ML19345B665
| ML19345B665 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 11/25/1980 |
| From: | Finnan C, Milos R, Yoli A NUCLEAR ENERGY SERVICES, INC. |
| To: | |
| Shared Package | |
| ML19345B662 | List: |
| References | |
| TASK-02-04, TASK-03-06, TASK-2-4, TASK-3-6, TASK-RR 81A0039, 81A39, NUDOCS 8012020318 | |
| Download: ML19345B665 (26) | |
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DOCUMENT NO, 81A0039 REV.
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.-.ss NUCLEAR ENERGY SERVICES, INC.
PAGE.
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DESIGN OF AN EMERGENCY SERVICE WAT_R SUPPLY SYSTEM FOR LACBWR p#'p.
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'Hed 11/25/80 5101 C. F. Finnan t_
APPROVALS Ti TLE/ DEPT.
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// "2fdO Project Manager f,
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Engineerinz A
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FORM e NES 204 ? 37 80120 2 0 3IE
DOCUMENT NO. 81A0039 NUCLEAR ENERGY SERVICES, INC.
PAGE OF REVISION LOG P^
DESCRIPTdON APPROVAL DATE no N,
1 I
i FORM
- NES 206 9/78
81A0039 DOCUMENT NO.
PAGE OF m
NUCLEAR ENERGY SERVICES. INC.
TABLE OF CONTENTS Page 1.
INTRODUCTION 4
2.
SUMMARY
DESCRIPTION OF AVAILABLE DECAY HEAT REMOVAL AND FIRE PROTECTION SYSTEMS AT LACBWR FOR BOTH NORMAL ~
AND ACCIDENT COND;TIONS 5
2.1 Ncn-LOC A 5
2.2 Systems Nermally Available for Cere Cooling Following Less of Ccolant Accidents (LOCA) 7 2.3 Fire Protctica Water System Description 9
3.
DISCUSSION OF EFFECT S OF LOSS OF CRIB HOUSE ON RESIDUAL HEAT REMOVAL AND FIRE PROTECTION CAPABILITIES AT LACBWR 11 4.
EMERGENCY SERVICE W ATER SUPPLY SYSTEM ASSUMPTIONS, DESIGN REQUIREMENT, AND CONCEPTUAL DESIGN 12 4.1 Basic Assumptiens 12 4.2 General Design Criteria 12 4.3 Conceptual Design 13 4.3.1 Implementation of General Design Requirements 13 4.3.2 Rimp Capacity and Head Requirements and System Imp'ementation Time Requirements 14 5.
GENERAL DESCRIPTION OF DPC-PROPOSED ESWSS IM PLEMENTATION 6.
REFERENCES 20 LIST OF F' CURES AND TABLES Figure 1 High Pressure Service Water System 21 Figure 2 Proposed ESTSS Arrangement 22 Table 1 ESWSS Required Flowrates and Implementation Times f or Various Accident Scenarios 23 s
FC A*.* :NES 205180
A0 m DOCUMENT NO.
11 NUCLEAR ENERGY SERVICES. INC.
- 1. INTRODUCTION Preliminary results of the NRC's evaluation of liquef action potential at LACBWR site indicate that the reactor building, turbine building and, in all probability, the "B" emergency diesel generator building are safe against liquef action (Reference 1).
However, the NRC has determined that the u..b house and the underground service water piping conected thereto are ny saf e f rom liquef action. As a remedial measure, DPC has proposed to provide a dedicated saf e shutdown system which would eliminate the need to rely on the crib house and buried piping in the event of the SSE. This system would irsure the availability of an adequate supply of cooling and service water to critical systems and components and provide a continuous heat sink f or reactor decay heat for shutdown conditions ranging f rom " normal" (i.e., no damage to the reactor and its auxiliaries with cold shutdown achieved by as near to normal methods as possible) up to the worst case LOCA scenario.
