ML20050Q739
| ML20050Q739 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 04/12/1982 |
| From: | Kirkwood R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20050N853 | List: |
| References | |
| NUDOCS 8204140485 | |
| Download: ML20050Q739 (7) | |
Text
,1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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LONG ISLAND LIGHTING COMPANY
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Docket Number 50-322
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(Shorehan Nuclear Power Station,
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Unit 1)
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NRC STAFF TESTIMONY OF ROBERT KIRKWOOD REGARDING PASSIVE MECHANICAL VALVE FAILURES (SUFFOLK COUNTY CONTENTION 11) l 6
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OUTLINE OF TESTIMONY l
Intervenor contends that Applicant has not demonstrated that valves used in the safety-related systems will not fail in an undetectable or unsafe mode, thereby jeopardizing the safe operation of Shoreham and i
violating 10 C.F.R. 50, Appendix A, GDC 23, 34, 35, 37 and 40. This allegation is incorrect. The Applicant has designed, fabricated, erected, and tested the valves used in the safety-related systems at Shoreham in accordance with applicable codes and standards, to required environmental qualification criteria, and appropriate design and service loads.
During plant operation the valves will also be tested in I
accordance with an Inservice Testing Progran which also conforms to code requirements.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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LONG ISLAND LIGHTING COMPANY Docket Number 50-32?
(ShorehamNuclearPowerStation
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Unit 1)
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NRC STAFF TESTIMONY OF ROBERT KIRKWOOD ON SOC CONTENTION 11 Q.
Please state your name and position with the NRC.
A.
My name is Robert Kirkwood.
I am employed by the U.S. Nuclear Regulatory Comission as a Principal Mechanical Engineer assigned to the Mechanical Engineering Branch in the Division of Engineering. A copy of my professional qualifications is attached.
Q.
What is the purpose of your testimony?
A.
The purpose of this testimony is to respend to Suffolk County Contention 11: Passive Mechanical Valve Failures which states:
l Suffolk County contends that LILC0 has not demonstrated that the valves used in the safety-related systems at Shoreham will not fail in an undetectable or unsafe mode, thereby jeopardizing the safe operation of Shoreham and violating 10 C.F.R. 50, Appendix A, GDC 23, 34, 35, 37 and 40.
Specifically, my testimony addresses; (1) the adequacy of the valves in safety-related systems for Shoreham, and (2) the means to assure l
operational readiness of valves and provide assurance that these valves l
are in accordance with 10 C.F.R. 50, Appendix A.
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Q.
What codes and standards are used in the construction of valves in safety-related systems for Shoreham?
l A.
The valves in safety-related systems for Shoreham are designed, fabricated, erected, and tested in accordance with those codes and standards that were required by the NRC at the time of the procurement of these components. Specifically, valves in the safety-related systems for Shoreham were constructed to the codes and standards in effect on the purchase order date of the components. These codes and standards are:
1.
ANSI B 31.1.0, Power Piping 2.
Draf t AStiE Code for Punos and Valves for Nuclear Power, Class I, Class II, or Class III.
3.
ASf1E Section III, Class 1, 2, or 3 The codes and standards used in the construction of the valves in safety-related systens are identified in FSAR Tables 3.2.1-1 and 3.2.1-2.
Q.
What additional means are available to assure that valves constructed to approved standards meet the required specifications for structural integrity and function?
A.
Prior to plant operation, based on SRP 3.10 and SRP 3.11, the NRC Staff audits the applicant's equiptrent qualification files and reviews the results of tests and analyses performed by the applicant to assure the proper implementation of criteria and that qualification has been adequately demonstrated for all valves. The staff verifies that all design and service loads including the loadings inposed by the safe shutdown earthquake, postulated accidents, and loss-of-coolant accidents have been
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properly defined and accounted for in the test and analyses performed.
The NRC staff also verifies that the valves meet the environmental qualification requirements and are capable of performing their design
e j safety functions under all normal, abnormal, accident, and postaccident s
environments and for the length of time during which the valve function is required.
Q.
What means are used to ensure that safety-related valves will be in a state of operational readiness to perform necessary safety functions throughout the life of the plant.
