ML19221A238
| ML19221A238 | |
| 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 | |
| NUDOCS 7905210047 | |
| Download: ML19221A238 (69) | |
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NUCLEAR REGULATORY CCMMISSION
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/E 3i MEMORANDUM FOR:
E. L. Jordan, Assistant Director for Technical Progr.us, Division of Reactor Operations, IE:HQ FROM:
R. C. Lewis, Acting Chief, Reactor Operations and Nuclear Support Branch, Region'II
SUBJECT:
POTENTIAL PROBLEMS IDENTTFIED AT BABCOCK AhT WILCOX FACILITIES I: REGARD TO THE THREE MILE ISLiND OCClTRRENCE Based upon preliminary ir. formation from Reactor Inspectors dispatched to the Crystal River and Oconee Facilities to investigate generic concerns of the Three Mile Island incident, the following potential problens 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 c o rre ct.
In some instances these actions result in pri=ary system cooldown.
2.
Pressurizer level ins t rumenta ta tion 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 instru=entation does not have sealed reference legs which r::eans the reference leg may flash during transir.ats resulting in erroneous level indication.
These features differ from Westinghouse PVR design which considers pressurizer 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 pricary and secondary systems, including feedwater, steam generator level, turbine, stean bypass and rod control are not safety related and are powered from a single vital electrical bus.
In addition. pressurizer level and pressure centrols are not safety related and powered from single power supply.
Problens associated with this design include the follcwing:
a.
Euring transients the turbine is scmetic es' automatically trans-ferred to manual at fixed load while the ICS is reducing reactor power.
This creates a ciscatch transient between turbine load and re.s cto r power.
CONTACT:
T. McHenry 70052100 0 l4 A
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APR 0 41979 E. L. Jordan b.
The failure of the ICS power supply causes all controls to de=and 50% which places transient on plant if power level is at other than 50%.
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4.
In regard to IEE-79-05, both Crystal River and Oconee Plant procedures i
require the operator to secure reactor coolant purps when the pressurizer empties during an accident.
This procedure negates the assertiu-un REB 79-05, Enclosure 2, that void formation on ecptying the pressurizer dispensed by force flow. In addition, the evaluation provided in are 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 pu=ps.
This fact would certainly negate the forced flow assertion.
5.
The accident analysis for a loss of feedwater flow from high power indicates in sc=e cases that the pressurizer may go solid resulttng in passing water through the power operated relief valve and the safety valves on the pressurizer. Two concerns as a result of this possibility 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 notor operated isolation valve down atrean 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
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external to the containment. These releases occur via the surp purps l
and reactor coolant drain system which pump autc=atically to tanis outside the containment. This release would continue until autocatic safeguards actuation occurs isolating containment.
In addition, the release of coolant could also result in the loss of water inventory assumed to be in the containment surp for NPSil requirements for decay heat purps during recirculation phase, 7.
Neither facility has hydrogen recombiners and rely strictly upon venting of containment for hydrogen control following an accident.
The review of these items indicates general applicability of all iters at both facilities except i t er.s I and 3.b which do not appear to be problers at the Ccocee Units.
In order for a thorough evaluation of generic aspects of the Three Mile Island incident to other Eabcock and ilcox racilities, the above iters
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APR 0 4 lc73 E. L. Jordan shoul<. 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.
()+m R. C. Lens, Acting Chief Reactor Operations and Nuclear Support Branch cc:
- 3. H. Grier, RI J. G. Keppler, RIII K. V. Seyfrit, RIV R. H. Engelken, RV b
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0-1 NRC "TAFF ISSUES ADVISORY TO
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BABCOCK & WILCOX OPERATING REACTORS
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m The NRC staf f has issued an Of fice of Inspection and Enforce-ment bulletin regarding the nuclear accident at the Three Mile Island Unit 2 reactor.
3 This bulletin is being sent to all utilities with an operating j
Babcock and Wilcox pressurized water reactor.
Babcock and Wilcox i
was the supplier of the Three Mile Island Unit 2 reactor.
There are seven other Babcock and Wilcox reactors in operation at five sites.
Unit 1 at the Three Mile Island site is shut down for refueling.
The utilities involved are being provided with early information on the Three Mile Island accident and are being asked to review immediately their facilities and procedures in light of the accident to preclude a similar event.
A formal response to the NRC is required in 10 days.
All other NRC reactor licensees and construction permit holders are being provided with this information as well, but no action is required.
In a related action, the NRC has dispatched a reactor inspector to every si:e with an operating Babcock and Wi!cox reactor.
These inspectors will function in a similar fashion to the resident inspectors who have been assigned to 16 operating rea: tor sites around the country.
(Eventually all operating sites and many con-struction sites will have resident inspectors.)
The newly assign 2d resident inspectors will conduct inspections of all phases of the facility operations in addition to the routine inspection program which is performed by inspection specialists from the NRC's Regional Of fices. The inspectors will be residing near the site and work on a full-time basis at their assigned sites.
(A list of the operating Babcock and Wilcox reactors is attached.)