The purpose of this report is to provide sufficient information relating to the requirements for and design of the auxiliary service water system to justify its aaoption as the answer to the remaining liquef action concerns.
FCRM:NES 2C5 2 80
DOCUMENT NO.
81A0039 PAGE 5
op 24 NUCt. EAR ENERGY SERVICES, INC.
j
- 2.
SUMMARY
DESCRIPTION OF AVAILABLE DECAY HEAT REMOVAL AND FIRE PROTECTION SYSTEMS AT LACBWR FOR BOTH. NORMAL AND ACCIDl}NT CONDITIONS 2.1 NON-LOCA Following a normal reactor shutdown, reactor decay heat is initially dissipated via steam bypass to the main condenser.
Cooling water is supplied to the condensor by a conventional, once-through condenser circulating-water system in which river water is supplied by two vertical, centifugal pumps located in the crib house intake structure. Cool-down by steam dump continues until primary system temperature has been rerbced to 470 F.
From this point, further cooldown to cold shutdown conditions is accomplished by means of the decay heat removal system.
The decay heat removal system contsists basically of a closed loop, a pump, and a heat exchanger. The DH pump continuously circulates a portion of reactor water through the decay heat cooler where it transf ers reactor decay and sensible heat to the component cooling water system. The component cooling water system is an intemediate closed cooling water loop, providing a heat transfer link f rom nuclear service coolers to the ultimate Seat sink (the Mississippi River) while at the same time serving as an additional physical barrier against the release of radioactive materials to the environment. Low Pressure Service Water (river water), originating f rorr. the low pressure service water punps in the crib house, circulates through the tube side of the two component cooling water coolers, absorbs the decay heat load and discharges to the river, thus completing the heat rejection path to the environment.
In the event that reactor shutdown occurs and the nain condenser is not available as a heat sink (as would occur, f or example, if the reactor building or turbine building main steam isolation valves were closed or if loss of off-site power terminated circulating water flow to the condenser), the overhead shutdown condenser in the reactor building is used initially f or deacy heat removal and for cooldown to decay heat cooling system initiation temperature of 470 F.
FCP.i
- NES 205 220
81A0039 DOCUMENT NO.
PAGE OF NUCLEAR ENERGY SERVICES. INC.
The ACS system, on the other hand,is the p-ul 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 the 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 hea4r (ring) at pressures up to 1400 psig. The spray header supplies the lines leading to a spray nozzle just above each fuel assembly. The cumps obtain power f rom the essential Nusses, which, in cases of loss of off-site power, are supplied by independer t emergency diesel generators.
Tests concucted at LACBWR show that it takes 3.2 seconds af ter the HPCS signal is given to star 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 af ter a major system leak or rupture, a low pressure supply line bypassing the emergency core spray pumps allows water to flow directly f rom the overhead storage tank (or service water li te) to the core spray header.
The core spray header above the top of the f ailed fuel location system tube support grid supplies the 72 spray lines. An individual 3/3 inch spray line is provided f or each fuel assembly.
FCP.1 :NES 205 2,80
81A0039 DOCUMENT NO.
11 7
24 PAGE OF NUCLEAR ENERGY SERVICES. INC.
The shutdown condenser is a horizontal U-tube heat exchanger. Reactor steam is condensed in the tubes and the cooling water is evaporated in the shell. The shell side is normally supplied with demineralized water through a 3 inch line.
This line contains a control valve automatically positioned by a level controller to maintain a constant shell-side level. The water is evaporated and exhausted to the atmosphere through a 14 inch line penetrating the containment shell. If the normal water supply f ails, the shell side is automatically supplied with water through a 3 inch line from the high pressure service water (HPSW) system.
2.2 SYSTEMS NORMALLY AVAILABLE FOR CORE COOLING FOLLOWING LOSS OF COOLANT ACCIDENTS (LOCA)
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 :equired core cooling for the full spectrum of pipe breaks:
1.
