ML19294C228
| ML19294C228 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/29/1980 |
| From: | Thomas Greene Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19294C229 | List: |
| References | |
| NUDOCS 8003070340 | |
| Download: ML19294C228 (9) | |
Text
'.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
SACRAMENTO MUNICIPAL UTILITY
)
Docket No. 50-312 (SP)
DISTRICT
)
)
(Rancho Seco Nuclear Generating
)
Station)
)
NRC STAFF TESTIMONY OF THOMAS A. GREENE ON CONTAINMENT OVERPRESSURIZATION PROTECTION (CEC Issue 5-2)
Question 1 Please state your name and your position with the NRC.
Response
My name is Thomas A. Greene.
I am a Senior Systems Engineer in the Containment Systems Branch, Division of Systems Safety, 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.
Question 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:
80030703 yg
_2 CEC Issue 5-2 Whether the containment building should be modified to provide over-pressurization protection with a controlled filtered venting system to mitigate unavoidable release of radionuclides?
Question 4 Does the Rancho Seco facility have an overpressurization protection system for the containment building?
Response
Yes, the Rancho Seco facility has an overpressurization protection system called the heat removal system.
It corisists of the reactor building spray system (RBSS) and the reactor building emergency cooling system (RBECS),
which remove energy from the containment atmosphere following an accident and prevent the containment from being overpressurized.
The RBSS serves only as an engineered safety feature and performs no normal operating function.
The system, which consists of two separate spray trains of equal capacity, is a seismic Category I system.
All active components of the RBSS are located outside the reactor building.
The RBECS consists of four fan-cooler units and four emergency upper dome air circulators.
Two of the fan-cooler units are equipped with activated charroal filters.
The fan-cooler units will perform no normal operating function, whereas the emergency air circulators may be used during normal plant operation to improve air circulation within the reactor building. The RBECS can be operated re-motely from the control room and is a seismic Category I system.
. Question 5 What was the design basis acclaent against which the overpressurization protection system was designed?
Resoonse The containment building for Rancho Seco is a steel lined, reinforced concrete structure with a net free volume of 1.98 million cubic feet and is designed for a pressure of 59 pounds per square inch gauge (psig).
The design pressure of the containment building is based on oostulated reactor coolant system pipe break accidents that were analyzed in the followins manner.
Mass and energy release rates to the containment during the blowdown and core reflood phase of a postulated accident were based on Babcock and Wilcox analytical models and assumptions that conservatively account for the nass and energy release rates to the containment.
The mass and energy release rate data were then used as input data to a containment computer code which performs thermodynamic calcula-tions with appropriate considerations of containment heat removal mechanisms to calculate the pressure and temperature response of the containment. tiini-mum containment cooling was assumed in the analysis, i.e., two of four fan-coolers of the reactor building emergency cooling system anJ one of the two spray trains of the reactor building spray systen were assumed not to operate.
A spectrum of pipe break sizes for both the hot and cold legs were analyzed.
The highest calculated containment pressure was 52 psig (a 5.0 square foot break in the hot leg) which provides an acceptable 12 percent margin below the containment design pressure of 59 psig.
. Question 6 Does the design basis accident for the containment building include consideration of core degradation or core melt?
Response
The design basis accident for the containment buildirg does not include concideration of core degradation or core melt.
The loss-of-coolant accident for the containment building should not be confused with the loss-of-coolant accident used for the design of systens such as the emergency a re coolina system.
In the containment building design loss-of-coolant accident, core degradation or core melt is unlikely, whereas in the emergency core cooling system design loss-of-coolant accident, core degradation or core melt is more probable.
The containment building design is based on a loss-of-coolant accident which maximizes the mass and energy release rates to the containment atmosphere to obtain the maximum containment building pressure. What tnis means is that almost all of the reactor core stored energy is released to the containment atmosphere.
Therefore, in the deisgn basis accident for the containment building, the reactor core fuel temperature remains very low and core degradation is unlikely.
The emergency core cooling system design is based on a loss-of-coolant accident which minimizes the energy release from the reactor core to maximize the fuel temperature and core degradation or e re melt is more probable.
The results of these differing assumptions can be seen . the peak containment building pressures calculated for these two design, asis loss-of-coolant accidents.
For the
- containment building design accident a peak pressure of 52 psig was calculated, and for the emergency core cooling systen a peak pressure of about 16 psig was calculated.
Although the contail.nent building design basis accident does not include considerations for core degradation or core nelt, two of the engineered safety feature systems do.
These systems are the containment building spray system and the combustible gas control system.
The containment building spray includes a system for injecting sodium hydroxide into the spray water to accelerate the removal of aerosol fission products released to the containment atmosphere.
The design of the spray system is based on the assumed release of twenty-five percent of the equilibrium radioactive iodine and one hundred percent of the equilibrium radioactive noble gas to the containment atmosphere.
The combustible gas control system design assumes the emergency core cooling system is
,5 a degraded, but rot total failed, condition.
