ML20062H572

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Affidavit in Support of Summary Disposition of Jf Doherty Contention 33 Re Doppler Effect.Ge Mathematical Model Used to Calculate Doppler Effect Does Not Rely on Spert Test Data.Prof Qualifications Encl.Pp 235-241
ML20062H572
Person / Time
Site: Allens Creek File:Houston Lighting and Power Company icon.png
Issue date: 07/29/1980
From:
BAKER & BOTTS, HOUSTON LIGHTING & POWER CO., LOWENSTEIN, NEWMAN, REIS, AXELRAD & TOLL
To:
Shared Package
ML19331C559 List:
References
ISSUANCES-CP, NUDOCS 8008190177
Download: ML20062H572 (7)
v • d • e


Text

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O UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING 6 POWER' COMPANY)

Docket No. 50-466 (Allens Creek Nuclear Generating Station, Unit No.1)

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AFFIDAVIT OF RICHARD C. STIRN State of California County of Santa Clara I, Richard C. Stirn, Manager of Core and Fuel System Design within the Nuclear Power Systems Engineering Department of the General Electric Company, of lawful age, being first duly sworn, upon my oath certify that the statements contained in the attached pages and accompanying exhibits are true and correct to the best of my knowledge and belief.

Executed at San Jose, Califo rnia July

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_,e Subscribed and sworn to before me this

,j (/ day of July, 1980.

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of S

S HOUSTON LIGHTING & POWER S

COMPANY S

Docket No. 50-466 S

(Allens Creek Nuclear S

Generating Station, Unit S

No. 1)

S Affidavit of Richard C.

Stirn l

i My name is Richard C.

Stirn.

I am employed at General Electric Company as a Professional Nuclear Engineer.

1 I have been so employed for fifteen years.

A statement of my experience and qualifications is set out in Attachment 1.

I.

Introduction The purpose of this affidavit is to address Mr.

Doherty's contention that the negative reactivity effect

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r from Doppler broadening has been overstated because General l

Electric used experimental results obtained during testing l

when particles of fuel dispersed into the reactor coolant l

rather than upon tests using a contained, pelletized oxide form.

II.

Doppler Reactivity Effect 1

In nuclear reactor physics calculations, the l

probability that a given neutron will be absorbed by a i

236 l

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nucleus is dependent upon the energy of that particular neutron and the inherent absorption characteristics of that particular nucleus.

The absorption characteristic of a given nucleus is given by the absorption " cross section" and is represented in units of area.

The absorption cross section is variable over a range of neutron energies and is unique to that nucleus.

There exists for every particular nucleus specific neutron energy ranges (resonances) at which the probability of absorption is very high compared to neutrons of other energies.

As fuel temperature in a reactor increases, the velocities of the target nuclei increases.

This increase results in a conecmitant increase in the ranges of neutron energies with a high probability of absorption.

This

" broadening" of the absorption cross section is known as

" Doppler Broadening."

Doppler Broadening in a BWR results in a negative reactivity effect as reactor fuel temperature I

increases because the parasitic absorption (non-fissioning absorption) by Uranium-238 and Plutonium-240 produces the primary influence which greatly offsets the increased absorption in Uranium-235 and fissile Uranium-238.

III.

The Centention The sole support for Mr. Doherty's contention is the allegation that the General Electric topical report ~1/

1/

" Generation of Void and Doppler Reactivity Feedback for Application to BWR design," NEDO-20964 (December, 1975). 237 l

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analyzing the Doppler reactivity effect for BWRs relies on data generated from the Special Power Excursion Tests (SPERT) conducted at the Idaho Nuclear Experimental Laboratories.

Mr. Doherty further asserts that this reliance is erroneous because, during the SPERT tests, the fuel dispersed into the l

reactor coolant, thereby creating the appearance of Doppler feedback reactivity ~2/ which would not actually occur in an operating BWR with intact fuel pellets.-3/

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Quite simply, Mr.