It sprays water directly over the core z.nd thus it is suitable for both above-core and below-core LOCAS.
2.
It is autcmatically actuated into operation by a key reactor parameter (reactor low water level of -12 inches or containment building pressure of
+5 psig).
3.
Its principal components are redundent.
4.
Its operation is independent of off-site power.
5.
It has sufficent head to inject water at any expected reactor system pressure.
FCR*.':NES 205 2cE0
0039 DOCUMENT NO PAGE O F-NUCLEAR ENERGY SERVICES, INC.
The low pressure ACS system includes two diesel-driven service water pumps which take suction f rom the river.
When containment building pressure exceeds 5 psig, both diesels will start autornatically and supply cooling water to the reactor vessel 4 inch nozzle through either of the two motor operated valves which open automatically when containm ent building pressure reaches 5 psig coincident with reactor vessel water level of -12 inches. The cooling water f alls f rom the 4 inch nozzle down through the tube bundle of the high pressure core spray system, impinges on the perf orated 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 distributico of the cooling water to all f uel elem ents.
Flow to the vessel commences when reactor vessel pressure drops to approximately 150 psig and f ull flow of approximately 900 gpm is reached when reactor vessel and containment vessel pressures are equalized on a LOCA.
A full range of breaks has been evaluated in the analysis of Loss of Coolant Acciends (LCCA) at LACBWR. Breaks of various sizes both below and above the reactor core were investigated.
These results have shown that the HPCS is adequate, even when limited to one pump, to maintain core parameters within 10 CFR 50 Appendix K criteria. The I.OCA analysis, as reported, also as';umed that feedwater flow would be t(rminated at the time of the break anc 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 shut off.
The plant responses to mitigate the consquences of a LOCA are fully automatic and are acutated without operator action required.
l FCOM:NES205 2 80
m00M DOCUMENT NO.
PAGE OF NUC' EAR ENERGY SERVICES, INC.
The only LOC A which requires AUS operation for short tecm 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 sufficent to keep the core cool (Reference 2).
The depressurization of the reactor to 150 psig has been calculated to take about 23 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.
2.3 FIRE PROTECTION W ATER SYSTEM DESCRIPTION Fire ;.rotection water at the Lacrosse Boiling Water Reactor is derived from a
.ombined usage water system called the High Pressure Service Water System (HPSW). Mississippi River water is supplied to this system via two automatic, diesel motor-driven vertical turbine fire pumps located in the crib house. These pump's each have a nominal rating of 750 gpm, 320 feet net head, at 1760 rpm.
'Ihe pumps are e onnected in parallel and discharge into a six inch steel underground main that loops the plant. One leg of the loop is run overhead through the grace tloor of the turbine building at the west end. This section of the main is isolable f rom the yard main by means of two underground, key-operated gate valves. The underground yard main supplies five, 6 inch outside fire hydrants which are spaced at approximately 200 foot intervals around the
,lant. All other HPSW services are fed from the overhead main in the turbine building.
HPSW system pressure is maintained at 30 to 120 psi by a 500 gpm,150 f oot net head electric motor-driven centrifugal pump located at the grade floor in the west end of the turbine building. This pump takes suction f rom the 16 inch Low PCGY NES205 2:80
81A0039 DOCUMENT NO HELg 10 24 PAGg NUCLEAR ENERGY SERVICES. INC-I Pressure Service Wate (LPSW) main. LPSW is supplied by two electric motor-driven vertical turbine pumps located in the crib house. A 1000 gallon surge tank located on the east end of the turbine building mezzaniane floor connects to HPSW header to stabilize HPSW system pressure.
Services suppaied by the HPSW sy, tem are:
1.
Turbine building wet-pipe, automatic sprinkler systems supply.
2.
Turbine building, reactor containment building, and waste handling building fire hose station supplies.
3.
High Pressure Core Spray System backup supply (loss of coolant accidents only).
4.
Alternate Core Spray System supply (loss of coolant accidents only).
