ML20150D939
| ML20150D939 | |
| Person / Time | |
|---|---|
| Issue date: | 06/28/1988 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Gallo D HOUSE OF REP. |
| Shared Package | |
| ML20150D941 | List: |
| References | |
| NUDOCS 8807140244 | |
| Download: ML20150D939 (5) | |
Text
h v it t n.D _) & A u.)
7
/
'o UNITED STATES
~g 8"
NUCLEAR REGULAYORY COMMISSION n
E WASHINGTON, D. C. 20555
%*****]
June 28, 1938 The Honorable Dean Gallo United States House of Representatives A shington, D.C.
20515
Dear Congressman Gallo:
Your letter of May 6, 1988 transmitted for our information and comment a con-stituent's correspondence which expressed serious concern over the safety of boiling water reactor (BWR) nuclear power plants with Mark I pressure sup-pression containment designs.
In particular your constituent voiced concern over his proximity to the Fitzpatrick Station, Nine Mile Point Unit 1, Hope Creek Station and Oyster Creek Station.
After review of the correspondence we concluded that the author makes three general assertions:
(1) that the Mark I design was defective and cited as such by the NRC, (2) that the Mark I containment is stated by the NRC to have 45 percent probability of failure over the remaining life of all U.S.
reactor plants, and (3) that the Mark I containment design is essentially the same as that used at Chernobyl.
With regard to the adequacy of the Mark I containment design, the NRC con-ducted extensive evaluations of the design to assure all required safety standards and criteria were met prior to licensing plants using that containment.
The acceptance criteria included the ability to withstand nut.trous events and still provide a "containment" function.
The events, known as design basis accidents, encompass a large spectrum of incidents, such as failure of electric power supplies or rupture of major piping systems, some of which are considered very unlikely.
The NRC has concluded that the Mark I containment designs are acceptable based on their ability to survive and successfully mitigate these accidents.
With regard to the issue of containment failure probability, we believe your constituent's assertion is based on 1985 HRC testimony before the Subconmittee on Energy Conservation and Power, where we stated that the probability of experiencing a core meltdown in the U.S. in the next 20 years may be 45 percent.
To consider this testimony as supporting a claim of 45 percent Mark I containment failure probability is incorrect.
This estimate represented the likelihood of a plant arriving at a point in an accident sequence where there was insufficient core cooling to maintain fuel integrity.
Having incurred a failure of that type does not mean that the accident sequence necessarily progresses to either con-4 tainment failure or large radioactive releases, neither of which occurred n'
during the Three Mile Island accident.
Also, the 1985 testimony was an
$vl average cver all U.S. plants and containment types--not specifically the d
Mark I containment plants, fq#
M u
e,jd 5
f.h MlFWs44 gggg
?
PL M
a The Honorable Dean Gallo P Additionally, it should be understood that the 45 percent estimate was cal-culated by assuming a core melt value of 3x10 4 per year for each plant, which was the average of a number of early probabilistic risk assessment studies, only one of which was for a Mark I plant.
The testimony in which that number was referred to also indicated that it may have been an overestimate.
More recent work by the NRC's Office of Research has provided a rebaseline of plant risk based on more advanced models of plant behavior and assessment of various plant modifications which have taken place since the Three Mile Island accident.
This study, known as NUREG-1150, "Reactor Risk Reference Document" (draft),
included an evaluation of the Peach Bottom plant, which has a BWR Mark I containment. The study showed that plant improvements and more accurate analytical models resulted in a core damage estimate of 8x10 8 per year (less than one chance per 100,000).
The study also considered post core melt containment behavior and indicated a result of very low public risk.
Risk estimates are affected by plant specific design features, plant locaticn and containment performance behavior following a core melt; therefore, estimates of the average risk of a group of plants contain considerable uncertainty.
Later in this response, we identify actions undtrway to examine all operating reactors individually to confirm judgment of low public risk and to identify further reductions in plant risk which may be warranted.
The final concern of your constituent is that the Mark I containment design is the same as that at Chernobyl, which failed to mitigate serious accident consequences.
The Chernobyl plant did not have a containment at all comparable to the essentially airtight pressure vessel common to U.S. reactor plants.
The NRC has studied the Chernobyl accident to determine any safety implications for U.S. reactors.
