ML19294C230
| ML19294C230 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/29/1980 |
| From: | Meyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19294C229 | List: |
| References | |
| NUDOCS 8003070341 | |
| Download: ML19294C230 (11) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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SACRAMENTO MUNICIPAL UTILITY
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Docket No. 50-312 (SP)
DISTRICT
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(Rancho Seco Nuclear Generating
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Station)
NRC STAFF TESTIMONY OF DR. JAMES F. MEYER ON CONTAINMENT OVEP, PRESSURIZATION PROTECTION (CEC Issue 5-2)
Question 1 Please state your name and your position with the NRC.
Response
My name is Dr. James F. Meyer.
I an a Reactor Engineer in the Advanced Reactor Branch, Division of Project Management, Office of Nuclear Reactor Regulation.
Question 2 Is a statement of your professional qualifications attached to this testimony?
Response
Yes, a copy of this statement is attached to this testimony.
Qu stion 3 What contention is addressed in your testimony?
Response
My testimony is to partially respond to California Energy Commission (CEC)
Issue 5-2, which reads:
80030703c_._
2 CEC Issue 5-2 Whether the containment building should be modified to provide overpressurization protection with a controlled filtered venting system to mitigate unavoidable release of radionuclides?
Question 4 What is meant by controlled filtered venting?
Response
Controlled filtered venting is a process in which a portion of the containment atmosphere is deliberately released to the environment in a cor. trolled manner thrcugh a system of filters and energy absorbers.
Such a pressure relief system would be actuated to reduce containment presure, a pressure which could otherwise fail the containment and thereby allow the unc'ontrolled and unfiltered release of radiation into the atmosphere.
This system would only be used to mitigate the effects of a severe accident beyond the design basis accident.
Depending on the characteristics of specific designs, different radioactive isotope attentuation factors could be realized for these filtering systems.
For example, attenuation factors can vary from 50% to nearly 100% for organic iodine.
For all designs, the attenuation factors for particulates and molecular iodine are better than 98%.
Whatever the fina'l choice, the Filtered Vented Contaiment Systen (FVCS) will result in a considerable reduction in societal risk relative to an uncontrolled, unfiltered containment failure.
Question 5 For what level of contaiment pressure would controlled filtered venting be used?
Response
The pressure level would depend on the specific set of dominant-risk-contributor accident sequences for a given nuclear power plant.
The " dominant-risk-contributor accident sequences" are those several accident sequences which, when using quantitative risk analysis, are considered the major contributors to the societal risk from a given nuclear power plant.
For example, for the PWR reactor used in the Reactor Safety Study (Reference 1), a major contributor was a trans.
with loss of all on-site and off-site AC power.
From these sequences, containment pressure histories can be calculated from which judgements can be made regarding appropriate pressure levels to initiate venting.
Basically, the levels cannot be too high so as to vent too late for the venting system to accommodate the pressure surge, nor too low so as to vent too early, possibly then unnecessarily A pressure presently being considered for a passive w nt system is about 60 psia, or approximately the design basis pressure of a large-volume dry containment.
Question 6 Does the containment basis accident include consideratien of core degradation or core melt?
Response
In terms of the definition of the containment design basis accident (C3A), it does not since the DBA is defined - and the Emergency Core Cooling System (ECCS) and Engineered Safety Features (ESF) systems are designed - such that there is
_4 only minor damage to the core with no fuel melting.
However, the same safety systems which are included in the plant design to accommodate (with virtually no core damage) the DBA, also have the capacity to accommodate certain accidents beyond the DBA.
This is because these safety systems are designed conservatively.
Examples of safety systems which are conservatively designed and would help in mitigating accidents beyond the DBA are:
DBA 10 CFR 50 DESIGN ACTUAL REFERENCE Containment Pressure Part 50, App A, Accommodation 59 psig 118 osig Crit. 50 Contaiment Steam Energy Part 50, App A, 6
6 Accommodation (Sprays +
240 x 10 480 x 10 Crit. 38 Coolers)
BTU /HR BTU /HR The nature of the conservativisms is as follows.
The DBA must be accommodated by redundant systems, both of which have the full capability to do the job intended. Thus, unless a system fails, both would be available for accommodating an accident and thus be able to remove considerably more heat energy from the containment than that produced in a DBA.
Thus, in the sense of the margin designed into the safety syster.is, certain accidents beyond the DBA can be accommodated.
Question 7 What is the status 'f the Filtered Vented Containment System (FVCS) studies to date?
Response
Various types of FVCSs have been installed or are being installed in Fast Breeder Reactor facilities both here and abroad.
For example, the Zero-Power Plutonium Reactor (ZPPR) test facility (Reference 2), the Fast Flux Test
. Facility (FFTF), and the German SNR-300 prototype LMFBR (Reference 3) all have FVCSs, or are installing them.