5.
Shutdown Condenser Backup Supply.
6.
Eductor on the turbine condenser monitor-backup supply.
7.
Travelling screen wash water supply.
Except during ar. accident or fire, the consumption is limited to items 6 and 7.
This flow is about 75 gptr, mostly f or screen wash at the crib house.
I A simplified sketch c. the high cressure service water system. showing its interfaces with the lire protection system, ACS system, and the shutdown condenser, is provided in Figure 1.
FCR$.'sNES 205 2.30
81A0039 DOCUMENT NO.
11 mr
=
NUCLEAR ENERGY SERVICES. INC.
- 3. DISCUSSION OF EFFECTS OF LOSS OF CRIB rIOUSE ON RESIDUAL HEAT REMOVAL AND 19RE PROTECTION CAPABILITIES AT LACBWR 3.1 It is evident f rom the descriptions above that loss of the crib house and its associated piping has severe consequences, since it results in isolation of the plant f rom its ultimate heat sink f or both normal and accident conditions, and also deprives the fire protection system of its water source.
In particular, severance of the underground service water lines and destruction of the crib house will have the f ollowing eff ects:
3.1.1 Circulating water flow to the ma2n condenser is lost. The main condenser becomes unavailable as a heat sink.
3.1.2 Low pre ssure service water flow is lost. This eliminates the primar y supply of water to the HPSW system. This also directly results in termination of cooling water flow to the component cooling system coolers, the waste gas compressor lube oil cooler, waster gas compressor inter-and af ter-coole s, and recombiner condenser. Loss of LPSW to the CCW coolers result.s in elimination of the decay heat removal system as an available core cooling mechanism.
3.1.3 The entire HPSW sytem is depressurized anc rendered non-functional.This impacts saf ety related systems in several ways:
1.
HPSW is no longer available as a backup supply to the shutdown condenser.
2.
HPC'V is no longer available as a backup water source to the overhead storage tank (via the DMW system) or the HPCS system.
3.
No fire protection water is available f or the turbine building wet-pipe sprinkler system, or f or the interior or exterior hose. stations and hydrants.
4.
Alternate core spray becomes unavailable as either a stcrt-or long-term core cooling system in the event of a LOCA.
FCPA:NES 2C5 0 80
81A0039 DOCUMENT NO.
12 20 PAGE OF NUCLEAR ENERGY SERVICES. INC.
- 4. EMERGENCY SERVICE WATER SUPPLY SYSTEM ASSUMPTIONS, DESIGN REQUIREMENTS, AND CONCEPTUAL DESIGN 4.1 BASIC ASSUMPTIONS The following assumptions are basic to establishing design requirements for the Emergency Service Water Supply System (ESWSS):
4.1.1 It is assumed that all buildings and systems at LACBWR other than the crib house and its associated pemps and piping remain undamaged and functional, with one exception: off-site powei is assumed lost. Emergency on-site AC power is assumed to be available.
4.1.2 It is postulated that a fire will not occur concurrently with a loss of cool-ant accident (in the short-term; in the long-terr.i, fires may occur). For the purpose of this study, the short-term is defined as the period of time required to re-fill the vessel and cover the core. The long-term is defined as that period of time, f ollowing the short-term, in which coolant injection requirements are limited to restoring inventory depleted by boil-of f of water by decay heat.
4.1.3 For non-LOCA cases,it is arbitrarily assumed no fire occures within 21/2 hours f rom reactor shutdown.
4.2 GENERAL DESIGN CRITERIA General design criteria f or the ESWSS are as f ollows:
4.2.1 The system shall be capable of supplying sufficient water to ensure adequate core cooling and decay heat removal capabilities f or all reactor shutdown atuations up to and including the MCA.
4.2.2 The ESWSS sha!! allow f or decay heat removal and controlled cool-down to cold shutdown using norm ally available procedures, s ystems, and components to as great an extent as possible. Intentional degradation of rcm = nES 20s 2.80
81A0039 DOCUMENT NO.
