ML20031H400

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Testimony of Dl Peterson Re Doherty Connection 28 Re Control Rod Ejection.Prior Affidavit Adequately Addresses 790731 Version of Contention
ML20031H400
Person / Time
Site: Allens Creek File:Houston Lighting and Power Company icon.png
Issue date: 09/18/1981
From: Peterson D
GENERAL ELECTRIC CO., HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML20031H319 List:
References
NUDOCS 8110270445
Download: ML20031H400 (2)
v • d • e


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Scptemb r 18, 1981 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 2

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 3

In the Matter of

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S HOUSTON LIGHTING & POWER COMPANY S

Docket No. 50-466 S

I (Allens Creek Nuclear Generating S

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Station, Unit 1)

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DIRECT TESTIMONY OF DONALD L. PETERSON 3

REGARDING DOHERTY CONTENTION NO. 28 - CONTROL ROD EJECTION C

Q.

Mr. Petersd'n, have you reviewed your prior affidavit

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10 on Doherty Contention No. 28, which affidavit is attached 11 hereto as Attachment DLP-l?

12 A.

Yes I have.

Q.

Are the statements contained thereir still true and 13, correct?

74 A.

Yes they are.

Q.

Do you believe your affidavit addresses Mr. Doherty's 16 contention as it is described at page 42 of the Board's Order 17 of September 1, 19817 s0 A.

Yes.

I have reviewed my prior affidavit in light.

13 of the statement of Doherty Contention 28 set forth in the 20 Board's Second Order Ruling Upon Motions for Summary Disposition, and I have determined that the prior affidavit adequately addresses the July 31, 1979 version of the contention.

To reiterate, there is no rod ejection accident with consequences more severe than the rod drop event.

In this regard, I l

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8110270445 810918 PDR ADOCK 05000466 T

PDR

1 specifically addressed the scenario whereby tha'cntiro Control 2

Rod Drive and housing become detached from the vessel (see 8-10), which is the specific event discussed Affidavit, pp.

3 in the contention.

I have also reviewed Mr. Doherty's reply to Applicant's l

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Motion for Summary Disposition wherein Mr. Doherty alleges 5 i that the drive housing weld would fail because of bulging strains and distortions.

The two exhibits to Mr. Doherty's I

reply indicated the alleged source of bulging is a damaging 9

pressure pulse following a rod drop.

Mr. Holt claw and Dr.

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' 10 williams have previously addressed this issue and have shown that this is no pressure pulse associated with a rod drop 33 accident (Tr. 11718-12035).

Furthermore, bulging of the reactor itself would not affect the ability of the rod 13 support structure to prevent housing ejection, as the support 14 beams are welded to the reactor pedestal.

Even if we 1

assumed failure of the drive housing weld, the rod cannot be

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16 ejected more than three inches.

17 Finally, I would note that the design of the big h

13 Rock Point reactor is not comparable to Allens Creek.

The reported cracking is in a part which has been eliminated in the gg Allens Creek design, and furthermore, does not increase the likelihood of a rod ejection.

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Attcchmsnt DLP-1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER COMPANY

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f Docket No. 50-466 (Allens Creek Nuclear Generating

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Station, Unit No.1)

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AFFIDAVIT OF DONALD L. PETERSON State o'f California County of Santa Clara I, Donald L. Peterson, Manager Control Rod Drive Line, 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 state-ments contained in the attached pages and accompanying exhibits are true and correct to the best of my knowledge and belief.

Executed at San Jose, California, July 29,1980.

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SubsJp@gM86296M&cMM&fe this 29 day of July

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NOTARY PUsuc.CAUFORNIA g S but SANTA CLARA COUNTY J

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Attcchm3nt DLP-1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of S

S HCUSTON LIGHTING & POWER S

COMPANY S

Docket No. 50-466 5

(Allens Creek Nuclear S

Generating Station, Unit S

No. 1)

S Affidavit of Donald L. Peterson

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My name is Donald L. Peterson.

I am employed at General Electric Company as a Professional Nuclear Engineer, as I have been for 25 years.

A statement of my experience l~

and qualifications is set out in Attachment 1.