Babcock & Wilcox Coerating Reactors
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,,,0conee,1, 2, 3 South Carolina Duke Power Co.
/ Rancho Seco California Sacramento Municipal Pcwer District Davis 3 esse 1 Ohio Toledo Edison Co.
Three Mile Island 1,
2 Pennsylvania etropolitan Edison Co.
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__ Arkansas 1 Arkansas Arkansas Pcwer & Light Co.
Crcstal River Florida FleEida Power Co.
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uh ATTACHMEtiT 1
~f-PRELIMIrlARY RECOMEtlDATI0 tis' q_
RECOMMEtlDATI0t15:
(1)
Verify that the auxiliary feedwater system is properly aligned and opera.ble (including automatic actuation.)
In the event of a loss of all "in feed flow, auxiliary" feed flow is essential for core cooling (ECCS ~
is not intended for this condition in a Westinghouse Plant.)
(2)
Verify operating procedures for failure of a relief of safety valve to close, failure of a pressurizer relief valve to recle' e is considered an ASME upset condition.
There procedures should recgon. 'e that the pressurizer will fill with water and that water could be ented if containment in the pressurizer relief tank failure disc.
These procedures should recognize the following points:
(a)
The isolation motor operated valve should be left to stop RCS blowdown through a power-operated relief valve when RCS pressure returns to a pre-relief valve actuation pressure.
(b)
Pressurizer steam bubble will continue, pressurizer will be water solid and waste relief will result, and this is to be excetted.
(c)
ECCS maintaining pressure, ECCS flow is necessary to maintain RCS pressure will above that corresponding to saturation temperature in hot leg or core cutlet.
(d)
Heat removal and cooldown by steam generator is needed.
(e)
ECCS oepration should continue until cold shutdown (below 200 degrees F) reached with further heat removal by RWR.
(3)
Recheck procedures for containment isolation and pumping from containment building sump to auxiliary building lielic waste storage tanks, I.E., sump pump operation.
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ATTACHMENT 1
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Based upon a preliminary review of the FSAR for Three Mile Island, the_.
differences listed below should be cons'dered in evaluating the applicability of the Three Mile Island incident sequence to your plant.
In evaluating your plant, there may be variations of the Three Mile Island accident j
sequence that will need to be considered.
i (1)
The once-thru sheerheat steam generators at Three Mile Island have considerable different characteristics than Westinghouse U-Tube units, in particulate.
(a) A loss of feed on Three Mile Island design thermally affects primary system conditions with a resultant pressure rise, this close coupling of effects between the primary and secondary systems is a function of mass flow and heat transfer characteristics.
On a Westinghouse unit, a loss of feed water will have little thermal effect in the primary system until a low steam generator water level reactor trip is reached.
Reactor trip will then reduce primary temperatures and pressures.
(b)
The Three Mile Island steam ger.erators have considerably less water inventory on the secondary line than Westinghouse steam generators, for Westinghouse plants, the larger inventory provides a larger heat removal capacity.
(c) "atural RCES circulation may have played a significant role during the transient, the Three Mile Island, tall, straight, steam generator necessities and so foot elevation difference between parts of the hot leg and cold leg piping which may have made it more susceptible to vanca blockage.
The difference in configuration leads to different natural circulation characteristics.
(2)
On Three Mile Island, turbine trip accest directly actuate a reactor trip, on all operating Westingnouse plants, a trubine trip is a reactor trip input signal at high power levels.
This feature avoids pressure increases in the primary side following a loss of load.
(3)
The Three Mile Island auxiliary feedwater system design includes one automatically actuated steam-turbine driver pump and two manually actuated motor driven pumps.
Westinghouse design feedwater pumps, all automatically actuated.
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ATTACH"ENT 1 PAGE 3
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The follcwing commentary relates to the behavior of the Westinghouse PWR plant following an inadvertent for depressurization caused by a stuck open pressuizer relief or safety valve, emergency procedures should be~~
checked to ensure that undesired actions which could complicate the transient are not taken.
The following preliminary information is a brief descripiton which may help you focus on the evaluation.
The expected response to an.open primary relief valve will be a rapid depresssurization of the ECS.
This will appear to the operator as decreasing pressurizer pressure with increasing pressurizer level. If the open valve is a power operated relief valve.
There is a motor operated isolation valve in the depressurization, if the depressurization continues, reactor trip, turbine trip, man feedwater isolations auxiliary feed activation, and safety injection actuation will occur automatically, RCS pressure will stabilize when safety injection flow balances the discharge from the relief valve.
It is inportant to maintain high pressure safety injection flow to keep pressure above saturation pressure for the hot leg temperature, similarly, it is important to maintian auxiliary feedwater flow to remove decay heat.
Depressurization should follow control in a gradual, controlled manner using safety injection in control pressure until AWR can be cut in.
Water for high pressure safety injection comes from the refueling water stchage tank, when the tank is empty, recirculation from the containment sump should be initiated, still using the high head or low head injection pumps as appropria ce to control pressure.
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