PAGE OF NUCLEAR ENERGY SERVICES. INC.
existing, operable systenc for the purpose of implementing ESWSS operation shall be avoided.
4.2.3 ESWSS pump drives, valve operators, and controls shall be independent of off-site power supplies.
4.2.4 The ESWSS shall have the capability of supplying the internal fire hose stations and sprinkler systems with adequate fire protection water.
4.2.5 The ESWSS will be placed into operation manually.
4.3 CONCEPTUAL DESIGN 4.3.1 Imolementatier Of General Design Requirements
!! is evident frem consideration of the system descriptions in Section 1 and from study of Figure 1 that the HPSW header in the turbine building serves a pivotal role as the link between the river water supply and the ACS system, fire protection system, and all HPSW primary and bac'kup services.
This suggests that in the event of pipe breaks in the underground service water lines or destruction of the crib house, continuity of all these services can most easily be secured by means of the ;oilowing design modifications:
1.
Install accessible isolation valves on all connecting lines between tne un ground service mains and the HPSW header, at locatiens within the turbine building. Closure of these valves immediately f ellowing the seismic event preserves the integrity of the HPSW 1 eader by isolating it from the postulated breaks in the buried lines or crib house.
2.
Provide additional fixed or portable river water pumps whose operation.or readiness will not be affected by the maximum credib'e seismic event at the site. These pumps must provide adequate flow at suf ficient head to satisfy all safety-related demands which can be generated.
FCOM :NES 205 2 80
?
81A0039 DOCUMENT NO.
PAGE OF NUCLEAR ENERGY SERVICES. INC.
3.
Provide fixed or temporary conduits which are likewise immune from earthquake effects, to convey river water to the turbine building.
4.
Make provisions for new connections to the HPSW header in the turbine building to allow for introduction of emergency service water from the ESWSS.
If, in addition, provisiens are made to circulate ESW through the tube side of CCW ccolers (supplied by fire hoses from the HPSW header),
the capability to use the decay heat ccoling system in its normal design mode is restored. Such a system will, in principle, satisfy the general design criteria established in 4.2.1 through 4.2.5 above. This system is shown in conceptual outline in Figure 2.
4.3.2 Pump Capacity and Head Requirements and System Implementatien Time Requirements Determinatien of these parameters is best made by looking individually at each scenario with which the system is dcsigned to cope.
For each scenario described below, it is assumed that an SSE has occured, and that the crib house and underground piping are destroyed.
In addition, assumptions 4.1.1 to 4.1.3 apply. It is expected that the LOCA scenarios will control both the capacity requirements and time-to-implement requirement.
1.
Non-LOQA_
For any shutdown for which there is no LOCA or loss of primary system inventory, reactor decay heat removal and cooldown will take place using the shutdown condenser (to 470 F or below) followed by the decay heat cooling system. Since the normal source of cooling water to the shutdown condenser, the DMW system, is assumed operable, there is no immediate requirement f or ESWSS operation. In
81A0039 DOCUMENT NO.
PAGE OF g
NUCLEAR ENERGY SERVICES, INC.
fact, calculations show that even with only one-half of the DMW stortge tank inventory available, decay heat could be removed by the shutdown condenser alone for approximately 2-1/2 days. Obviously, this scenario is not controlling from the standpoint of system implementation time requirements.
Following initial cooldown and depressurization by means of the shutdown condenser, the deacy heat system will be placed into operation using ESWSS flow to the CCW cooler in place of LPSW.
Cooling water flowrates en the order of 200 gpm to the CCW coolers will be required.
2.
LOCA A.
Below-Core Breaks For below-core breaks the HPCS system will automatically initie*e and maintain coo! ant flow to the core f or the short-term. The absolute minimum available duration of HPCS flow, based upon two HPCS pumps operating and no makeup to the overhead storage tank (OHST) is 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. At least this much time, theref ore, will be available f or ESWSS implementation.