This affidavit' responds to Intervenor Doherty's Contention No. 28 which po.stulates that a control rod can be ejected by containment pressure or by the pressure in 4

i the SCRAM Discharge Volume Tank (SDVT).

Ir.tervenor contends that such a rod ejection would create a more rapid reacti-vity insertion than a rod drop, the design basis reactivity insertion accident.

As support for this assertion, Inter-venor relies on the power excursion accident at the Sta-tionary Low Power Reactor (SL-1) in January.1961.

l

I.

Introduction For gross control of reactivity in the Allens Creek reactor, cruciform control blades are inserted be-i-

tween the fuel assemblies.

These control blades, or rods, i

r.nclose smaller rods filled with boren carbide, a n6utron absorbing material.

The reactor is controlled by driving these control blades (177 for ACNGS) into the reactor core to reduce reactivity (and thus power) cnd withdrawing the a

rods to increase reactivity (and power).

The control blades are moved by, hydraulic drives which insert or re-i tract the blades in small increments and continuously drive the blades in on a shutdown signal.

II.

The Rod Drop Event As testified to by Mr. Stirn in an affidavit also filed'in this proceeding, it is possible that a rapid 1/

removal of a high worth control rod could result in a potentially significant power excursion.

The design basis accident dealing with rapid removal of a control rod is the rod drop accident.

The worst case credible control rod drop accident for the ACNGS design is described as follows:

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Control rod worth is the measure.of reactivity which will be added to.the nuclear reaction is the rod is re. moved.

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a)

A fully inserted control rod drive must sustain a complete rupture, break, or disconnection from its crucife u control blade at or nunr the coupling.

b)

The blade must stick in the fully. inserted posi-tion as the rod drive is withdrawn.

c)

The blade falls by gravity after the rod drive is i

fully withdrawn.

As explained by Mr. Stirn, the above' described I

sequence assumes that the Rod Pattern Control System (RPCS) j is operating.

This is a reasonable assumption because the Rod Pattern Control System (RPCS) is a safety related, hard-wired, dual channel system which must operate to move 2,

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control _ rods and which cannot be bypassed by the operator.

The RPCS limits the amount of reactivity which may be inserted into the core by a dropped rod under the assump-tions listed above.

This is accomp.1.ished by physically preventing rod movement if the reactivity worth of the control rod chosen for movement by the operator is not in a 3/

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proper preplanned sequence which minimi::es rod worth.

2/

For a fuller description of the RPCS, see the affidavit of Mr. Lesyna, filed in this docket concurrently.

L 3f The sequence is provided in detail on pages 7.7-8.8 of the ACNGS PSAR.

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The enforced sequence assumes that the magnitude of the maximum control rod worth will be limited such that unac-ceptable fuel damage cannot occur in the event of a rod drop.

Hence, with the RPCS operating, the peak fuel enthalpies that can result frcm the postulated drop cf i

these rods has been shown to be well below the energy 4

deposition (i.e.,

fuel peak enthalpy) design limit thereby resulting in inconsequential fuel damage.

III. The Hvpcthesized Rod Ejection Accident According to Intervenor's contention, a control rod might be removed faster than has been assumed under the assumptions used for the rod drop event.

This-could theoretically produce higher fuel enthalpies (and thus worse fuel damage) than predicted.

Intervenor hypothesizes that containment pressure and/or pressure from the SCRAM Discharge Volume Tank (SDVT) would act to drive the rod out faster than would gravity acting on a decoupled rod.

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A.

The Rod Control System In order to understand Intervenor's conention, a simplified understanding of the Control Rod Drive (CRD)

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Hydraulic Control Units (HCU) is helpful.

A simple illus-tration is attached as Exhibit A.

The CRD is primarily a

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piston connected to the control rod and driven up or down by varying the pressure differential across the piston.

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During a SCRAlt, water at several hundred pounds per square inch above reactor vessel pressure automatically enters the CRD below the drive piston and pushes the rod into the core.

In order to displace the water above the piston for such a rapid insertion, a special flow volume called the scram discharge volume is available.

The scram discharge volume l

consists of header piping which connects to each control rod hydraulic control unit and drains into an instrument volume.

