ML19247A460
| ML19247A460 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/23/1979 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19247A452 | List: |
| References | |
| NUDOCS 7907310581 | |
| Download: ML19247A460 (3) | |
Text
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UNITE D STATES PJUCLEAR REGULATORY COMMisSIOfJ W ASHING IO N, D. C. 205L5 APR 2 31979
";MORANDUM FOR:
D. B. Vassallo, Assistant Director for Light Water Reactors, DPM FROM:
R. L. Tedesco, Assistant Director for Reactor Safety, DSS
SUBJECT:
BOARD NOTIFICATION - ANALYSIS OF SINGLE ROD DROP EVENT -
WESTINGHOUSE PLANTS Westinghouse (W) inforned us (telephone call on March 28, 1979 from K. Jordan of W to R. Tedesco of NRC) of a Part 21 notification concer-ning the analysis of the single rod drop event.
A neeting was held on April 12, 1979 hetween the NRC and W to discuss this natter further.
Infornation from this neeting is preseEted in this nenorandun.
The sinqle rod drop is a ONB-limited transient considered to be a modcrate f reque-ncy transient.
The calculated consequences for this event are decendent upon whether the reactor is being operated in an automatic or nanu31 control node.
The Part 21 notificacion is conce, ned with the analysis of a single rod drop event with the reactor in the automatic control mode.
If a sinqle rod drop event occurs when the reactor is in the automatic mode, the reactor control systen responds to both the reactor power drop (nismatch between turbine power and reactor power) and the decrease in the core average temperature and attempts to restore both quantities to their eriqinal values.
This restoration of reactor power by the reactor control system nay result in some power evershoot depending upon the excore rower sinnal that is used.
For the SARs this analysis has been perforcled usina a point ki netics model to track the core oower tine-dependent behavior.
In 2-and 4-lcop W plants the power signal to the reactor control systen is either auctioneered from amanq the four excore detectors to obtain a high value or averaced.
This results in a power overshoot of the riagni-tude predicted in the SAR.
In a 3-loop W plant the situation is somewhat dif ferent in th3t tne reactor control system cD1; ins its power signal frcm a dedicated excore detector.
Recent spatial analyses by W indicate that for a dropNd rod in the core quadrant adjacent to the dedicated e.; core detector, the pNer overshoot is greater than the value calculated by the rethods used in the sal.
This could then lead to exceediaq the D.'.3 l i m i t.
For the SAR analysis, no credit is taken for the negative 488 278 700731067/l
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APR 2 31979
' O. c. ','i s pfi l o '
s flux rate trip for the plants which have this feature; credit is taken for the turbine runback and rod stop for those plants which have these f eatures except for the Indian Point-3 4-loop plant.
These recent W calculations which nodel spatial effects have considered, amng other tiiings, different fuel cycles and times in fuel cycles.
Based on these calculations, W has proposed changes to the high negative flux rate tri p setpoints to assuro that all dropped rod events result in a reactor trip for plants which have this feature.
The negative flux rate : rip setpoint would be decreased tron 5 percent to 3 percent and the rate-lag tine constant would be decreased from 2 seconds to 1 second.
flo changes are proposed to plants which have the turtaine runback and rod stop features.
NRC Office Letter fio. 19 calls for a determination of the safety sigri-ficance of new information by evaluating "whether the information could reasonably tie regarded as putting a new or _ different light upon an issue befo re Boards or raising a new issue."
Staff evaluation has reached a stag which concludes that this matter could be interpreted as putting a net or different lient on the analysis of the single rod drop event.
More infornation from a source or sources external to the staff is requi ed to further study this issue.
I therefore conclude that appro-priate Boards should be ratified.
In accordance with NRR Office Letter No.19 requirenents, I ao providir g you the following:
1.
the item of notification; 2.
considerations regarding relevancy and materiality; 3.
statenent of perceived significance; and 4.
relation to projects.
1.
The Iten A recent analysis of the single rod drop event in Westinghou;e 3-loop plants when the reacter is operating in the automatic control mode and which have the negative flux rate trip feature indicates that the power overshoot is larger than stated or inplied in the analysis provided in the SAR.
This power over-snoot could lead to exceeding DN3 limits.
2.
Relevancy and Materiality _
Thi s analysis deficiency is relevant and material to all Westing '
house 3-loop pl3nts which have the negative flux rate trip feature.
779 ASS
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Arf ( 2 31979 f
. Vas)/ ilo 9 3.
_Si oni fica nce Since the recent analysis of the single rod drop event for a Westing-house 3-loop plant which has the negative flux rate trip feature indicates a larcer power overshoot than for the analysis in the SARs, Westinghoure has recommended that the affected plants change the setpoint and ' ate-lag time constant in the negative flux rate trip circuit to ensure a reactor trip when a single rod drops.
This change is in a conservativ; direction frcm a safety standpoint.
4.
Relation to Projects For plants that the CP review has been completed (i.e., SER published, I recommend that letters be sent to the applicants stating that this cencern will be reviewed at the OL stage.
If a Licensing Board is in existence, they are also to be notified.
For plants that the OL review has been completed (i.e., SER published, I recom,end tnat letters be sent to the applicants requesting that they address this concern prior to the issuance of an OL.
If the Licensing Board is in existence, they are also to be notified.
For any plants currently under review in the CP or OL stage, the applicant will be requested to address this issue during the normal review process.
For oper ating reactors, DDR rorommends making the cha 'e to the set-point and rate-lag time constant in the negative flux dte trip feature of af fected plants.
Westinghouse intends to provide the NRC with a topical report on this matter in about six-nonths.
m
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Gr'j ].-d W +OO Robert L. Tedesco, Assistant Director for Reactor Safety Division of Systems Safety cc:
R. Mattsen/F. Schroeder D. Ross V. Stello/D. Eiserhut P. Check A. Schwoncer K. K r. i el D. Fieno
( b 3nce