For below-core breaks it is highly desirable to flood the containment building as quickly as possible up to the level of the core midplane. This allows the shutdown condenser to be used to dissipate core decay heat to the environment, and core cooling to become independent of long-term pum ped-water iruection.
This function would ordinarily be serve:d by the ACS system, by which one or both ACS pumps would inject river water through the ACS nozzle at the top of the rez.ctor vessel, through the core, and cut through the break into the containment vessel.
FCP.i NES 2CS 2 E
81A0039 DOCUMENT NO.
PAGE OF NUCLEAR ENERGY SERVICES, INC.
I i
In the event of the less of the crib house, this function must be assumed by the ESWSS.
This suggests a required minimum pump head / capacity of 320 feet, 750 gpm fer the ESWSS pump (s). This is the equivalent of a single ACS pump, one of which has been shown to be capable of providing adequate core cooling / vessel flooding flow following a large below-core LOCA.
B.
Above-Core Breaks Above-core breaks may consist of breaks in the main stream lines, shutdown condenser inlet / outlet lines, or in the ACS or HPCS connections.
For above-core breaks it is desirable to minimize the quantity of water dumped into the containmcat vessel, since f or these breaks the shutdown condenser can not be used to dissipate heat to the environment. Overflooding of the containment vessel complicates post-accident pressure control'since it significantly reduces containment free volume.
For the case of a break in a steam line or shutdown condenser inlet / outlet line, the HPCS system will autorna.ically provide core cooling flow within a very short time of the break. The LACBf/R ECCS analysis has shown that +he minimum HCPS flow rate of 50 gpm is adequate for short-term cooling and in f act will maintaincontinuous wetting of the fuel rod surfaces within 3 to 4 minutes af ter the reactor has been scrammed.
Consquently, itis obvious that this same flow rate will be satisfactory for long-term cooling with the HPCS/HPSW system.
Also, calculationshave been performed which show that the HPCS/OHST short-term cooling system has sufficent inventory (15,000 gallonminimum) to reestablish the reactor vessel water level above the core f or any above-core break.
FCA'.t :NES 205 2.80
A DOCUMENT NO.
PAGE OF NUCLEAR ENERGY SERVICES. INC.
e This suggests that the ESWSS would not have to provide ACS, but would merely be required to provide makeup to the OHST via the HPSW system to support long-term cooling by the HPCS/ Low pressure core spray systems. For this function,100 gpm would be more than sufficient, and would be required no sooner than 2-1/2 hours af ter the start of HPCS.
For a break i.1 the ACS line between the ACS nozzle in the reactor head and the ACS line check valve, identical requirements are generated fc-the ESWSS is for the case of a steam line break.
A different situation arises with a break in the HPCS system.
For this scenario,,the only means initially available to carry off decay heat is the upward flow of steam and water through the core resulting frem the blow-down of primary system inventory through the break. For the largest HPCS break, the duration of this blowdown period is 23 minutes. Theref ore, core injection from the ESh 55 (via the ACS piping and nozzle) must be available no later than 23 minutes following the initiation of the LOC A event. An ACS equivalent flowrate of 900 gpm is desirable for rapid reflood of the reactor vessel to cover the core.
3.
Fire Protection Reauirements By assumption, fires are not postulated to occur in the "short-term".
Therefore, fire scenarios au not control ESWSS implementation time.
Reference 3 demonstrated that a single diesel-driven, 750 gpm 320 foot net head pump was capable of providing sufficient fire water f or a major fire in the turbine building, while simultaneously providing suf ficient cooling water to the shutdown condensers to dissipate the design decay heat load.
FCR'.?:NES2052SO
DOCUMENT NO.
PAG 6 OF NUCLEAR ENERGY SERVICES, lNC.