The header piping is sized to receive and contain all the

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water discharged by the drives during a scram, independant cf the instrument volume During normal plant operation the scram discha:ge volume is empty, and vented to atmosphere thrcugh an open vent and drain valve.

When a scram occurs a signal from the Reactor' Protection System closes these vent and drain valves i

to conserve reactor water.

Lights in the control room indi-cate the position of these valves.

During a scram, the scram discharge volume partially fills with water displaced from above the drive pistons.

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Folleving the scram, the control rod drive seal leakage from the reactor continues to flow into tne scram discharge 4/

There are two leakage paths across the drive seals:

one from the CRD supply pumps and another from re-actor pressure.

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c volume until the discharge volume pressure equals the reactcr vessel pressure.

A check valve in each HCU prevents reverse flow from the scram discharge header volume to tha drive.

When the initial scram signal is cleared from the Reactor Protection System, the scram discharge volume signal is overridden witn a key lock override switch, and the scram discharge volume is drained and returned to atmospheric pressure.

3.

The Hypothetical Accident As explained above, the SDV is at atmospheric pressure at all times except after a SCRAM.

Hence, at these times the pressure from the SDV could not approach the value necessary to move any of the 177 CRD pistons.

After a SCRAM, l-as the SDV is filled by seal leakage frem the reactor, the pressure above the CRD piston is at all times equal to re-i actor pressure.

Consequently, the fact that the SDV has It pressurized to reactor pressure is of no consequence.

would add no additional pressure to drive the rod out of the core.

Furthermore, since check calves are in all 177 lines i

l leading to the SDV, any pressure helf by the SDV could not be transmitted back to the CRD unit.

Obvicusly, then, the I

pressure er lack of it in the SDV has no effect on the Con-trol Rod Drives.

-Intervenor has also suggested that containment pressure would add to or produce a rod ejection event.

Since the Control Rod Drive units as well as their associated Hydraulic Control Units are sealed, there is no way in which the drive water would see containment pressure unless the system were ruptured.

Since pressures in the CRD units are close to reactor operating pressure (several hundred psi) and contair. ment design pressure is at a maximum 15 psi, the effect of a rupture on the control rod drive would be to relieve pressure above or below the drive piston.

The re-

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lease of pressure below the piston in discussed in detail below; the consequences are well accounted for.

The release of pressure above the piston would drive the control red into the core.

Hence, containment pressure could nct contribute to a rod ejection.

Although not covered by Intervenor's contention, one can postulate an event in which the pressure below the CRD drive piston is relieved by some incredible failure while i

reactor pressure acts above the pistons.

However, even if the pressure below the drive piston were relieved with the drive latched (as it would normally be), no control rod withdrawal would occur.

The CRD col'let latch mechanism would hold the the CRD in place because the collet is held in the latching position with a heavy re urn spring and requires at.

a

least 100 psi above reactor pressure in the area over the piston to unlatch.

If the pressure were relieved below the drive piston while the CRD was being moved (unlatched posi-tion), the hydraulic pressure which unlatches the CRD would drop and spring force would cause the collet to latch and stop rod withdrawal.

Furthermore, even if the collet were assumed to jam in the open position, the steady-state control rcd withdrawal velocity would be 2 ft/sec compared with 5 ft/sec for a rod drop event.

This number was calculated by i

an analysis of the forces and hydraulic losses in the drive 4

during this postulated event.

Since lower withdrawal velocities result in lesser reactivity insertions (and thus lower peak

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enthalpies and potential fuel damage), this postulated event does not produce a rod ejection accident that compares with the rod drop event.

One other event which might be postulated is the failure of the drive housing such that the entire Control Rod Drive and housing becomes detached from the vessel.

This scenario would result in the following sequence of events:

The housing would separate from the vessel.

The control rod drive and housing would be blown downward against

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the support structure by reactor pressure acting on the l

l cross-sectional area of the housing and the drive.

The downward motion of the drive and associated parts would be H

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determined by the gap between the bottom of the drive and the suppcrt structure and by the deflection of the support structure under impact.

The support structure is composed of a series of deep I beams located under the reactor vessel and attached to the reactor pedestal (not the reactor vessel).

Suspended from these beams are tie rods attached to support blocks placed under the individual drives.

