ML20010C337
| ML20010C337 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 08/13/1981 |
| From: | Hodges M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20010C329 | List: |
| References | |
| NUDOCS 8108190363 | |
| Download: ML20010C337 (6) | |
Text
k UNITED STATtS OF AMERICA NUCLEAR REGULATORY 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 (Allens Creek Nuclear Generating
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Station, Unit 1)
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HRC STAFF TESTIMONY OF MARVIN W. (WAYNE) H0DGES REGARDING POSITION INDICATION FOR SRV (D0HERTY CONTENTION 42)
Q.
Please state your name and position with the NRC.
A.
My name is Marvin H. (Wayne) Hodges.
I am employed by the U.S.
Nuclear Regulatory Commisison as a Section Leader in the Reactor Systems Branch of the Division of Systems Integu tion. A copy of n1y professional qualifications is attached.
Q.
What is the purpose of your testimony?
A.
The purpose of this testimony is to respond to Doherty Contention 42, which reads as follows:
Intervenor contends his health and safety interest will be injured because the infomation system giving the position of power operated relief valves and safety valves to the reactor operators is ambigucus and in need of improve This is a finding of the " Lessons Learned Task Force" ment.
on Three Mile Island, Unit 2 (TMI-2), reported in NUREG-05/8, on p. 7, and is further supported by the fact Applicant states it "... will comply with the recommendations te alleviate this problem (Letter from E. A. Turner, Applicant to H.
Denton, NRC, August 9, 1979, Attachment A, p. 2).
- However, Applicant does not say how this will be done, nor is there evidence this is even possible. Hence, this intervenor contends Applicant must show how the recommendation will be complied with at the cor.struction license hearing.
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Q.
Is the position indication for safety / relief valves at ACNGS ambiguous and in need of improvement?
A.
No.
In Amendment 57 to the Allens Cred PSAR, the applicant has conunitted to a system in which safety / relief valve position indication will be determined by pressure sensors in the discharge pipe which are redundant, safety grade, seismically and environaentally qualified, and powered from Class 1E power sources. An alarm indicating that a safety /
relief valve is open will be provided in the control room.
This system will provide an unambiguous indication of valve position.
Q.
How does this valve position indication differ from that proposed for Allens Creek and used in other reactors prior to the accident at Three Mile Island?
A.
Prior to that accident, most, if not all, reactors used thermocouples to indicate flow through relief and safety valves. At THI-2, both thermocouples and an indication of an open or clored signal to the power operated relief valves were used.
Neither of these two signals is unambiguous. The open/close signal indicates what the valve was commanded to do, not what it did. The thermocouple can heat up due to a leaking valve or remain hot for a period of time after a vah has closed, thus providing the operator with a false position indication.
As a result of the lessons learned from TMI-2, hil reactors are now required to have a direct, unambiguous indication of valve position for safety and relief valves. The pressure sensor coricept propcsed for ACNGS has been revie..ed and approved by the NRC Staff for the Grand Gulf ficility.
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Q.
Why is the system to be used at Allens Creek unambiguous?
A.. When the safety / relief valve is closed, the pipe pressure should be near the containment pressure.
However, as soon as the valve is opened, an almost instantaneous pressure rise would be detected by the sensor and would trigger the alarm.
For the reasons noted in the previous answer, the thermocouple systems used in the past could give an indication that the valve was open when, in fact, it was closed.
J.
Does the pressure sensor to be used at Allens Creek provide a i
direct indication of valve position?
A.
To respond to this question with precision, the pressure sensor provides a direct indication of flow through the valve. However, since flow is the parameter of interest, the design is acceptable to the Staff, and as noted above, a similar design has been approved by us previously for the Grand Gulf facility.
Q.
How sensitive will the pressure sensors be with regard to detecting flow through safety / relief saives?
i A.
Typical pressure sensors are calibrated to read full fluw, i.e.
safety / relief valve fully ope 9 and reactor pressure in the normal operating range, and will not be very sensitive to low flows (flow less than approximately five percent of full flow).
Small leakages may, i
therefore, not be indicated by the sensor.
However, this small leakage is not significant from a safety standpoint. There is other instrumentation installed in the power plant which would allow the operators to deduce leakages above typical tech spec limits, i.e. Sgpm.
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4-Q.
Is there any concern over the feasibility of the position indication system to be used at Allens Creek?
A.
No. The pressure increase in the pipe running from the safety / relief valve to the suppression pool will provide an easily measured and definite signal to the operators, alerting them io the status of the
-safety / relief valve. This type of pressure sensing device has had substantial successful operational use in both nuclear and non-nuclear facilities for many years.
Q.
What is your conclusion en regarding this contention?
A.
The system to be used at Allens Creek represents a significant f
improvement in the valve position indication system previously used in operating reactors.
It will provide a direct indication of flow througn the valve and thus an una$aiguous indication of valve position to the operators. Further, there is no question that the system is feasible a3 demonstrated by the use of similar systems in different types of industrial facilities for many years.
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Marvin W. (Wayne) Hodges Professional Qualifications
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Reactor Systems Branch Division of Systems Integration U. S. Nuclear Regulatory Commission I an employed as a Section Leader in Section B of the Reactor Systems Branch, Division of Systems Intrgration.
I graduated from Auburn University with a BS degree in Mechanical Engineering I received a Master of Science degree in Mechanical Engineering from in 1955.
Auburn University in 1967..
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In my present work assignment at the NRC, I serve as principal reviewer in e s the area of boiling water reactor systems.
I also participate in the review of analytical mocels use in the licensing evaluations of boiling water reactors and I have the technical review responsibility for many of the modifications and analyses being implemented on boiling wa.ter reactors after the Three-Mile Island, Unit-2 accident.
As 6 nember of the Bulletin and Orders Task Forca which was formed af ter the TMI-2 accident, I was responsibic for the review of the capability of BWR systems to cope with loss of feedwater transients and small break loss-of-cool-ant acciderts.
I have also served at the NRC as a reviewer in the Analysis Branch of the NRC I served as in the area of thermal-hydraulic performance of the reactor core.
a consultant to the RES representative to the program management group for the BWR Blowdown /Er..argency Core Cooling Program.
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' Prior to joining the NRC staf f in Mirch,1974, I was employed by E.1. DuPont At SRL, I conducted at the Savannah River Laboratory as a research engineer.
hydraulic and heat transfer testing to support operation of the reactors at I also performed safety limit calculations and the Savannah River Plant.
participated in the development of analytical models for use in transient analyses at Savannah River. My tenure at SRL was from June 1967 to March 1974.
From September 1965 to June 1967, while in graduate school, I taught courses 4
in thermodynamics, statics, mechanical engineering measurements, computer During programming and assisted in a course in the history of engineering.
the summer of 1966, I worked at the Savannah River Laboratory performing hydraulic testing.
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