Study of the pump heau,.apacity curves reproduced in Reference 3 indicates that a pump with similar characteristics has considerable capacity beyond rated conditions and could, easily handle simultaneous fire-protection flow and CCW cooler tube side flow.
Using calculated results from Reference 3 as guidance, fire protection demand is approximately 161 + 472 = 633 gpm; estimated LPSW demand is approximately 200 gpm; total demand 333 gpm. An ACS-equivalent purrp can easily accomodate this demand.
4.
Conclusions ESWS$ set-up and initiation time is determined by the HPCS-line LOCA scenario:
2S minutes from time of break.
Capacity requirements are determined by the below core break scenario: 900 gpm to the ACS nozzle f or containment vessel flooding.
This requirement (along with fire protection requirements) suggest a pump rated at 750 gpm and 320 foot head, rc=u s t.Es 205 2 80
81A0039 DOCUMENT NO.
19 PAGE OF 24
. NUCLEAR ENERGY SERVICES. INC.
- 5. GENERAL DESCRIPTION OF DPC-PROPOSED ESWSS IMPLEMENTATION DPC proposes to have available two semi-portable, trailer-mounted pumps f or ESWSS service. When not in use these pumps will be stored in the turbine building. If the need f or ESWSS arises, the pumps will be transported to the river, suction hoses will be connected and placed into the water; fire hoses will then be routed f rom the pump discharge connections to designated siamese / hose valve connections installed in the HPSW header inside the turbine building. The pumps will then be started and primed.
When sufficient discharge pressure is available, all valves between the pur' o discharge and HPSW header are opened, allowing river water to fill the header and restore HPSW and related services.
i l
CCOM
- .ES 2:5 2/80 I
DOCUMENT NO.
20 24 OF NUCLEAR ENERGY SERVICES. INC.
- 6. REFERENCES 1.
Letter from Dennis Crutchfield (NRC) to Frank Linder (DPC),
Subject:
Liquef action Potential at LACBWR, dated August 6,1980.
2.
NES SI A0019," Single Failure Analysis of the LACBWR Emergency Core Cooling Systems", November,1975.
3.
NES 81 A0036, Revision 1, " Lacrosse Boiling Water Reactor Fire Protection System Combined Water Demand Analysis", January,1980.
- CAM : NES 2Cs 2. 80
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81A0039 DOCUMENT NO.
22 24 PAGE OF NUCLEAR ENERG SERVICES, INC.
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DOCUMENT NO.
PAGE OF F
NUCLEAR ENERGY SERVICES. INC.
TABLE 1 ESWSS REQUIRED FLOW RATES AND IMPLEMENTATION TIMES FOR VARIOUS ACCIDENT SCENARIOS Scenario: (Earthquake, Time Available for Plant Maximum Required Loss of cribhouse, loss Personnel to Make ESWSS Flows for Orderly of Of f-site Power, +'")
Fully Operational Cooldowr., Decay Heat Removal / Accident Mitigation And Fire Protection No other damage 2 days 200 gpm to CCW cociers as replacement f or LPSW No other damage 2-1/2 hours 200 gpm to CCW but major fire in
+ 633 gpm f or turbine building Fire Protection occurs
= S33 gpm turbine total flow Below-core LOC A 2-1/2 hours 900 gpm f or continuous core cooli
- and building reflood up to core midplane Below-core LOC A 2-1/2 hours Same as above until with major fire in building is flooded turbine building to core midplane in long-term level; then 633 gpm Fire Protection +60 gpm maximum boil-off makeup - o93 gpm total All above-core LOC AS 2-1/2 hours 100 gpm makeup except HPCS line up to OHST f er break long term supply to HPCS pumps / low-pressure core spray line Same as above 2-1/2 hours 633 gpm - fire with turbine protection + 100 building fire gpm OHST assumed in the makeup = 73 > gpm leng-term total cCAV2'CC "
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00B DOCUMENT NO.
PAGE OF NUCLEAR ENERGY SERVICES, INC.