Shock-absorbing springs at the tie rod / beam connection limit the maximum housing movement to aporoximately three inches under the loads anticipated from reactor pressure and driveline weight.

In the current design, maximum deflection of the support structure after impact by the CRD housing is approximately 3 inches.

If the collet were to remain latched, no further f-control rod ejection would occur and the housing would not I

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drop far enough to clear the vessel penetration.

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If the basic housing failure were to occur while m

the control red is being withdrawn (this is a small fraction of the "otal drive operation time) and if the collet were to stay unlatched, the following sequence of events is pos-(

sible:

The housing would separate frem the vessel.

The 1

the con-drive and housing would be blown downward against trol rod drive housing support.

Since in this instance the r

i control rod is unlatched, the control rod will continue to withdraw a'fter the control rod drive housing has been 5

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stopped by the drive housing support.

The equivalent steady-state rod withdrawal velocity for this occurrence would be 0.3 ft/sec-5/ as compared with 5 ft/sec for the rod drop event.

Clearly, this event also does not produce a rod ejection accident with consequences more adverse than the rod drop event.

The withdrawal speed for the instance where pressure below the CRD piston has been removed and the collet fails (2 ft/sec) is higher than the velocity for the case where the CRD housing has detached and the collet fails (.3 ft/sec) because in

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nhe former case reactor pressure above the CRD drives the rod out with no pressure below the drive piston.

In the latter case pressure below the piston is not removed during the event.

There are no other credible rod ejection scenarios with more adverse consequences because the failure of all known mechanisms which could produce uncontrolled withdrawl of control rods have been reviewed and analyzed.

III.

The SL-1 Event Intervenor has used the SL-1 accident to support

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the contention that a rod ejection event would produce a 5/

This withdrawal velocity was calculated by the same means used to determine the withdrawal speed for the instance where pressure is relieved below the drive pisten. _

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l power excursion larger than a rod drop event.

The SL-1 was a military pressurized water reactor which used control rods inserted into the top of the reactor.

At the time of the accident, the reactor was shutdown (all controls roda in) and the control rod drives were undergoing maintenance.

Results of investigation showed that a control rod, detached from its drive, had been renoved to a point, where the reactor began a power excursion.

As a result of the power excursion, control rods were blown out of the core and the i

reactor was demolished.

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The SL-1 reactor and accident are in no way rele-vant to ACNGS.

First, when the ACNGS reactor is shut down l..

i (all control rods in), the complete removal of one control l-rod would not bring the reactor critical.

Secondly, rod i

ejection for ACNGS is physically prevented or the effects are bounded by the rod drop event.

The SL-1 reactor was not 1

so designed.

Hence, Intervenor's reference to the SL-1 event to support this contention is totally inappropriate.

IV.

Conclusion After thoroughly analyzing the range of events which could rapidly remove a control rod from the reactor, it is apparent that the consecuences of a design basis rod drop event are bounding.

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ATTACHMENT 1 PROFESSIONAL QUALIFICATIONS DONALD L. PETERSON CONTROL ROD DRIVE'LINE DESIGN GENERAL ELECTRIC COMPANY My name is Donald L. Peterson.

My business ad-dress is 175 Curtner Avenue, San Jose, California 95125.

I am Manager, Control Rod Drive Line Design Unit for the Boiling Water Reactor System Department of the General Electric Company.

In this position I am responsible for the design of components such as the control rod drive, control rod, hydraulic control unit and other components associated with the control rod drive system.

I have a Bachelors Degree in Mechanical Engineering i~

from the University of Washington, Seattle.

After two years of service a's a Gunnery Officer in the U.S. Navy and a snort term as an Instructor in Mechanical Engineering Department of the University of Washington, I'

joined General Electric Company.

Prior to transferring to General Electric Company's boiling water reactor operations, I worked at the Hanford operation for 7 1/2 yez My re-L sponsibilities included design of remote handling facilities, test facilities and control rod drive mechanisms.

For the past 25 years I have been responsible for the design of components for boiling water reactors, being u

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the responsible design engineer for control rod drives on early General Electric reactors.

I have continuously served as Technical Leader or Manager of the group responsible for control rod drives and associated components since that time.

I am a Professional Engineer in the State of California.

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