ML19224C339
| ML19224C339 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/04/1979 |
| From: | Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jordan E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7907020176 | |
| Download: ML19224C339 (3) | |
Text
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[pMeg UNITED STATES NUCtEAR REGULATORY COMMISSION o
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2 101 MARIETTA STREET, N.W.
ATLANTA. GEORGIA 3o3o3
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APR 0 41979 SSINS 8150 MEMORANDUM FOR:
E. L. Jordan, Assistant Director for Technical Programs, Division of Reactor Operations, II:HQ FRCM:
R. C. Lewis, Acting Cblef, Reactor Operations and Nuclear Support Branch, Region II SU3 JECT:
POTENTIAL PROBLEMS IDENTIFIED AT BABC0CK AND WILCOX FACILITIES IN REGARD TO THE T11REE MILE ISLAND OCCURRENCE Dased upon preliminary infomation 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 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 1er =av flash during transients resulting in erroneous level indication.
These features differ from Westinghouse PWR design which considers pressurizer level safety related and provides reactor trip and safeguards functions and furtbar designed with scaled reference legs to orevent flashing.
3.
The integrated control system (ICS) which controls primary and secondary systems, including feedwater, steam generator level, turbine, steam bypass and rod control are not safety related and are powered from a single vital electrical bus. In addition. pressurizer level and pressure controls are not safety related and powered from single power 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 transteat between turbine load and reactor power.
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b.
The failure of the ICS power supply causes all controls to denand 50% which places transient on plant if power level 14 at other than 50%.
4.
In regard to IEB-79-05, both Crystal River and Oco.ee Plant procedures require the operator to secure reactor coolant pumps when the pressurizer empties during an accident. This procedu.e negates the astertion 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 7ower is assumed to result in 4 loss of the reactor coolant pumps. This fact would certainly negate the forced flow assertion.
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t 5.
The accident analysis for a loss of feedwater flow from high power indicates in some cases that the pressurizer may to solid resulting it passing water through the pouer operated relief calve and the safety valves on the pressurizer. Two concerns.as a result of this possibility have be:e identified.
a.
T.'.e safety valves have been designed to pass water, a concern ex.'sts as to whether the power oceea'ted relief is also designed to ) ass water?
b.
There is a limitorque motor operated isolation valve down stream of.he power operated relief.
The concern for this valve is whet her 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 a id operatir g procedures allow for autcmatic release of reactor coolant t externM to the containment. These releases occur via the sump pumps j and rea ttor coolant drain system which pump automatically t'o tanks outside 'be containment. This release would continue until automatic safeguardc actuation occurs isolating containment.
In addition, the release of coclant could also result in the loss of water inventory assumed to be in the containment sump for NPSH requirements for decay heat pumps during recirculation phase.
7.
Neither facility has hydrogen cecombiners and rely strictly upon venting of containment for hydrogen control following an accident.
The review of these items indicates general applicability of all items at 4
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 items om' s u
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APR 0 4 ;973 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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R. C. Lewis, Acting Chief Reactor Operations and Nuclz?.r Supa rc Branch cc:
B. H. Grier, RI J. G. Keppler, RIII K. V. Seyfrit, RIV R. H. Engelken, RV 49 264 293 TMI-2 50-320