TABLE 1 (Cont)
(Continued)
Scenario: (Earthquake, Time Available f or Plant Maximum Required Loss of cribhouse, loss Personnel to Make ESWSS Flows f or Orderly of Of f-sitePower, +'")
Fully Operational Cooldown, Decay Heat Removal / Accident Mitigation And Fire Protection HPCS line rupture 23 minutes 900 gpm f or rapid core reflood Same as above with 23 minutes Same as above; turbine building fire then: 633 gpm for assumed in the long-fire protection +
term 60 gpm f or cecay heat boil-off makeup = 693 gpm total flow 9
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ATTACHMENT 2' DAMES-& MOORE STATEMENT OF SURFACE STABILITY BEHAVIOR OF LACBWR SITE 00RIlm AND AFTER-AN SSE PRODUCItiG 0.20 PEAK }iDRIZONTAL GR0lEID SURFACE, ACCELERA l
It was nentioned earlier that the probability of occurrence of
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1 an SSE associated with a 0.129 peak ground surface acceleration was very low. The probability of occurrence of an SSE associated with 0.20g peak grou.nd surface acceleration is significantly lower at LACBWR site.
Extensive investigatier. sand analyses perfonned have shown that the free field areas of the site will be under marginal conditions 1
during the'O.29 carthouake.
llowever,' soils under the critical i
I structures supported by piles posses adetecte p3rgin of safety cgainst liquefaction. The above conclusion was substanticted by neasured r
1 high 5Pi-N Yalues amidst pik clusters.
The portion of the LACBWR site closer to the river may experience measurable strains during the strong motion oeriod.assoe!ated with a
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0.29 peak ground-acceleration..Some collapse of river bank may tske f
place. Minor p9ve:nent cracks n3y aopear at the surface due to differential h
settlenents. - However, the LACDWR sands are highly raervicus and will therefore most probably diuicete the excess pore pressures generated durino the SSE relatively fast.
Because of the fast redistribtufon of excess pore pressur e, no mjor cost carth:aste ' flows', or ' sink holes' I
are expected to develop.
Based on a knowledge of LACWR site conditions and reported da lage during similar carthquakes, we believe that any surficial obstructions created as a reFult of an SSE of.0.Eg Deak ground surface acceleration can easily be noneuvered by a: vehicle carrying the proposed dedicated water system equipent.during an e ergency cooling oNretion, gs 9 @,
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~ ATTACHMENT 2 DAMES & MOORE-STATEMENT OF' SURFACE STABILITY-BEHAVIOR Or LACBWR SITE DURING' AND 4FTER A!i SSE PR000CING,0 EoVEAX HORIZONTAL GROUND SURFACE,f.C_CLERAT10ii:
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Extensive field and laboratory investigations have been-ocrformed on LACBWR plant site soils to evaluate _the stability of the site under seismic icadings. All the analyses performed so far to evaluate the liquefaction potentici at the LACBWR site have repeatediv shown thet:
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- 1) the probability of an SSE occurring at the site during the remainina
. plant life is extrenely ' low; 2)'in-the event of an occurrence of an SSE, the nargin of safety against liovefaction potential is adequate under the free-field concition; and 3) the surgin of safety against liquefaction cr. der the tenttinent, tsrbine building and other critical structures is higher than that under the free-field conditions.
Sarause of the above findings and based on reported descriptions of danaces csused t'y aarthquakes sir.iiar to the costulated SSE at similar sites, we believe that the dansges to tha subsurf ce and surficial soils will n:t he Wiceable during-the very short duretion of strong rotion essociated with the SSE.
Therefore, we believe that the-e will be little or no hinderance
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to the movement of vehicles used to tr6nsport the proposed dedicated water systen e:vicaant for the ener,ency cooling oceratiens.
After 1
the earthouake motion ceases, the ground surface is expected-to be stable with mini;nal noticeable displacement.
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