacter Startup Line water level.
Reset modification re-quired.
Diverse parameter for isolation will be high containment building pressure.
Condensate to Seal--Check Valve Injection Reser-voir Containment---------Automatic valve closure on high primary Ventilation system pressure, high containment build-ing pressure; any one of three detectors -
high immediate particulate,high delayed particulate, high gaseous activity.
Reset modifications required on all of above parameters.
Diverse parameter for isolation will be low reactor water level.
Retention Tank------Automatic valve. closure on high contain-Discharge to ment building pressure.
Waste Water Storage Tank Reset modification not required.
Diverse parameter for isolation.will be low reactor water level.
1866 276
_c_
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.4 NOTE:
The discharges from the following systems (Cont'd) are piped in common with the above system, therefore, the above modification encompasses the systems listed below which are:
Sump Pump Discharge to Retention Tanks Retention Tank Discharge to Evaporator Feed Tank Purification--------Manual valves normally closed except Resin Sluice for sluicing during reactor operation, Line but containment integrity provided by closed vent valves during sluicing.
Heating Steam-------Heating Steam has check valve.
and Condensate Return Condensate Return has Automatic Vcive Closure on high containment building pressure.
Reset modification required.
Diverse parameter for isolation will be low reactor water level.
Reactor Cavity------Check valve Purge Containment Vessel--Automatic valve closure on high primary Off-Gas Vent system pressure and high containment Header to Stack building pressure.
Reset modification required.
Diverse parameter for isolation will be low reactor water level.
Containment Build---Manual valves, normally closed.
ing Drain Line Installation of the modifications described above will be initiated following plant shutdown on March 11, 1980.
1866 277
Mr. Harold Denton, Director LAC-6769 Office of Muclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems and 2.1.8.a Improved Post-Accident Sampling Capability.
Post-accident purging has been included as a design basis for licensing LACBWR (See Section 3.4 of the full term license application for LACBWR, Reference 6).
Dedicated penetrations for post-accident purging will be provided as part of the modification to comply with the requirements of Section 2.1.8.a of NUREG 0578 (Improved Post-Accident Sampling Capability).
These penetrations will provide the purging capacity identified in the design basis.
A review of existing post-accident sampling capabilities at LACBWR against the requirements of Section 2.1.8.a of NUREG 0578 has been completed.
Existing sampling provisions do not meed the requirements, since currently available sample locations would be inaccessible in the post-accident scenario postulated by the Task Force.
The facility modifications required to bring the plant into full compliance have been identified.
A new, additional remote sample station will be installed at an accessible location outside of the containment building.
Procedures are available to perform boron and chloride chemical analyses assuming a highly radioactive initial sample Full implementation of the post-accident sampling system will be complete by January 1, 1981.
With respect to improved post-accident sampling capability, liquid reactor coolant samples will be conveyed from the sampling point in centshment to the designated post-accident sampling station in the turbine building via 3/8" stainless steel tubing.
The sample line will be equipped with:
a manual isolation valve close to the source, a remote-manual containment boundary valve, a sample cooler, a remote-manual flow control pressure reducing valve, pressure breakdown orifice, sample cylinder, miscellaneous check valves and flow directing valves, and instrumentation to measure sample flowrate, pressure and temperature.-
1866 278
_ 8-
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.8.a Sample fluid will flow in a closed loop from the sample (Cont'd) point, out to the post-accident sample station, and back into containment where it will be directed to the reten-tion tanks.
Provisions will be made to introduce demin-eralized water into the sample stream to dilute it, if required for safe handling, before it reaches the sample station.
Samples will be obtained by:
a) purging to the retention tanks until fluid representative of the recirculating reactor coolant fills the line; b) adding demineralized dilution water to the sample stream to the degree required; c) after equilibrium is reached, divert-ing the aample stream to a sample cylinder, which is sub-sequently isolated, manually removed, and transported to the health physics laboratory for analysis.
