ML20031D684
| ML20031D684 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 10/09/1981 |
| From: | Hou S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20031D662 | List: |
| References | |
| NUDOCS 8110140015 | |
| Download: ML20031D684 (7) | |
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10/09/81 UNITED STATE 1 0F AMERICA NUCLEAR REGUu!.YORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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HOUSTON LIGHTING AND POWER COMPANY )
Docket No. 50-466
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(Allens Creek Nuclear Generating
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Station, Unit 1)
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NRC STAFF TESTIMONY OF SH0N-NIEN HOU REGARDING CONTROL R0D EJECTION
[Doherty Contention 28]
Q.
Please state your name and position with the NRC.
A.
My name is Shon-nien Hou.
I am a Principal Mechanical Engineer in the Mechanical Engineering Branch, Division of Engineering. A copy of qy professional qualifications is attached.
Q.
What is the purpose of your testimony?
A.
The purpose of this testimony is to respond to Doherty Contentior. 28 which basically alleges that certain off-normal conditions might develop within and with regard to the reactor pressure vessel such l
that control rod ejection can occur that would have consequences more serious in te(ms of reactivity insertion than the rod drop accident, the i
l design basis reactivity insertion accident.
Board's Order of l
September 1, 1981, p.42.
f Q.
Has the postulated control rod ejection accident, the sole concern of Contention 28, ever occurred in any operating Boiling Water Reactor i
(BWR) plant?
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A.
No. The rod ejection accident has never occurred in any operating BWRs. Although an accident did occur in the S-t-1 military reactor in 1961, it was a different type of reactor and occurred under unique circumstances. The S-L-1 was a Pressurized Water Reactor (PWR) of early design and the event occurred because established procedures during maintenance were not properly followed.
Q.
Why do you call the rod ejection in a BWR a " postulated" accident?
A.
If we look into the causes of rod ejection and how it might result in an accident, we have to assume that three low probability events occur simultaneously:
(1) An instantaneous and complete circumferential break of the Control Rod Drive (CRD) housing occurs near the connection weld to the vessel.
(2) The specific control rod is at that instant in an unlatched position which allows it to withdraw continuously due to a failure of the latch mechanism.
(3) The specific control rod withdrawn has to be of high worth.
Q.
Why is an instantaneous and complete separation of CRD housing from the reactor vessel a low probability event?
A.
The CRD housings at the vessel penetrations are made of Alloy 600 seamless steel pipe. The material is austenitic and is strong and ductile with a minimum tensile strength of 80,000 psi. The design is in accordance with the rules in Section III of ASME Code for Class 1 components with ample margin, good quality assurance, and adequate inservice inspection. The Staff's latest generic review of the primary piping of similar material has concluded that such piping and associated l
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welding mater.ial has ample fracture toughness, and that the potential for failure is very low even under the most severe flow transients and strong motion earthquake. Accordingly, the Staff believes that it is highly unlikely to have an instantaneous and complete circumferential pipe break at the vessel connection. We further believe that if an austenitic steel pipe did fail at the vessel connection, it is much more likely to have a detectable leakage crack prior to a complete circumfer-ential break. Thus, adequate inservice inspections should avert a complete break. However, the Staff still reviews the consequences of a complete break-away of CRD housing from the vessel and evaluates protective measures to assure that the Commission policy of defense-in-depth is satisfied.
Q.
Since a complete break-away of the CRD housing from the vessel is postulated, have you reviewed the protective measures in Allens Creek iluclear Generating Station, Unit 17 A.
Yes.
In Allens Creek Unit 1, there are assemblies of protective structures installed directly underneath the CRD housings during the reactor operation and these structures may be disassembled during the reactor shutdown. The design purpose of these protective structure assemblies is to serve as barriers to limit any housing drop distance to 3 inches in the postulated housing break-away accident.
Furthermore, a latch mechanism, the so-called " collet fingers", is built-in to prevent any unintended control rod withdrawal.
Consequently, the rod can only withdraw one notch distance (6 in.) at a time. The latch is always in a
" lock-in" position by the spring force and can only be temporarily released by a hydraulically pressured piston after receiving a releasing signal.
i Q.
Do you consider that those protective measures are adequate to ensure safety for the postulated control rod ejection accident?
A.
Yes.
From a Mechanical Engineer's point of view, I do not believe that a complete break-away of CRD housing from the vessel is going to happen. But, even if a housing break-away is postulated to occur, it can only happen under the following conditions:
(1)
If it should occur during reactor operation, since the rod and the drive are structurally linked and the rod is latched during normal plant operation, the rod will withdraw together with the housing to a distance of not more than 3 inches.
I have reviewed the magnitude of loads and allowable stresses used in the design of protective structure assemblies installed directly underneath the CRD housings.
I conclude that the loads are conservatively calculated and the allowable stresses are acceptable to ensure structural capability for limiting the drop distance of the CRD housing. A 3 in. withdrawal of any rod in any rod pattern during reactor operation will not constitute a safety problem and is bounded by the design basis rod drop accident.
(2)
If the postulated accident should occur during reactor operation and a rod of relative high worth is at the instant of unlatched position (which is again another event of extremely low probability of occurrence due to very brief period of unlatched time), the rod will be with-drawn to the next notch position (within 6 in.) in addition to the 3 in. drop with the housing.
Further withdrawal is unlikely since a latch mechanism failure (i.e., failure of the hydraulic system to exert pressure on the latch piston) could mean more difficulty for rod withdrawal due to incapacity of the latch release.