The amount of hydrogen contributed by the core metal water reaction as a result of degradation is assumed either to be five tirres the total amount of hydrogen calculated in the emergency core cooling syster 's design basis accident er the amount that would result from reaction of all tne metal on the outside surface of the cladding cylinders surrounding the fwel to a depth of 0.00023 inch, whichever amount is In addition, the F drogen produced from the radiolytic decomposition greater.
/
of the cooling water in the reactor core and in the containment sump is based on a fission product distribution model that assumes that one hundred percent of the noble gases, fifty percent of the halogens and one percent of the solids present in the reactor core are released to the containment.
. Question 7 What type of pressure transients would be expected inside the containment building as a result of a feedwater transient at Rancho Seco?
Response
Normally, a feedwater transient will not result in a pressure rise inside the containment building.
Because of the containment building's large volume and the containment ventilation system, it usually requires an enormous amount of energy to be released to the containment atmosphere over a relatively short period of time (a matter of minutes) before the containment building will experience an increase in pressure. The amount of energy needed to obtain a containment building pressure increase usually requires a pipe break in a system containina a high energy fluid.
Even at Three Mile Island, Unit 2 (TMI-2) where the feedwater transient developed into a loss-of-coolant accident and large amounts of energy were released to the containment building atmosphere, the pressure remained relatively low.
In fact, for approximately the first 9-1/2 hours after the start of the accident at TMI-2, the pressure in the containment building remained below 4 psig except for a short period of time (at approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the accident) when the containment pressure increased to abcut 4.3 psig.
There was, however, a pressure pulse of about 28 psig inside the containment building at about 9-1/2 hours after the start of the accident.
This pressure pulse was believed to be caused by the burning of hydrogen and lasted a short period of time.
By 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the start of the accident the containment
. pressure was again below 4 psig and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later the containment pressure was near atmospheric pressure where it has remained.
Question 8 Is the present Rancho Seco containment design adequate?
Response
The present Rancho Seco containment design is adequate.
The design of the containment building is based on a conservative pressure calculation resulting from release of the reactor coolant +o the containment atmosphere in the event of a loss-of-coolant accident.
The containment design basis includes the effects of stored energy in the reactor coolant system, decay energy, and energy from other sources such as the secondary system and metal-water reactions and assumes a single failure in the heat removal system.
The highest calculated pressure was 52 pounds per square inch gauge which is 12 percent Selow the containment design pressure of 59 pounds per square inch gauge.
It should be pointed out that the containment is capable of withstanding pressure in excess of 59 psig before containment integrity is lost.
In addition, the containment building meets the applicable design criteria of 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, which establish necessary design, fabrication, construction, test and performance requirements for structures, systems, and components important to safety.
e
(
THOMAS A. GREENE PROFESSIONAL QUALIFICATIONS CONTAINMENT SYSTEMS BRANCH OFFICE OF NUCLEAR REACTOR REGULATION
'I am a senior Systems Engineer in the Containment Systems Branch, Office of Nuclear Reactor Regulation.
In this position I am responsible for the technical review, analysis and evaluation of the containment and secondary containment functional design; containment subcompartment analysis, contain-ment heat removal systems, containment isolation system, combustible gas in containment, and containment leak testing program as described in Safety Analysis Reports to assure that nuclear power plants can be built and oper-ated without undue risk to the health and safety of the p>blic.
I also as-sist in the preparation of standards, guides, and codes for the design and operation of reactors which deal with the containment system.
From 1973 to present, I have been employed with the Atomic Energy Commission and the Nuclear Regulatory Commission which was established by the Energy Reorganization Act of 1974.
I have been the principal reviewer for a number of nuclear power plants and am a former member of the American Nuclear So-ciety Standard Committee 56.1, " Design Basis for Hydrogen Treatment in Con-tainments."
From 1969 to 1973 I was employed as a Nuclear Safety Engineer at Combustion Engineering, L'indsor, Connecticut.
My major area of responsibility was the analysis of the thermal-hydraulic response of a nuclear power plant during
.?-
I a hypothetical loss of coolant accident by the use of computer blowdown codes.
During this time period I closely followed the LOFT semiscale test progran at the National Reactor Testing Station, Idaho Falls, Idaho.
Fron 1958 to 1969, I was employed as a Nuclear Engineer at the San Francisco Eay Naval Shipyard, Valiejo, California.
My duties were in all areas of engineering related to the overhaul and refueling of nuclear power systems on naval sub arines and surface vessels.
Also, at this time I taught a night course in Basic Nuclear Engineering at John F. Kennedy University, "artir.e, California.
I received a vaster of Science Degree in Nuclear Engineering frcr the University of Arizona in 1969 and a Bachelor of Science Degree in Engineer-ing Physics from the University of Oklahoma in 1956.
While at the University of Arizona, I was a graduate assistant aad instructor in the radioisotopes and instrumentation course.
Since graduating from college, I have attended various courses in reactor technology and safety.
I have been a menber of the American Nuclear Society since 1965.
f
.