With respect to dedicated penetrations for external recombiners or post-accident purge systems, post-accident hydrogen content of containment air will be monitored by continuously passing a small sample stream of the contain-ment atmosphere through a hydrogen analyzer located outside of containment.
A small compressor will provide the motive force to draw sample air from containment, route it through the analyzer, and return it to containment in a closed loop.
Continuous readout of H2 concentration will be provided in the control room.
A block valve is installed downstream of the sample compressor.
A bypass line around this valve contains a sample cylinder.
A discrete contain-ment air sample can be obtained manually by bypassing the circulating containment sample air through this line, isolating the sample cylinder, and removing it to the laboratory for analysis.
Manual sampling will be performed in a shielded, accessible area in the turbine building.
A piping connection for a post-accident containment building purge will be provided in the sampling loop downstream of the sample compressor.
This purge line will be routed to the plant vent stack.
The line will contain instrumentation and control valving (manually operated) to provide the capability to conduct a controlled, post-accident venting of the containment volume for the purpose of combustible gas control over the full range of containment building pressures from design pressure of 52 psig to sub-atmospheric.
_9_
1866 279
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.8.a Purge system piping and valve sizing will be based upon (Cont'd) the flow requirements established in the plant design.
System design is such that no single active component failure will result in the loss of venting capability.
With respect to Section 2.1. 5.a and 2.1.8.a, f ull implementation will be completed by January 1, 1991.
2.1.6.a Integrity of Systems outside containment Likely to Contain Radioactive Materials (Engineered Safety Systems and Auxiliary Systems) for PWRs and BWRs.
At LACBWR, all systems which would contain large inventories of radioactive materials following a serious accident or transient are located entirely inside containment.
As a recult, the special leak measurement and leak reduction measures described in Section 2.1.6.a of NUREG 0578 are not applicable to this plant.
DPC will continue to strive to maintain leakage from components in normally readioactive systems to as low as practical levels.
Leakage detection is normally achieved with the plant on line by monitoring local and stack activity which is indicated and recorded continuously in the control room.
Since all plant building and equipment off gas vents are routed ultimately to the stack, there are no locations where leakage could continue undetected.
At LACBWR, maintenance of systems with indicated leakage has always been accomplished on an expedited basis.
2.1.6.b Design Review of Plant Shielding of Spaces for Post-Accident Operations.
In accordance with the position outlined by NUREG 0578, and clarified by Reference (2), a radiation and shielding design review of areas which will require occupancy by an operator to aid in the mitigation of or recovery from an accident has been performed.
Three areas have been identi-fled in which access will be necessaryt The control room, the post-accident sampling station, and the health physics laboratory.
The design review has found that no new sources, except the samples obtained, would be generated in any of these areas as the result of a post-accidental release of radioactivity equivalent to that described in Regulatory Guide 1.3.
~1866 280
- to -
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.6.b I.
Control Room (Cont'd)
The dose calculations performed by Allis-Chalmers in ACNP-66564 and accepted by the NRC in granting Pro-visional Operating License NO. DPR-45 have been verified during this design review to be correct as reported in the LACBWR Final Safeguards Report, Section 10.2.
These calculations are conservatively based upon an assumed 100 percent release of fission products from the fuel.
As indicated in the LACBWR Final Safeguards Repert, the average whole body dose received an an operator during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the control room is 6.3 Rem.
As detailed by Operating Procedures, all essential systems at LACBWR can be fully controlled from the control rocm during a maximum credible accident.
Since the cumulative 24-hour doses received in the control room slightly exceed GDC 19 limits, shielding modifications and changes to post-accidental procedural controls are necessary to meet the GDC 19 limits.
II.
Post-Accident Sampling Station A post-accident sampling station, as required by Section 2.1.8.a of NUREG 0578, does not exist at present.
A location accessible from the control room has been identified for a sampling station and the latter will be designed to the requirements outlined by NUREG 0578, Section 2.1.8.a.
III.
