ML20031A767
| ML20031A767 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 09/17/1981 |
| From: | Crimmins T ALLEGHENY ELECTRIC COOPERATIVE, INC., PENNSYLVANIA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20031A766 | List: |
| References | |
| RTR-NUREG-0785, RTR-NUREG-785 NUDOCS 8109250392 | |
| Download: ML20031A767 (11) | |
Text
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RELATED CO20tESit.W.NCT4 September 17, 1981 UNITED STATES' OF AMERICA g('
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Docket Nos. 50-387 and
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50-388
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ALLEGHENY ELECTRIC COOPERATIVE, INC. )
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(Susquehanna Steam Electric Station, )
Units 1 and 2)
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TESTIMONY OF THOMAS M. CRIMMINS REGARDING CONTENTION 21 (Scram Discharge Volume Subsystem Pipe Breaks) 1.
Contention 21 states that "[T]here is a potentially dangerous flaw in the Applicants' reactor in the design of the primary cooling system inasmuch as radioactive water frem a i
break in the scram discharge volume subsystem can disable the major safety systems including the residual heat removal system, the reactor core isolation cooling system, the core sprays and the high pressure coolant injec?lon pumps in a brief period of l
l time".
2.
A recent Nuclear Regulatory Commission (NRC) report entitled " Safety Concerns Associated with Pipe Breaks in the BWR O$0 7
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.. Scram System" (NUREG 0785) describes a postulated sequence of events in which a br eak occurs in the Scram Discharge Volume (SDV) piping of r Boiling Water Reactor (BWR).
In the NRC.
postulated sequence of events, the damage resulting from the break includes the loss of all Emergency Core.Coviing F steos (ECCS).
This postulated loss is predicated on the fact that all the ECOS pumps are typically located on the basement floor of the reactor building.
The fluid f.isiharged from the postulated break in the SDV is assumed to be delivered to the reactor building basement elevation by a variety of flow paths, in-cluding not only the normal discharge path via floor drains but also down stairwells and through hatchways above the ECCS equipment.
NUREG 0785 (which is based or a BWR design typical of a Mark I containment) postulates that, eventually, all the ECCS pumps will fail due to the cascading of discharged water through stairwells and hatchways and onto the ECCS pump motor essemblies or due to the general flooding of the ECCS pump rooms themselves.
3.
The SDV is part of the Control Rod Drive (CRD)
System.
The CRD system at the Susquehanna Steam Electric Station (SSES) consists primarily of the CRD's, the CRD vithdrawal lines, two SDV's, and the valves associated with j
performing the control rod insertion (1,e., scram) function.
These components are depicted on the attached Figures 1, 3 and 4.
The CRD Syster, downstream of the scram exhaust valves, is w
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- normally dry and not pressurized during plant operation.
During a scram, water from the volume above the CRD piston within the CRD is discharged to the CRD withdrawal line.
It flows through the scram exhaust valves to the SDV.
The scram exhaust valves are normally closed but open upon receipt of a scram signal.
The SDV vent and drain valves are normally open during plant operation and close automatically upon receipt of the scram signal.
The SDV then partially fills with the water discharged from the CRD's.
Upon completion of the reactor scram the small quantity of water flowing past the CRD seals from the reactor, and w7cer flowing past the pisten seals from the CRD make-up pump continues ;o flow into the SDV.
This flow continues until the pressure in the scram discharge volume ic equal to the reactor pressure.
The scram signal is reset by the operator resulting in the scram exhaust valves closing and the SDV vent and drain valves opening.
The SDV then drains and returns to atmospheric pressure, configuring it for normal operation again.
4.
NUREG 0785 postulates a break in the SDV piping i
without regard to the high quality of material and fabrication procedurcs used in the SDV's.
The SDV's are designed to high standards commensurate with their function (4SME Section III, Class 2) and, therefore, failure of the SDV piping has an extremely low probability of occurrence.
In addition, the SDV will be subjected to in-service inspection in accordance with Section XI, Class 2 of the ASME code.
The SDV materials are l
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carbon steel (SA-lO6 Grade B) and highly re,sistant to cracking phenomenon.
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5.
