ML20136B514
| ML20136B514 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/04/1979 |
| From: | Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| NUDOCS 7909060405 | |
| Download: ML20136B514 (3) | |
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101 M ARIETTA FTEEET, N.W.
ATLANTA. GEORGI A 3o3o3 e
2 APR 0 41979 e
SSINS 8150
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./V MEMORANDUM FOR:
E. L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations, IE:HQ FROM:
R. C. Lewis, Acting Chief, Reactor Operations and Nuclear Support Branch, Region II
SUBJECT:
POTENTIAL PROBLEMS IDENTTTIED AT BABCOCK AND WILCOX FACILITIES IN REGARD TO THE THREE MILE ISLAND OCCURRENCE Based upon preliminary information from Reactor Inspectors dispatched to the Crystal River and Oconee Facilities to investigate generic concerns of the Three Mile Island incident, the following potential problems have been identified.
Some of these matters may be known and evaluated, however, they were identified in our initial review.
Specifically, the following actual or potential problems have been identified:
1.
Most transients from high power cause fluctuations in pressurizer level that require operator action to correct.
In some instances 3
these actions result in primary system cooldown.
2.
Pressurizer level instrumentatation is not safety related beyond General Design Criterion 13 of Appendix A to 10CFR50, therefore, pressurizer level is not necessarily available to mitigate an accident or for post accident recovery.
In addition, this instrumentation does not have sealed reference legs which means the reference leg may flash during transients resulting in erroneous level indication.
These features differ from Westinghouse PVR design which considers pressurizer i
level safety related and provides reactor trip and safeguards functions and further designed with sealed reference legs to prevent flashing.
3.
The integrated control system (ICS) which controls primary and secondary systems, including feedwater, steam generator level, turbine, steam j
bypass and rod control are not safety related and are powered from a i
single vital electrical bus. In addition. pressurizer level and pressure controls are not safety related and powered from single pc,wer supply.
Problems associated with this design include the following:
a.
During transients the turbine is sometimes' automatically trans-ferred to manual at fixed load while the ICS is reducing reactor power.
This creates a mismatch transient between turbine load and reactor power.
CCNTACT:
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APR 0 41979 E. L. Jordan b.
The failure of the ICS power supply causes all controls to demand 50% which places transient on plant if power level is at other than 50%.
4.
In regard to IEB-79-05, both Crystal River and Oconee Plant procedures require the operator to secure reactor coolant pumps when the pressurizer empties during an accident. This procedure negates the assertion in IEB 79-05, Enclosure 2, that void formation on emptying the pressurizer are dispensed by force flow. In addition, the evaluation provided in IEB 79-05 also addresses transients as a result of a loss of off site power. In all cases, the loss of site power is assumed to result in a loss of the reactor coolant pumps.
This fact would certainly negate the forced flow assertion.
5.
The accident analysis for a loss of feedwater flow from high power indicates in some cases that the pressurizer may go solid resulting in j
passing water through the power operated relief valve and the safety 4
valves on the pressurizer. Two concerns as a result of this possibility l
have been identified.
a.
The safety valves have been designed to pass water, a concern exists as to whether the power operated relief is also designed to pass water?
b.
There is a limitorque motor operated isolation valve down stream of the power operated relief.
The concern for this valve is whether the valve is designed to close against full designed flow through the power operated relief?
6.
Should a small loss of coolant exist, the current system design and operating procedures allow for automatic release of reactor coolant external to the containment. These releases occur via the sump pumps and reactor coolant drain system which pump automatically to tanks outside the containment. This release would continue until automatic safeguards actuation occurs isolating containment.
In additico, the release of coolant could also result in the loss of water inventory l
assumed to be in the containment sump for NpSH requirements for decay heat pumps during recirculation phase, 7.
Neither facility has hydrogen recombiners and rely strictly upon veuting of containment for hydrogen control following an accident.
The review of these items indicates general applicability of all items at both facilities except items 1 and 3.b which do not appear to be problems at the Oconee Units.
In order for a thorough evaluation of generic aspects of the Three Mile Island incident to other Babcock and Wilcox facilities, the above itema j
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APR 0 41979 E. L. Jordan should be investigated at. other B & W facilities. We are presently continuing our investigation into these items at crystal River and Oconee.
We will advise when more detailed or additional information is available.
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W R. C. Lewis, Acting Chief Reactor Operations and Nuclear Support Branch ec:
B. H. Grier, RI J. G. Keppler, RIII K. V. Seyfrit, RIV R. H. Engelken, RV
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