Vent-filter systems for LWRs have received attention since 1975, when Norwegian and Swedish studies on underground siting considered the use of the surrounding soil and rock as a filtering medium (Reference 4).
Subsequently, a UCLA study group presented a conceptual design of a vent-filter system for LWRs (Reference 5) comprised of a graded sand and gravel b:=d with downstream HEPA and charcoal filters. Their design included the use of Lydrogen burners to minimize the likelihood of hydrogen explosions and air cooling fans to prevent overheating of the charcoal filters.
More recently, the use of a controlled vent-filtered system for core melt accidents was considered in a conceptual study of underground nuclear plants for the California Energy Commission (CEC) (References 6, 7).
The CEC design was completely passive, with the principal filtering structure being an underground pressure relief volume filled with crushed rock and gravel.
Although most of the emphasis in the United States has been focused on atmospheric venting, the zero-release concept of venting directly into a separate vacuum containment building has been incorporated into some of the Canadian multi-unit CANDU reactors (Reference 8).
The basic conclusions of the studies to date are two-fold.
First, systems can be designed and implemented which can vent large volumes of cases and vapors in a controlled manner and which can attenuate (absorb) virtually any radioactive isotope known to be harmful.
Basically the technology is in place to do the job required although it should be pointed out th'.c some of the more sophisticated systems are very expensive.
The second conclusion is that, although these systems could contribute significantly to the reduction in societal risk from a nuclear power plant, much more work needs tc be done before all the negative as well as positive aspects of a FVCS can be factored into n integrated risk reduction assessment.
Open questions include possible interference with other engineered safety features, possible exacerbation of low-consequence accidents into high-consequence accidents, possible increase of hydrogen explosion potential, impact of uncertainties in various phenomenological and cost evaluation areas, and reconcilation of vent-filter systems with the current regulatory position requiring essentially leaktight containment.
Question 8 Does the NRC have a program to address the safety, licensing and value impact of FVCS?
Response
The United States Congress, in the Fiscal Year 1978 Budget Authorization Act, directed the Nuclear Regulatory Commission to prepare a plan for the development of new or improved safety systems for nuclear power plants.
In April 1978 the NRC submitted such a plan to Congress, outlining seven key areas of research to be conducted over 3 years at a total estimated cost of
$14.9 million.
Of the various research projects proposed by the NRC, a program for the develop-ment and analysis of FVCS conceptual designs was accorded particularly high priority by the NRC and the Advisory Committee on Reactor Safeguards (ACRS).
. Funding of this program was subsequently approved.
This program combines risk reducticri analysis for specific plants with actual conceptual design analyses being performed at Sandia Laboratories (Reference ").
Concurrent with and complementary to the above program, NRC Staff has initiated, as part of its TMI Action Plan (NUREG-0660, currently in Draft), specific licensing and safety evaluation studies centering on FVCSs. Central to this licensing effort, is a planned rulemaking on mitigation features for severe accidents including the FVCS.
Question 9 What is the NRC Staff's position on requiring FVCS? In particular, should the Rancho Seco containment building be modified to provide overpressurization protection with an FVCS?
Response
It is the NRC Staff's position that a nuclear power plant which conforms to all the licensing requirements, criteria, and regulations presently in place is sufficiently safe to operate.
That is, the health and safety of the public is sufficiently assured. As discussed in the NRC Staff Testimony of Thomas A.
Greene on Containment Overpressurization Protection, Rancho Seco is such a plant.
Since the TMI-2 accident, attention has been drawn to severe accidents, including core melting and consideration has been given to additional efforts to reduce the probabilities of their occurrence and also to mitigate their consequences.
At TMI-2, large amounts of radioactive material were released into the containment.
The containment retained its integrity and very little radioactive material was released to the atmosphere.
Consideration of mitigating the consequences of severe accidents arises from the potential for containment failure while
. large amounts of radioactive material are present in it.
In this regard, the Staff has proposed to the Commission (NUREG-0660, Task II.8) that safety reviews be extended to include consideration of degraded or melted cores.
The Staff is recommending that this issue be the subject of rulemaking.
The rulemaking would include,among other items, consideration of the use of the filtered vented containment system (FVCS) to mitigate the potential consequences of core degradation and core melt accidents.
The Commission has not yet acted on the Staff's proposal.
REFERENCES 1.
Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, WASH-1400, USNRC, 10/75.
2.
H. Lawroski, et al., Final Safety Analysis Report on the Zero Power Plutonium Reactor (ZPPR Facility, ANL-7471 (Argonne, 11: Argonne National Laboratory, June 1972).
3.
L. Bohm, S. Jordan, and W. Schikarski, "The Off Gas Filter System of the SNR-300," Pror.13th USAEC Air Cleaning Conference, San Francisco, Ca, August 1974, pp. 620-031.
4.