Health Physics Laboratory The presently-existing health physics laboratory, which will be needed during post-accident operations, has been found to have shielding superior to the control room and is accessible under accident conditions.
Based upon control room doses ard minor shielding additions along the transit path from the control room to the sampling station to the laboratory, accessibility is assured with dose rates during transit significantly below GDC 19 limits.
The design review was completed on January 17, 1980.
1866 281.
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.7.a Automatic Initiation of the Auxiliary Feedwater Systems for PWRs.
Not applicable to boiling water reactors.
2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generators for PWRs.
Not applicable to boiling water reactors.
2.1.8.c Improved In-Plant Iodine Instrumentation.
The equipment for compliance with improved iodine monitoring is on order and is scheduled for delivery February 8, 1980.
The necessary equipment is identified in Reference (5).
2.1.9 Transient and Accident Analysis.
Our letter, Reference (3), discussed the need for a meeting with the NRC staff to clarify the scope of the accident and transients analysis.
Referring to Table B-2, Task and Responsibilities Timetable, the following status report is submitted.
Task Description Completion Dates 1.
Small Break LOCA Small Break LOCA analysis pre-analysis and prepar-sented in Reference (8).
NRC ation of emergency staff determined LACBWR guidelines procedure guidelines.
not necessary due to nongeneric design.
2.
Implementation of Retraining of operators completed small break LOCA by 1/1/80 for all except 4 oper-emergency procedures ators on vacation who were re-and retraining of trained on 1/4/80.
operators.
3.
Analysis of inade-The following analytical report quate core cooling is presented.
and preparation of emergency procedure guidelines.
1866 282 Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section Title No.
2.1.9 The Analysis of Inadequate Core Cooling Used to Modify (Cont'd) the Operating Procedure for Recognition by Operating Personnel of Inadequate Core Cooling.
The LACBWR systems designed to provide emergency core cooling (ECC) are the High Pressure Core Spray (HPCS) and the Alternate Core Spray (ACS) systems.
The HPCS system is the principal short-term ECC system since it is expressly designed to provide the required core cooling for the full spectrum of pipe breaks:
1.
It sprays water directly over the core an1 thus it is suitable for both above-core and below-core LOCA's.
2.
It is automatically actuated into operation by a key reactor parameter (reactor low water level of -12 inches or containment buiiding pressure of +5 psig).
3.
Its principal components are redundant.
4.
Its operation is independent of off-site power.
5.
It has sufficient head to inject water at any expected reactor system pressure.
The ACS system, on the other hand, is the principal long-term ECC system:
1.
It is directly supplied by an unlimited water source (river).
2.
The ACS pumps are self-powered with diesels.
3.
Its principal components are redundant.
4.
It is automatically actuated upon low reactor water level of -12 inches coincident with containment pressure of +5 psig.
The water reaches the vessel when reactor pressure decreases to 150 psig.
The HPCS system includes two pumps which normally take suction from the 42,000-gallon overhead storage tank and discharge to the core spray header (ring) at pressures up 13 -
1866 283
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No. _
Title 2.1.9 to 1400 psig.
The spray header supplies the lines (Cont'd) leading to a spray nozzle just above each fuel assembly.
The pumps obtain power from the essential buses, which in case of loss of off-site power, are supplied by independent emergency diesel generators.
Tests conducted at LACBWR show that it takes 8.2 seconds after the HPCS signal is given, to start the emergency diesel generator and get the spray into the core.
The core spray pumps are started simultaneously and both pumps supply 100 gpm of water to the spray header.
These pumps are used for core spray when the reactor remains pressurized, as in the case of a small leak.
When the reactor and containment building pressures are equalized, as after a major system leak or rupture, a low pressure supply line bypassing the emergency core spray pumps allcws water to flow directly from the overhead storage tank (or service water line) to the core spray header.
The core spray header above the top of the failed fuel location system tube support grid supplies the 72 spray lines.
An individual 3/8-inch spray line is provided for each fuel assembly.
