ML20009E311

Transcript of MR Lane 810720 Testimony Re Doherty Contention 41 (Reactor Water Level Indicators).Prof Qualifications Encl

ML20009E311
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	07/20/1981
From:	Lane M GENERAL ELECTRIC CO.
To:
References
NUDOCS 8107280035
Download: ML20009E311 (9)

Text

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July 20, 1981

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. ! q, 1 UNITED STATES OF AMERICA Y' NUCLEAR REGULATORY COMMISSION 2

, L_

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD L' 3

4 In the Matter of S s  :

5 HOUSTON LIGHTING & POWER COMPANY Docket No. q-3Nd 6

(Allens Creek Nuclear Generating Station, Unit 1) S S7 ,

3 7 2 jul T N

• o 8

DIRECT TESTIMONY OF MIMON R. MNE ON DO @i Th h # f '

CONTENTION 41, - REACTOR WATER LEVEL INDI $RS 9

Q. Please state your name and place of emplo M 10 A. My name is Milton R. Lane and I am employed by 11 General Electric Company as a Principal Engineer. My business 12 address is 175 Curtner Avenue, San Jose, California 95125.

13 Q. Please describe your professional qualifications.

14 A. My professional qualifications are set forth in 15 Attachment MRL-l to this testimony.

16 Q. What is the purpose of your testimony?

17 A. The purpose of my testimony is to address the 18 consolidation r" Mr. Doherty's Contention #41 and TexPirg's 19 Additior.al Contention #54 which states that Applicant's reactor water level indicators are unreliable, as indicated 20 by events at Three Mile Island and Oystar Creek. Intervenor 3

contends that the Applicant should develop a system whereby the reactor water level is sensed by redundant to type and redundant to function water level indicators.

24 po3

_, _ 5 8107280035 810720 I PDR ADOCK 05000466 T PDR

1 Q. Has the reactor water level instrumentation been 2 recently reviewed?

3 A. Yes. As indicated in Appendix 0, Page -94 of 4 the ACNGS PSAR, the water level instrumentation has been 5 reviewed in response to Item II.K.3.23 of NUREG-0718, 6 " Licensing Requirements for Pending Applications for Con-7 struction Permits and Manufacturing License."

8 Q. As background information please give the purpose 9 of the reactor water level indication system?

10 A. The purpose of the reactor water level indicator 11 system is to provide the reactor operator and various safety 12 systems with information regarding vessel water level. The 13 BWR water level instrumentation provides multiple level 14 indications displayed on the reactor control console or near-In addition, 15 by panels in full view of the operator.

16 multiple indicating trip units provide wide range and narrow 17 range reactor level safety related trip signals and related 18 alarms. Safety control and information functions provided 19 by the level instruments include scram, containment isolation, 20 ECCS initiation, RCIC initiation, ADS initiation (contribu-21 tion), feedwater control, recirculation pump shutoff, MSIV 22 closure, level readout, level recording, and level alarm 23 functions in the control room for normal, transient, and 24 post-accident ce"ditions.

1 Q. Describe the present BWR reactor water level 2 indicator system uhich will be used on ACNGS.

3 A. Reactor vessel water level is measured by differen-4 tial pressure transmitters which measure the difference in 5 static head between two columns of water. One column is a 6 " cold" (ambient temperature) reference leg outside the reactor vessel; the other is the reactor water in the annulus 7

g area inside the reactor vessel. The measured differential pressure is a function of reactor water level.

9 The cold reference leg is filled and maintained 10 full of condensate by a condensing chamber at its top which continuously condenses reactor steam and drains excess con-densate back to the reactor vessel through the upper level tap connection to the condensing chamber. The upper vessel level tap connection is located in the ' team zone above the 15 normal water level inside the vessel. Thus the reference 16 leg presents a constant reference static head of water to 17 one side of the differential pressure (d/p) transmitter.

18 The other side of the transmitter is piped to a lower-level 19 tap on the reactor vessel which is located belou the normal 20 The low-pressure side of the water level in the vessel.

2~1 transmitter thus senses the static head of water / steam 22 inside the vessel above the lower vessel level instrument 23 tap. This head varies as a functic. of reactor water level 24 above the tap and is the " variable leg" in the differential 1 pressure measured by the transmitter. Lower taps for 2 various instruments are located at various levels to accom-3 modate both narrow and wide range level measurements.

4 Q. Describe the redundancy of function of the level 5 indicator system.

6 A. The Allens Creek reactor vessel water level indica-7 tion system is shown schematically in Attachment MRL-2. This

+

8 system provides redundant channels of level indication of g

overlapping ranges of instrumentation which are calibrated f r specific plant conditions. Four redundant and separate 0

instrumentation channels are provided for the normal operating range. Three redundant and separate channels are provided for transients which can be postulated during reactor operation and LOCA and post-LOCA conditions. One shutdown range channel 14 is provided for cold shutdown and maintenance conditions 15 when the reactor is depressurized and flooded. One upset 16 i range channel is provided for monitoring high water level 17 transients. The shutdown and upset range instruments 18 are effectively redundant for monitoring water levels up to

the main steam lines. Two redundant and separate fuel zone O channels provide water level recording / indication for accident i

21 monitoring when the reactor is hot and depressurized. Also, 22 in response to the recent review of water level instrumentation 23 mentioned earlier, water level is to be recorded continuously 24 from the bottom of the core support plate to the centerline

(

)

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1 of the main steam line for post-accident monitoring.