Rock Siting of Nuclear Power Plants from a Reactor Safety Standpoint, Final Report (Sweden: Centrala Draftledningen, November 1975).
5.
B. Gossett, et al., Post-Accident Filtration as a Means of Improving Containment Effectiveness, UCLA-ENG-7775, December 1977.
6.
F. C. Finlayson, et al., Evaluation of the Feasibility, Economic Impact, and Effectiveness of Underground Nuclear Power Plants, Final Technical Report, ATR-78(7652-14)-1 (Aerospace Corporation, May 1978).
7.
J.E. Ward, et al., Conceptual Design and Estimated Cost of Buried
' Berm Contained' Nuclear Power Plants, Contract No. 154-064 (S&L Engineers, January 1978).
8.
E. W. Fee and G. E. Shaw, " Vacuum Containment Systems for Multi-Unit Nuclear Power Stations," 7th Int. Congress on the Confinement of Radioactivity in the Utilization of Nuclear Energy, Societe crancaise de la Radioprotection, Versailles, France, May 1974.
9.
A. S. Benjamin, " Program Plan for the Investigation of Vent-Filtered Containment Conceptual Designs for Light Water Reactors," NUREG/CR-1029, Sandia Laboratories (October 1979).
PROFESSIONAL QUALIFICATIONS JAMES F. MEYER ADVANCED REACTORS BRANCH DIVISION OF PROJECT MANAGEMENT My name is James F. Meyer.
I am employed by the U.S. Nuclear Regulatory Commission, Washington, D. C. 20555.
I am a Reactor Engineer and, as such, am responsible for analyzing and evaluating technical input for licensing actions in the general areas of severe (core degradation, melt or disruption) accidents in nuclear power plants.
I attended Valparaiso Univer..cy, Valparaiso, Indiana from 1958 to 1963, where I received Bachelor degrees in Electrical Engineering and Physics.
Upon com-pletion of my undergraduate studies in 1963, I enrolled in the Nuclear Engineering Department at the Pennsylvania State University.
In 1965, I received by M.S.
degree and in 1968 my PhD, both in the subject area of nuclear engineering.
Following graduate studies, I worked for Argonne National Laboratory for about six years.
During the first five years, I worked in the Applied Physics Division at ANL on Development, planning, execution, analysis, and reporting of experiments on Zero Power (Plutonium) Reactors. My experience was in reactor analysis, fast reactor experiments, and general engineering (design and development).
In addition to completing the abcVe tasks, my responsibilities included being a reactor supervisor on two plutonium test reactors.
From October 1973 to November 1974 I was on loan from ANL to the Atomic Energy Comission (now NRC) working in the Liquid Metal Fast Breeder Reactor Branch.
For about 4 years (including the one year I was on loan from ANL), my duties included analysis, assessment, and evaluation of safety / licensing issues associated with the Clinch River Breeder Reactor with specific responsibility in the areas of fuels, reactor physics, accident analysis, and core disruption nalysis. Accomplishments durino this period included publishing reports, establishing licensing criteria and preparation for licensing hearings.
From about August 1977 to August 1978, I had similar responsibilities for the Fast Flux Test Facility (FFTF) culminating in contributions to the Safety Analysis Report for FFTF.
The NRC, especially during the past year, has been participating in the Administration's Non-Proliferation Alternative Systems Assessment Program and the International Nuclear Fuel Cycle Evaluation Program.
I have had lec.d responsibility, representing the Of; ice of Nuclear Reactor Regualtion, in conducting an independent comparative evaluation of the safety, environmental, safeguards, and licensing issues for the advanced reactors under consideration.
Our preliminary assessment has been sent to the Department of Energy and a Report to Congress was prepared.
Because of my experience in the area of severe (core melt) accidents for advanced reactors, I was given similar parallel responsibilities, during the sumer of 1979, in the area of PWR and BWR severe accidents and severe accident mitigation features.
These responsibilities include analysis and evaluation of Filtered Vented Cor,tainment Systems, Hydrogen Control Systems,
2 and Core Retention Devices.
Particular licensing applications include the Zion / Indian Point Task Force consideration of mitigating features for these power plants and the scheduled Rulemaking on consideration of dearaded or melted cores in safety reviews.
While working in all of the above three areas, I have been responsible for managing a large ($1 million) technical assistance program at various laboratories and universities. Also, I have made numerous presentations before the Advisory Committee on Reactor Safequards.
During this time period in the evenings I have taaght reactor physics courses at the University of Maryland. As a " Visiting Lecturer" I taught three 3-credit graduate level courses in the Fall of 1976 and 1977 and the Sprina of 1978.
My honors include Sigma XI, a number of Scholarships and Fellowship awards, and a "High Quality Certificate" presented in July 1978.
I am the author or co-author of several papers involving reactor physics and reactor safety, and I am a member of the American Nuclear Society.