4 Doherty's assertion is incorrect.

The General Electric mathematical mcdel developed to calculate the Doppler effect does not rely in any way on the SPERT test data.

The model was derived based on fundamental principles of Doppler Broadening known for decades and universally accepted.

The

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increase of neutron absorption cross sections as a function r

of temperature increases has been thoroughly investigated

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and experimentally quantified.

Knowing the character of this phenomenon, it is a straightforward. precess to analytically determine the resulting neutron population.

Inputs to the

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2/

Intervenor's reference to p. 15 of NEDO-20964 is also incorrect because this section discusses only mcderator void reactivity feedback, not Doppler reactivity feedback.

3/

The desica basis limit on scecific fuel enthalov adocted cy the NRC is'230 calories per gram.

This limit is incorporated

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in Section 4.2 of the Standard Review Plan because it has been demonstrated that fuel dispersal will not occur below this limit. 238

model are empirical (experimentally determined) cross sections equally applicable to all uranium fueled reactors.

The model has been verified against :he widely-known Hallestrand tests. ~4/ These tests measured the temperature dependence of resonance neutron absorption in clad uranium dioxide fuel rods.

Lastly,.the model was compared to the most appropriate (e.g., not water-logged fuel) SPERT tests.

The results compared extremely well as shown on the graphs in Exhibit A taken from the General Electric topical "Ro1 Drop Accident Analysis For Large Boiling Water Reactors," NEDC-10527 (Figure 5-7).

Thus, there was no " reliance" by General Electric cn the SPERT results, only a secondary comparison to further support the Hellestrand verification of the Doppler reactivity model.

4/

E.

Hellestrand, P.

Blomberg and S. Horner, "The Temperature 5cefficient of the Rescnance Integral for Uranium Metal and Oxide," Nuclear Science and Engineering, Vol.

3, pp. 497-506 (1960).

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ATTACHMENT 1 QUALIFICATIONS Richard C. Stirn, Manager Core and Fuel Systens Eesign General Electric, hB G My name is Richard C. Stirn. My business address is 175 Ortner Avenue, Mail Code 740, Sm Jose, California, 95125.

I am a registered Professional Nuclear Engineer in the State of Califomia (hU 630). As Shnager I have the responsibility of directing core and fuel systems design for the General Electric Company, SEG.

I graduated from Tennessee Technological thiversity in 1962 where I re-ceived a Bachelor of Science Degree in Engineering Science. During the Summer of 1962 I worked for the Arnold Engineering Eevelopment Center in Tullahoma, Tennessee as an Engineer.

In the Fall of 1962 I entered Purdue Lhiversity on an AEC Fellowship, and in August of 1964 I received a Sbster of Science Ecgree in Nuclear Engi-neering. Upon completion of my studies at Purdue I entered the LSiversity of Arizona to work toward a PhD degree in Nuclear Engineering; however, I left school in Februa:y of 1965 to work for the General Electric Company, NEG before completing the PhD requirements.

Upon joining General Electric I entered the Engineering Training Program and had assignments dealing with light water moderated themal reactors, steam cooled fast reactors, and sodium cooled fast reactors. After com-pleting my training assignment in October 1967, I was appointed to the position of Technical Leader of Core Dynamics and Poactivity.

In June 1972 I was appointed to the position of Shnager of the Nuclear Safety Analysis Component of the Core Nuclear Engineering thit.

In this position I co-authored or contributed to three papers and three reports on the topic of nuclear reactor excursion analysis.

I also participated in the development of the control rod drop accident boundary value approach for the reload licensing submittal.

La September of 1974 I assumed my present responsibilities as Shnager, Core and Fuel Systems Design.

In this capacity I am responsible for the development of system requirements for the Core and Fuel Perfomance, Core Perfomance Transient, and Fuel Mechanical Systems. Mditional responsibili-ties include core ther-al hydraulics evaluations, the development and issuance of core physics design require ents for reactivity contral systems, the perfomance of criticality analyses of the fuel storage and handling facili-ties, and the development of core physics design bases for plant transients.

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