4 A.
During plant operation certain valves in safety-related systems will be tested for operational readiness in accordance with the requirements of Subsection IWV, " Inservice Testing of Valves in Nuclear Pcwer Plants", of the ASME Boiler and Pressure Vessel Code Section XI,
" Rules for Inservice Inspection of Nuclear Power Plant Components", as required by 10 C.F.R. 50 Section 50.55a(g)," Codes and Standards". This program includes baseline preservice testing and periodic inservice testing. The program provides for both functional testing of the valves in the operational state and for visual inspection for leaks and other 1
signs of distress.
Q.
What are your conclusions concerning this contention?
I A.
Contrary to Intervenor's allegations, the valves are designed, I
i fabricated, erected, and tested in accordance with the applicable codes and standards and will be tested during plant operation in accordance with an approved inservice testing program. Therefore, the Intervenors allegations are without merit.
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9 Robert Kirkwood Professional Qualifications Mechanical Engineering Branch Division of Engineering Office of Nuclear Reactor Regulation I am a Principal Mechanical. Engineer in the Mechanical Engineering Branch responsible for the review and evaluation for compliance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50, pertaining to the codes and standards under which pressure-retaining components of the reactor coolant pressure boundary are designed, fabricated, erected, and tested.
Included in this review and evaluation is the classification system pertaining to the codes and standards under which pressure-retaining components of other fluid systems important to safety of a nuclear power plant are designed, fabricated, erected, and tested.
I am responsible for the revies and evaluation pertaining to the identification of structures, fluid systems and components and other mechanical components that should be designed to withstand the effects of the Safe Shutdown Earthquake (seismic Category I) and within the scope of a Quality Assurance a
Program that is in compliance with 10 CFR 50, Appendix B.
I graduated from Belfast College of Technology with a B.S. degree in Mechanical Engineering.
I am a registered professional engineer in the State of California.
From January 1949 to March 1951, I was employed by Monsanto Chemical Company, Montreal, Quebec, Canada. During this period, I was engaged in chemical plant layout and associated design of pressure vessels, storage tanks, piping systems and mechanical equipment.
From March 1951 to September 1956 I was employed by Atomic Energy of Canada Limited, Chalk River, Ontario, Canada.
During this period, I was engaged in the design of fuel elements, fuel handling equipment, remote handling equip-ment, experimental loops, pressure vessels and nuclear piping systems.
From October 1956 to November 1959, I was employed by General Dynamics /Convair Scientific Research Laboratory, San Diego, California. During this period, I was responsible for the design, fabrication and installation of laboratory equipment.
From November 1959 to July 1968, I was employed by Gulf General Atomic, San Diego, California. During this period, I was the engineer in charge of a design group in the High Temperature Gas-Cooled Reactor Division, responsible for nuclear power plant arrangement. designs including detailed piping layouts for the nuclear steam supply system. Other areas of responsibility included conceptual and preliminary design of reactor concepts utilizing prestressed concrete reactor vessels, reactor vessel internals, and associated design of fuel handling equipment.
In August 1968, I started work for the U.S. Atomic Energy Commission as a specialist in the Engineering and Component Branch, Division of Reactor Standards, where I. participated in the development of reactor standards, codes and criteria.
In October 1971, I was reassigned to the Mechanical Engineering Branch, Division t
of Reactor Standards, where I participated in the development of ASME nuclear codes and standards.
In April 1972, I was reassigned to the Reactor Systems Branch, Division of Reactor Standards, where I participated in ANS standards development and C.P. review of the applicability of codes and standards and seismic classification of structures, systems, and components of light-water-cooled and gas-cooled nuclear power plants.
I remained through the transition into the U.S. Nuclear Regulatory Commission.
In December 1975, I was reassigned to the A/D for Plant Systems with duties similar to the above.
In January 1977, I was reassigned to the Auxiliary Systems Branch, where I participated in ANS standards development and C.P. and 0.L. review of the applicability of codes and standards and seismic classification of structures, systems, and components of nuclear power plants.
In September 1980, I was reassigned to the Mechanical Engineering Branch with duties similar to the above.
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