The results of this study are documented in draft NUREG-1251, "Implications of the Accident at Chernobyl for Safety Regulation of Commercial Nucler Power Plants in the United States" (August 1987) and clearly indicate that the designs are not even remotely similar and that the containment vul-nerabilities of Chernobyl are not present in any commercial U.S; facilities.
While we believe that severe accident risks are low at operating nuclear plants, our goal is to identify and evaluate improvements that would achieve even lower levels of public risk.
To assure that our risk conclusions are applicable to j
all operating units, a number of programs are underway to address the severe accident sequences--those that would result in core damage if not arrested.
One such major effort is the Individual Plant Examination, a plant specific study by each utility to identify any severe accident vulnerability, both from the perspective of accident frequences and from containment performance following a postulated core melt.
As part of the program, any problems identified will be dealt with by the utility.
Even though the risk posed by Mark I containment facilities has been found by past studies to be acceptable, the NRC has continued to investigate means to improve the containment performance for those plants, and thus improve everall plant safety.
The impetus for this continued research stems from concern over
=.
The Honorable Dean Gallo uncertainty regarding the Mark I containment performance following a core melt, as reflected in NUREG-1150, as well as a belief that reasonably cost-effective plant improvements can be identified.
To that end, the NRC has initiated the Mark I Containment Performance Improve-ment Program.
The principal objective of the program is to evaluate technical issues associated with core melt accident phenomena and evaluate potential improvements to the Mark I design.
This program is being implemented as a high priority and should provide a firm and timely basis for deciding an ap-propriate course of action.
An interim report to the Commission on these activities is expected by the end of July 1988.
This report will address differences in existing risk studies and indicate whether existing analyses justify changes on the Mark I plants.
A final report to the Commission is scheduled for September 1988.
In summary, the calculated failure probability for Mark I containments does not constitute an unacceptable risk to the public i
health and safety.
Nonetheless the Commission, consistent with its defense-in-depth philosophy, is pur'uing methods for further reducing overall risk.
s Sincerely, (Signed) T. A, nchm ictor Stello, Jr.
> Ex9cutive Director for Operations
- See previous concurrence i
PSB: DEST PSB: DEST PSB: DEST SAD: DEST SAD: DEST D: DEST
- CTinkler:dj
- JKudrick
- JCraig
- MRubin
- AThadant
/ gMuhy VStello j
T rtin
'ezek 6/2.t/88 i/88 6TLT,/88 6/ /88 6FT/88 8) l
-.e
DISTRIBUTION:
Docket file FGillespie PSB Rdg ELJordan ED0 #3706 SECY EDO Rdg NRC POR TMurley/JSniezek AThadani Gillespie OGC-Beth OCA VStello DMossburg(0003706)
LShao CTinkler FMiraglia TMartin BSP Green Ticket File i
t b
r W
f J
i i
i
(
)
f~
r
.,y:Au%
9 h,
UNITED STATES 1
g
- [
g NUCLEAR REGULATORY COMMISSION j
~
E WASHINGTON, D. C. 20655 r
5 0,,,,,
EDO Principal Correspondence Control FROM:
DUE: 06/02/88 EDO CONTROL - 0003706 DOC DT: 05/06/88 FINAL REPLY:
Rnp. Dean A.
Gallo TO:
Ed Jordan, AEOD FOR SIGNATURE OF:
- GRN CRC NO:
Executive Director DESC:
ROUTING:
ENCLOSES LETTER FROM HENRY GLUCKSTERN CONCERNING-Jordan SAFETY OF NUCLEAR GENERATING STATIONS THAT USE THE OCA GENERAL ELECTRIC MARK I PRESSURE SUPPRESSION SECY CONTAINMENT SYSTEM DATE: 05/18/88 g$
ASSIGNED TO:
CONTACT:
i I
NRR Murley
() y SPECIAL INSTRUCTIONS'OR REMARKS:
NRR RECEIVED:.MAY 19,.1988 y
ACTION:
H DEST:SHAOL '
$[Y NRR ROUTING:
MURLEY/SNIEZEK MIRAGLIA MARTIN GILLESPIE j
ACTION mSSsUaG DUE TO NRR DIRECTOR'S 051CE BY N td-A7/
U
-