If a leak were to develap, the majar cafety systems identified in the contention are protected against flooding.
The SSES is a BWR with a " Mark II"_ containment structure, and as such, is designed so that the Reactor Core Isolation Cooling (RCIC) system pump and the F79S equipment (the Residual Heat Removal (Riia) system pumps, the High Pressure Coolant Injection (HpCI) system pump and the Core Spray (CS) system pumps) are i
located on the bottom floor of the reactor building.
Their location is depicted on the attached Figures 2 and 4.
Ea ch train of the abcve systems including their respective pumps, is located in a compartment which is watertight with respect to one another.
In addition, the stairwells are provided w.th watertight dcars which isolate them from this equipment.
This desiEn prevents localized flooding from having broad (i.e.,
multisytstem) consequences.
6.
If the SDV's were to develop a crack, the leakage would be quickly detected and corrective action taken before the crack could propagate into a rupture.
The leak detection indi-cations available include area radiaticn monitor alarms, reactor building sump level alarm, reactor building exhaust vent high radiation alarms, reactor building differential pressure indi-cator, loss of reactor building ventilation alarms, ECCS and RCIC pump room level alarms, control rod drive high temperature e
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alarms, control rod position indicators, an,d direct operator observation of the leakage.
7.
In addition, the operators at the SSES receive training, including specific procedures, ubich addresses leakage occuring outside primary containment.
These will enable the operators to deal effectively with SDV laakage.
The corrective action taken by the operator includes two major actions:
A)
Maintaining coolant rater to the reactor vessel.
The preferred system for decay heat removal while the reactor rems.f.ns pressurized in-cludes thn use of the main feedwater pu:aps, the condensate pumps and the condensor.
In addition, the CRD make-up pumps have the capability for limited flow high pressure injection of coolant into the reactor system.
Also, the RHR service water system is capable of low pressure injection into the reactor system following reactor depressurization for the purpose of maintaining core cooling.
The main feedwater pumps, the condensate pumps, the CRD make up pumps and the condensor are located in the turbine building.
The RHR scrvice water pumps are located in the Emergency Service Water (ESW) pumphouse.
Since these two buildings are physically isolated from the reactor building where the SDV's are located, they are not subject to the flooding that is postulated to occur after an SDV break.
Both the main feedwater pumps and the RHR service water pumps are controlled from the main control room and provide adequate, independent, and physically remote capability to preserve core cooling following an SDV break.
ECCS equipment and the RCIC system are avail-able, if required, to maintain coolant water to the shutdown core for the removal of decay heat.
It is important to note that the con-4 sequences, in terr of loss of coolant inven-tory, of a broken SDV without closure of the scram exhaust valves is within the makeup
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. capability of any one of the nine, individual ECCS pumps or the RCIC, system.
B)
Reducing or eliminating coolant leakage.
The operator will reset the scram signal in accordance with operating procedures.
This maintains the control rods in the inserted position but closes the scram exhaust valves thus stopping the flow of ' water to the broken SDV.
If scram reset fails to isolate the leak, the reactor coolant syctem will be depressurized using normal procedures or manual actuation of the Automatic Depressurization System.
This significantly reduces the leakage rate.
The operators will then close the manual isolation valves in the scram withdrawal lines to isolate the broken SDV from the source of water.
Followin' isolation, corrective action in the form of suitable repair work will be undertaken.
8.
If, in spite of the watertight provisions described in paragraph 5 above, general area flooding we"e to occur, it would take a mit.imum of 2-1/2 hours to flood the basement of the reactor buildiag to a one foot depth.
This flooding rate does not take into account operation of the reactor building sump pumps which could reduce the water level nor does it account for j
the reduction in flooding rate which would be achieved by depres -
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surizing the reactor system.
Because the ECCS pump motc:s are l
l six feet above the floor, more than a "brief period of time" l
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the reactor and control the leakage which might occur after an SDV break.
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Given the abundance' of options available to respond to the postulated accident, the long period of time'which is available to respond in, the training given to operators, and the low probability of an SDV break occuring in the first place, it is my opinion that the SDV break event does not present a potentially dangerous flaw in design of the SRES.
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