ML20032B208
| ML20032B208 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 10/30/1981 |
| From: | Brooks W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20032B200 | List: |
| References | |
| NUDOCS 8111050291 | |
| Download: ML20032B208 (3) | |
Text
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.. M 10/30/81' UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
HOUSTON LIGHTING AND POWER COMPANY )
Docket No. 50-466
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(Allens Creek Nuclear Generating
)
Station, Unit 1)
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NRC STAFF TESTIMONY OF WALTER L. BROOKS REGARDING DOHERTY CONTENTION 15 y.
Please state your name and position with NRC.
A.
My name is Walter L. Brooks.
I am employed by the U.S. Nuclear Regulatory Commission as a Senior Reactor Physicist in the Core Performance Branch.
I have previously testified in this proceeding.
Q.
What is the purpose of your testimony?
A.
The purpose of rqy testimony is to respond to Doherty Contention 15, which reads as follows:
Intervenor contends his health and safety interests are inadequately protected because the industry standard power excursion theory (WIGLE) is in-adequate to represent the increase in heat energy due to rapid increase in reactivity in a Design Based Power Excursion Accident (DB-PEA).
Experi-ments reported in IN-1370 Large Core Dynamics, pp. 48-87, where a burst of neutrons was %jected in the side of reactor, give results which when compared to WIGLE, indicate this industry standard DB-PEA theory might underpredict the energy yield of a power excursion theory predicts the energy yield per gram of fuel in a PEA will be about 70%
of the design safety limit (280 cal /ga) for fuel rods.
'See Regulatory Guide 1.77, May 1974, PSAR, Montague I & II, pp. 4.3-29, and 15.143-55.)
1 Further, the National Reactor Testing Station 8111050291 811030 PDRADOCK05000g
. (NRTS) recommended in 1970, a special research program to resolve this underprediction, (IN-1370,
- p. 18).
Hence, Intervenor contends that Applicant's one dimensional time code (described in Supp. No. 2 to the SER for this system because the product generated is too small compared to data resulting from the neutron burst experiments reported in IN-1370 (supra), as is the data generated by WIGLE.
(Note: This Intervenor does not contend Applicant's NSSS vendor uses WIGLE or relies upon it, but rather that Applicant's analytic method generates the scram reactivity function for the DB-PEA theory as does WIGLE.)
Hence, Applicant or Applicant's NSSS provider should be required to provide data from power excursion tests from full scale reactors as was recommended by the AEC's test laboratory in 1954 (See
" International Report," PTR-738, "A Review of the Generalized Reactivity for Water-Colled and -Moderated UO2 Fueled Power Reactor," G. O. Bright, Titsetal.),
and the BWR system be redesigned to reduc reactivity poter.tial.
Q.
To be sure that the matter is clarified, is the WIGLE code used in the analysis of the design basis power excursion accident?
A.
It is not.
Q.
Is the WIGLE code used in the analysis of g transient or accident?
A.
It is not.
Q.
Is the WIGLE code used to obtain the scram curve for the design basis power excursion accident?
A.
It is not.
Q.
Is any one-dimensional space-time transient code used to obtain the scran curve for the design basis power excursion accident?
A.
No.
Let me explain. Mr. Doherty appears to have been confused by the Staff's reference to a one-dimensional code in the section of the
i
, Allens Creek SER ich he cites in the contention. That SER section presents a brief outline of the physics calculational models u*ed by the Applicant.
In the Staff's discussion of these models, we noted that a one-dimensional, time-dependent, spatial calculation was used to generate the scram reactivity. The Applicant uses that scram reactivity.. derived in the manner described, in his pont-kinetics plant transir; t, analyses.
Thes>3 scram reactivity results were not used by the Applicant in performing his control-rod-drop, accident analyses.
Instead, the scram curve for the de:ign basis powe excursion (rod drop) accident is obtained from a three-dimensional static code described in NED0-10527 entitled " Rod Drop Accident Analysis For Large Boiling Water Reactors." That code would underpredict the amount of reactivity inserted by scram and would thus be conservative.
y.
Is the W1GLE code comparable at all, as Mr. Doherty suggests, to the analytical method used by the Applicant to generate the scram reactivity results for the design basis power excursior. accident?
A.
Nc. The NED0 report cited above, together with its two supplements involve, as _ stated, multi-dimensional calculations and complex calculational procedures necessary to obtain a conservative, bounding type analysis for the design basis power excursion event. This analysis has little semblance to the rather simplistic WIGLE code.
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