ML20008F973
| ML20008F973 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 05/11/1981 |
| From: | Fields M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20008F949 | List: |
| References | |
| NUDOCS 8105120417 | |
| Download: ML20008F973 (4) | |
Text
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05/11/81 O
UNITED STATES OF AMERICA NUCLEAR REGULATORY C0:4 MISSION BEFORE THE AT0i4IC SAFETY AND LICENSING BOARD In the 14atter of
)
HOUSTON LIGHTING AND POWER C0l1PANY Docket No. 50-466
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(Allens Creek Nuclear Generating
)
Station, Unit 1)
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HRC STAFF SUPPLEMENTAL TESTIM 0iiY OF MEL B. FIELDS REGARDING CONTAINMEi4T QUESTIONS --
C011PLIANCE WITH GDC 50 (ASLB 6) AND BYPASS LEAKAGE -(ASLB 9)
Q.
Please state your name and position with the NRC.
A.
My name is Mel B. Fields.
I an a Systens Engineer in the Containment Systems Branch of the Office of Nuclear Reactor Regulation.
A copy of my professional qualifications statement is attached.
n.
What is the purpose of this testinony?
A.
The purpose of this testimony is to respond to two Licensing Board questions regarding containaent systems. The first question (ASLB 6) asks whether the proposed Allens Creek facility will neet the current requirements of the Commission with respect to General Design Criteria (GDC) 50 -- Containment Design Basis.
The second question (ASLB 9) asks the Staff:
What is the technical basis for concluding that maintaining containment atmospheric temperature and relative humidity values within prescribed limits is a practical method for minimizing bypass leakage?
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I will respond to each of these questions separately below.
ASLB 6 Q.
What are the NRC's current requirenents'for containaent desigq basis?
l A.
The current requirements are set forth' in _GDC 50, Appendix A to 10 C.F.R. Part 50.
These requirements were established in 1978. See 43 Fed. Reg. 50162 (October 27,1978).
Q.
Does Allens Creek neet these GDC 50 requirenents?
A.
Yes. The spectrua of accidents that the Applicant was required to analyze include a variety of break sizes and single active failures.
The most conservative assumption regarding degraded emergency core cooling is loss of offsite power and loss of one of the two onsite emergency power supplies. This assumption, which has been analyzed by the Applicant, is more conservative than loss of one ECCS line due to a break in that line because one entire ECCS train is lost when only one ensite power supply is available. Also, for Mark III containments, neither the short-tera nor the long-tern pressure and temperature response of the drywell and containnent is affected by degraded emergency core cooling.
This is because the short-term response is controlled by the first fed seconds of the blowdown and the long-tera response is controlled by the i
amount of energy that can be removed from the suppression pool by one train of the RHR heat renoval s,,Len.
ASLB 9 Q.
The Board has questioned whether maintaining containment atnosphere temperature and relative hunidity values within prescribed l
p 3-1 limits is a practical method for ninimizing bypass' leakage.
Is the Staff-requirement of maintaining these values within prescribed'linits-associated with the bypass leakage'between.the drysell and containment?
Q.
No.
Containment atmosphere relative hunidity and tenperature limits are not used to control bypass leakage, but arel concerned with the sizing analysis for the containment vacuum breaker. system. - The Staff's 3
basis for this sizing analysis is set forth in Section 6.2.1(2) of SER
- Supplement No. 2 (HUREG-0515).
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t Professional Qualifications Mel B. Fields I an a Systens Engineer in'the Containment'Systeas' Branch of the Office of Huclear Reactor Regulation.
In this position I am responsible for the revicw and technical evaluation of safety aspects of containment systems.
I graduated fron the University of Arizona with a Bachelor of Science Degree in Nuclear Engineering in 1974 and received a MS in Mechanical Engineering fron Catholic University of America in Washington,_
D.C. in 1981.
In 1975 I accepted a position as a Reactor Engineer in the Containment Systens Branch, Division of Systens Safety, Nuclear Regulatory Connission. My responsibilities included the review and technical evaluatien of the safety aspects of containment systens.
In this position, I have been responsible for the evauation of the health
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and safety aspects related to containment systens for the following nuclear power plants:
Black Fox Station, Units Nos. 1 & 2, Grand Gulf Nuclear Station, Units Hos.1-A 2, North Anna Power Station, Units Hos.1
& 2. Jamesport Huclear Station, Units Hos.1 & 2 and Cherokee and Perkins Nuclear Station, Units Not 1, 2 & 3.
For the Black Fox Station, I was responsible for reviewing the staff positions and writing the section of the Safety Evaluation Report on the Mark III containment system.
In early 1977, I was transferred to another branch, the Power Systens Branch, in the same division where I remained for approximately 1-1/2 years before returning recently to the Containment Systens Branch.
I an currently involved in the reviet of the liark III Containment Test Progran being conducted by General Electric.
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