The low pressure ACS system includes two diesel-driven service water pumps which take suction from the river.
When containment building pressure exceeds 5 psig, both diesels will start automatically and supply cooling water to the reactvr v coci 4-inch nozzle through either of the two motor operated valves which open automatically when containment building pressure reaches 5 psig coincident with reactor vessel water level of -12 inches.
The cooling water falls from the 4-inch nozzle down through the tube bundle of the high pressure core spray system, impinges on the perforated flow deflector plates, and then flows through the deflector plates downward through the core.
A 3-inch skirt is welded on the outer periphery of the deflector plates to provide for even distribution of the cooling water to all fuel elements.
Flow to the vessel commences when reactor vessel pressure drops to approximately 150 psig and full flow of approximately 300 gpm is reached when reactor vessel and containment vessel pressures are equalized on a LOCA.
}866 284 Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.9 A full range of breaks has been evaluated in the analysis (Cont'd) of Loss of Coolant Accidents (LOCA's) at LACBWR.
Breaks of various sizes both below and above the reactor core were investigated and results reported in References 1, 2,
and 3 of this report and in response to Request for Additional Information Regarding Small Break LOCA Analysis.
These results have shown that the HPCS is adequate, even when limited to one pump, to maintain core parameters within 10CFR50 Appendix K criteria.
The LOCA analysis, as reported, also assumed that feedwater flow would be termin-aced at the time of the break and that the shutdown condenser would not be available.
The steam flow to the turbine was allowed to continue until the pressure dropped to 1000 psia in the steam dome at which point it was shmt off.
The plant responses to mitigate the consequences of a LOCA are fully automatic and are actuated without operator action required.
The only LOCA which requires ACS operation for short-term cooling is the break of the primary ECC piping, i.e.,
the 1-1/2 inch HPCS line, which is above the core.
For this break (and all other breaks at above-core level) the upward flow of steam through the core is sufficient to keep the core cool (Reference 4).
The depressurization of the reactor to 150 psig has been calculated to take about 28 minutes for the largest HPCS pipe break.
For smaller HPCS piping breaks, sufficient time is available to the operator to evaluate the problem and to initiate manual depressurization of the reactor system.
For larger above-core breaks (single-ended main steam line breaks) the core will be continuously cooled by the upward flow of the steam and water mixture, until the automatic actuation of the ACS system when the reactor pressure is finally decreased to 150 psig, provides long-term cooling.
An evaluation of the period of time that is available for operator action was performed using results from Reference (1) in the case of below-core breaks assuming that none of the high-pressure cooling systems (feedwater, HPCS, shut-down condenser) at LACBWR are available, as stipulated by Section 2.1.9, Inadequate Core Cooling.
This analysis estimated the time after the break by which the operator must make a decision to depressurize the vessel manually so 1866 285.
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.1.9 as to allow the ACS to start cooling the core and thus, (Cont'd) prevent damage to the fuel, which has been defined for the purposes of this calculation to be the stainless 0
steel cladding temperature exceeding 2300 F at the point in the core with the highest total peaking factor The calculation was performed by using conservative assumptions as more realistic methods would have required extensive computer code analysis, which was not feasible in the short period allowed for the responses.
These assumptions included adiabatic heatup of the fuel assem-blies following water level uncovery, thus ignoring cooling by convection and by radiation to steam above the two-phase mixture and to adjacent rods and the shrouds.
Various break sizes were evaluated up to the point of automatic ACS actuation which would result from rapid depressurization with the larger breaks.
Results indi-cate that following a below-core break with no high pressure cooling water, a minimum of 8.5 minutes are available for positive operator action before any fuel damage can occur.
References for Above Report:
1.
Gulf United Report, "Techn.' cal Evaluation of the Adequacy of LACBWR Emergency Core Cooling System",
SS-942, May 31, 1972.
2.
Gulf United Report, " Response to Questions by AEC with regard to Gulf United Report SS-942", SS-1075, Rev. 1, November 15, 1973.