Again, rod withdrawal in such a short distance (9 in.) will not con-stitute a safety problem and is bounded by the design basis rod drop i
accident.
l (3)
If the postulated accident should occur during reactor shutdown, l
any single rod withdrawal when all other rods remain inserted will not create a critical condition.
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Q.
What is your conclusion with respect to this contention?
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A.
Based on the foregoing, I conclude that the protective measures l
l in Allens Creek Unit 1 are adequate to ensure public safety for the l
postulated control rod ejection accident. As indicated earlier, the l
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postulated accident is not only of low probability but if it should occur, it has no safety significance because of the designed safety features and protective raeasures.
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r PROFESSIONAL QUALIFICATIONS DR. SHOU-NIEN,HR U. S. NUCLEAR REGULATORY COMMISSION MECHANICAL ENGINEERING BRANCH DIVISION OF ENGINEEkiNG I am a Principal Mechanical Engineer, an assistant to the Chief of the Mechanical Engineering Branch (MEB) for performing independent review of generic matters and coordinating technical position among the staff.
For nine years in MEB, I have reviewed plant design criteria, plant operating problems, plant safety considerations and dynamic analysis and testing of piping, equipaent, reactor internals and nuclear safety features.
I also served as the leader of the Seismic Qualification Review Team for conducting plant seismic audit, and the Task Manager for developing staff position on plant safety against the postulated pipe rupture event.
In addition, I am on several National Standard Committees and participated in several Regulatory Guide developments.
Born 1934 in China, I ccme to USA in 1957.
I received the B.S. in Civil Engineering from Taiwan University in 1955, the M.S. in Structural Dynamics from Virginia Tech. in 1956, and the Ph.D. in Structural Mechanics from M.I.T.
in 1968.
For 26 years after the B.S., I have had various work %J experiences in structural design, stress analysis, res a rch and development in space vehicle dynamics, and technical review ir. nuclear power plant safety.
I was a visiting lecturer to universities in England (1971), and Chile (1975), and to tae government in Taiwa.
China (1975).
I was the author of a dozen technical papers, a recipient of Ap,llo Achievement Award froia NASA (1969) and a High Quality Performance Increa;e from AEC (1975), and a member of Sigma X', Tau Beta Pi, Chi Epsile', AIAA and ANS.
During 1955-57 I served as a commissioned engineering liaison officer with the National Chinese Navy in various US-aid military projects, and passed the Examination of Professional Engineers.
In 1957 I came to the USA with teaching assistantship from VPI where I completed the M.5. degree in one year.
The thesis was entitled " Vibration Behavior of l
Parabolic Arches." I received "VPI Structural Ten" award after graduation.
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During 1958-64 I was a Bridge Design Engineer with Virginia Highway Department.
l For six years, I worked in stress analysis of steel, reinforced concrete and l
prestressed concrete structures.
In 1960, after taking short courses with IBM, I was assigned to perform ir, dependent studies in developing computer capability in design and analysis.
In 1963 I published the solution manual for "The Mechanics of Solid" authored by G. L. Rogers.
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In 1964 I received a Research Assistantship from MIT where I completed the Ph.D. degree in 1968.
My studies were emphasized on mechanical vibrations, t
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2-material behavior under various loading and temperature conditions, and stochastic processes in engineering applications.
Based on my research in random vibration theories, I completed the doctoral thesis entitled " Earthquake Simulation Models and Their Applications."
In MIT I was elected to honor societies of Sigma Xi in 1965, Tau Beta Pi in 1966, and Chi Epsilon in 1967.
During 1968-71 I worked as a member of the technical staff with the Space Vehicle Dynamics Department of Bellcomm, Inc., which was performing technical studies, system planning and analysis in the Office of Manned Space Flight within NASA headquarters, Washington, D.C.
My works were related to investigation, evaluation and development of dynamic analysis and testing technology in space vehicle and missile dynamics, as well as participating in task groups for solving problems such as P0GO, rover stability, wind simulations, and fuel tank sloshing etc.
I also had full responsibility for developing computer capability to perform large system dynamic analysis, such as modal synthesis.
I received " Apollo Achievement Award" by NASA in 1970 and
" Recognition of Accomplishment" by AT&T in 1971.
I was a visiting lecturer to England in 1971 and published ten technical papers in this period of time.
In January 1972 I joined the U.S. Atomic Energy Commission and have remained with this organization through the transition to the U.S. Nuclear Regulatory Commission.
During this time I have participated in the review of plant operating problems and design criteria and evaluation of over forty construction permits and operating license applications in the area of dynamic effects of LOCA, earthquakes, pipe rupture, and operating transients on systems, components, equipment, reactor internals and nuclear safety instrumentation.
I had served as the leader of SQRT (Seismic Qualification Review Team) for condu: ting plant seismic audit, and Task Manager of a generic program for investigating design criteria for plant protection against postulated high energy line rupture.
I am a roember of industry codc and standards writing bodies including:
ANSI N176,
" Design Basis for Protection of Nuclear Power Plants Against Effects of Postulated Pipe Rupture," and Standard IEEE-344 "Recommanded Practices for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations." During my service in NRC, I have received one "High Quality performance Step Increase" and published two technical papers.
In 1975 I was invited by the University in Chile to give lectures regarding seismic design of nuclear power plants and by the National Chinese government to discuss various subjects concerning nuclear power plant safety.
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