0 Describe the reliability of differential pressure 2

3 cells for water level measurement.

A. The differential pressure detectors currently 4

us d for water level measurement are recognized in the industry 5

as among the best quality available and have a history of being highly reliable. These sensors are rugged, contain few moving parts, are very accurate and have been fully qualified to recognized IEEE standards and NRC regulatori 9

guides. 3 complete redundancy in function of the water 10 level detection instruments which is provided is sufficient to 11 accommodate any ?.nticipated failure of the differential

. 12 pressure detectors.

13 Q. Describe the Oyster Creek incident referenced in 14 the contention.

~5 1

A. A loss of feedwater transients at the Oyster Creek l 16 facility on May 2, 1979 resulted in a significant reduction j 17 in water inventory above the reactor core areas as measured f 18 by one set of water 1% vel instruments (triple-low level) ,

19 while the remaining two sets cf level instrumentation in the 20 reactor annulus indicated water levels above the scram 21 setpoint.

22 The initiating event was a false high reactor 23 pressure scram. The signal resulted in a simultaneous l

i 2 reactor scram and the tripping c,f all operating recirculation

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1 pumps. Operator error led to the closure of all recircula-2 tion discharge valves. Oyster Creek is a non-jet pump BWR, 3 and, as such water from the reactor annulus area must pass 4

through the recirculation piping in order to reach the r ctor vessel lower plenum and the core shroud area. With 5

the recirculation pump discharge valves closed (discharge 6

I piping is over 2 f aet in diameter) , tLe only flow path back to the core region was via the 2-inch bypass lines around 8

the discharge valves. ".'he ef f ect of this flow restriction was to reduce the water level in the core region and to ir crease the level in the reactor annulus area.

11 Q. Did " spurious water level indication" at Oyster 12 Creek cause the operator to fail m take action as indicated 13 by the Intervenor?

14 A. No. There was no spuriour water level indication.

15 The water level instrumentation gave accurate indications of 16 water level in the reactor annulus and the core shroud area.

~7 1

The problem was operator error, as I explained above.

~8 1

Could this event scenario happen at ACNGS?

Q.

19 A. No. ACNGS is a jet-pump plant, and as such, there 20 is no way of restricting the flow from the reactor annulus 21 to the core shroud area for this type of event. Thus, the 22 water in the reactor annulus will accurately reflect the 23 water level in the core shrced area in a similar condition.

24 Q. Is the design of ACMGS comparable to TMI as 1 alleged in the contention?

2 A. No. The water level in the reactor vessel at TMI 3 was not directly measured, but was inferred from water level 4 instruments on the pressurizer. The pressurizer is a separate 5

vessel connected to the primary system of-PWR plants which 6

normally contains both steam and water and is used to maintain 7

system pressure such that roiling does not occu. in the g reactor. On BWR/6 plants such as ACNGS the reactor vessel wa er eve s m asur d re y and continuously. Thus the 9

I tervenor's reference to TMI is not applicable to ACNGS.

Q. What are your conclusions?

A. The reactor water level instrumentation at ACNGS 12 provides a direct measurement of reactor water level within 13 the vessel using reliable senscrs which are redundant as to 14 function.

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Attachment MRL-1 MILTON R. LANE My name is Milton R. Lane. My business address is General Electric Company, 175 Curtner Avenue, San Jose, California, where I currently hold the position of Principal Engineer in the Nuclear Power Systems Engineering Department.

I was graduated from the University of Maine College-of Engineering in 1953 with a Bachelor of Science Degree in Electrical Engineering, and with high distinction. During my junior year, I was elected to the engineering honor society,.

Tau Beta Pi, and during my senior year to Phi Kappa Phi, while also serving as student chairman of'the Local Chapter.

of the American Institute of Electrical Engineers. During my last year at the University of Maine I took many graduate courses ant completed nearly 50% of the hours necessary for a Masters Degree in Electrical Engineering.

My working career since receiving my B.S. degree in 1953 has been entirely with the General Electric Company and concerned with the Nuclear Plants Control, Instrumentation

& Electrical Engineering. I spent three years at the Knolls Atomic Power Laboratory in Niskayuna, New York, and then three years at the Atomic Power Development Associates Engineering offices in Detroit, Michigan on the Fermi Project. In 1959 I moved to California with the Atomic Power Equipment Depart-ment and have been there since that time in various capacities involving control and instrumentation and electrical design of nuclear power plant systems. I have participated in the preparation of preliminary and final safety analysis reports for various plants and in the design and evaluation of instru-mentation and control for safety systems. In my present capacity I provide technical leadership and individual con-tribution to develop systems design criteria and I work with systems engineers to resolve special control and instrument problems.

I serve on IEEE-NPEC working groups and I participate in the review and critique of Industry Codes and Standards

and NRC Regulatory Guides.

l l During the last two years I have participated in L review of the reactor vessel level instrumentation in' response to concerns which were precipitated by the TMI event.

l i

l 1

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STEAM SPACE SHUTDOWN RANGE REFERENCE , g g I

CCNDENSATE l CHAMBER. yPSET TYPICAL, OF 4

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(TYPICAL CF 9 NR D/P CELLS) '

b UPSET /SHUTDCWN LEVEL INITI ATE

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2. INDCAT10'JRECORD AND TMIP UNITS FCR WIDE RANGE USE COMMCN j D/P CELLS 1

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