3.
Gulf United Report, " Supplemental Information on the LACBWR Emergency Core Cooling System", SS-ll26, October 10, 1973.
4.
NES 81A0019, " Single Failure Analysis of the LACBWR Emergency Core Cooling Systems", November 1975.
1866 286 4
5 s;e sg a
5 a1 s
8 E84 k
5^
ga1
~,
ar,,
sn
~
h!
b s
s x
~
g gi;8
~
l s
g
~
Sis
%s
~
g:'
x gg j s:
%N NN
. gi DI
\\
N\\
N l}
x st9 5!
'N
\\
\\ Nl eI i
~
[i I
y
\\'
8P T
l
~
~
~
l i
i s
m 1866 287
,0 B
8 a
1 7
_1. _. _
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section
_ _N o.
Title 2.1.9 Task Description Completion Dates (Cont'd) 4.
Implementation of Retraining of operators completed emergency procedures by 1/1/80 for all except 4 oper-and retraining relat-ators on vacation who were re-ed to inadequate ccre trained on 1/4/80.
cooling.
5.
Analysis of accidents The scheduled completion of this and transients and analysis is June 1, 1980.
preparation of emer-gency procedure guidelines.
6.
Implementation of The implementation of emergency emergency procedures procedures and retraining is and retraining re-scheduled for September 1, 1980, lated to accidents three months after the analysis and transients.
completion.
7.
Analysis of LOFT Not applicable to LACBWR.
small break tests.
2.2.1.a Shift Supervisor Responsibilities.
The duties and responsibilities of the Shift Supervisor were delineated in accordance with the guidance of NUREG 0578.
The management directive was issued December 31, 1979.
2.2.1.b Shift Technical Advisor.
The position of Shift Technical Advisor, Reference (4),
was activated at 2400 on December 31, 1979, on a continuous on-site basis.
2.2.1.c Shift and Relief Turnover Procedures.
The Shift Turnover Check List was initiated to comply with the requirements of NUREG 0578.
The Quality Assurance Department will perform audits to evaluate the effectiveness of the shift and relief turnover procedures.
This imple-mentation was completed December 28, 1979.
}hhh 2hh 2.2.2.a Control Room Access.
The directive establishing a clear line of authority ror limiting control room access was issued December 31, 1979.
Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 Section No.
Title 2.2.2.b On-Site Technical Support Center.
The on-site Technical Support Center was identified as the new Administration Building (See Reference 4).
The Emergency Plan procedures and the Administrative Control Procedure to utilize the Technical Support Center were approved January 11, 1980.
The Technical Support Center (TSC) of the La Crosse Boiling Water Reactor is established in Room 115 of the Administration Building.
The TSC has access to the latest revisions of all plant drawings, procedures and safety analyses, as well as communication with the control room.
An Administrative Control Procedure descriptive of the Technical Support Center together with procedures foi its activities and staff have been Unplemented.
The upgrading of the Technical Support Center to meet the long range requirements will be accomplished during 1980.
This will include instrumentation readouts and capability for data transmission, installation of permanent monitors, modification of the ventilation system to permit use of particulate and charcoal filters as well as the ability to operate on emergency power supply, and dedication of breathing apparatus and potassium iodide to the TSC.
2.2.2.c On-Site Operational Support Center.
The site evacuation points of LACBWR completely comply with the NUREG 0578 position on On-Site Operational Support Centers.
The evacuation point is:
On site; Separate from the Control Room; Is the place where operating support personnel report in an emergency; Has communication with the Control Room; and Is identified in the Emergency Plan including methods and lines of communication and management.
This implementation has existed for several years prior to January 1, 1980.
1866 289 Mr. Harold Denton, Director LAC-6769 Office of Nuclear Reactor Regulation January 31, 1980 If there are any questions regarding this submittal, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE Frank Linder, General Manager FL:JDP:af cc:
J. Keppler, Reg. Dir., NRC-DRO III 1866 290
_20_
.