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W� NUr' EAR REGULATORY COMMISSION
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SEP 2 2 1983 -
Mr. 0. R. Lee, Vice President ( 7-
Electric Production 1al
Public Service of Colorado SEP 3 0 1983
P.O. Box 840
Denver, Colorado 80201 4REELEr. q
Dear Mr. Lee:
We have completed our review of the information you have provided related to
NUREG-0737, Item II. 8. 3, "Post-Accident Sampling System. " Our review has
concluded that Fort St. Vrain (FSV) meets the intent of 6 of the 11
criteria, that two of the criteria, those pertaining to boron and chloride,
are not applicable, and that additional information is required for the
three remaining criteria. The results of our review are contained in the
enclosed Preliminary Safety-Evaluation (SE).
We request that you review the SE and provide your comments for the three
criteria which have not been fully resolved, together with a schedule for full
resolution, within 30 days of your receipt of this letter. Upon completion of
our review of this additional information, we will issue a final SE.
Since this reporting requirement relates solely to FSV, fewer than ten
respondents are affected; therefore, OMB clearance is not required under
P. L. 96-511.
If you have any questions on this subject, please contact your NRC Project
Manager.
Sincerely,
C�ir�vyyr�-"�
frt. L. Madsen, Chief
OO Reactor Project Branch 1
Enlcosure:
Safety Evaluation
83O95O
11:10C'
Fort St. Vrain
cc list
C. K. Millen Chairman, Board of County Commissioners
Senior Vice President of Weld County, Colorado
Public Service Company Greeley, Colorado 80631
of Colorado
P. O. Box 840 Regional Representative
Denver, Colorado 80201 Radiation Programs
Environmental Protection Agency
James B. Graham, Manager 1860 Lincoln Street
Licensing and Regulation Denver, Colorado 80203
East Coast Office
General Atomic Company Don Warembourg
2021 K Street, NW, Suite 709 Nuclear Production Manager
Washington, DC 20006 Public Service Company of Colorado
P. 0. Box 368
J. K. Fuller, Vice President Platteville, Colorado 80651
Public Service Company
of Colorado Albert J. Hazle, Director
P. O. Box 840 Radiation Control Division
Denver, Colorado 80201 Department of Health
4210 East 11th Avenue
Denver, Colorado 80220
G. L. Plumlee
NRC Senior Resident Inspector Kelly, Stansfield & O' Donnell
P. O. Box 640 Public Service Company Building
Platteville, Colorado 80651 Room 900
550 15th Street
Denver, Colorado 80202
Director, Division of Planning Darrell G. Eisenhut, Director
Department of Local Affairs Division of Licensing
615 Columbine Building Office of Nuclear Reactor Regulation
1845 Sherman Street U. S. Nuclear Regulatory Commission
Denver, Colorado 80203 Washington, D. C. 20555
Preliminary
Safety Evaluation by
the Office of Nuclear Reactor Regulation
Related to Operation of
Fort St. Vrain Nuclear Generating Station
Public Service of Colorado
Docket No. 50-267
Post-Accident Sampling System (NUREG-0737, II.B.3)
Introduction
Subsequent to the TMI-2 incident, the need was recognized for an improved
post-accident sampling system (PASS) to determine the extent of core
degradation following a severe reactor accident. Criteria for an accept-
able sampling and analysis system for water-cooled reactors are specified
in NUREG-0737, Item II.B.3. For a gas-cooled reactor, these criteria need
to be modified to take account of the gaseous nature of the coolant and
the different gaseous species (e.g. CO, C02, H2, CH4 and certain radio-
nuclides) which need to be analyzed to indicate the severity of core
damage. The system should have the capability to obtain and quantitatively
analyze the helium coolant and reactor building atmosphere samples without
radiation exposure to any individual exceeding 5 rem to the whole body or
75 rem to the extremities (GDC-19) during and following an accident in
which there is core degradation.
To comply with the requirements for a PASS, the licensee should (1) review
and modify his sampling, chemical analysis, and radionuclide determination
capabilities as necessary and (2) provide the staff with information per-
taining to system design, analytical capabilities and procedures in
sufficient detail to demonstrate that the criteria for an acceptable PASS
are met.
- 2 -
Evaluation
By letter dated September 27, 1982, the licensee provided information on
the PASS in the format of the eleven criteria in NUREG-0737, Item II.8.3,
addressing the portions which were relevant to a gas-cooled reactor.
Two of the eleven criteria (#5 and #7) are not applicable to a gas-cooled
reactor, but the other nine provide a basis for assessing the acceptability
of a PASS. The terms "containment" and "reactor coolant" are to be
interpreted as "reactor building" and "gaseous helium coolant" for the
gas-cooled reactor.
Criterion (1 ):
The licensee shall have the capability to promptly obtain reactor
coolant samples and containment atmosphere samples. The combined
time allotted for sampling and analysis should be three hours or
less from the time a decision is made to take a sample.
Because the normal and post-accident sampling systems are the same, the
licensee has the sampling and analysis capability to promptly obtain
and analyze helium coolant samples and reactor building atmosphere
samples within three hours from the time a decision is made to take a
sample. The PASS electrical power supply is from the normal station
service power supply. In the event of a loss of off-site power, two
gas-driven generators provide emergency power for sampling. We find
that these provisions meet Criterion (1) and are, therefore, acceptable.
Criterion (2):
The licensee shall establish an onsite radiological and chemical
analysis capability to provide, within the three-hour time frame
established above, quantification of the following:
- 3 -
a) certain radionuclides in the reactor coolant and containment
atmosphere that may be indicators of the degree of core damage
(e.g. , noble gases; iodines and ceisums, and nonvolatile
isotopes) ;
b) hydrogen levels in the containment atmosphere;
c) dissolved gases (e.g. , H2) , chloride (time allotted for analysis
subject to discussion below), and boron concentration of
liquids;
d) Alternatively, have in-line monitoring capabilities to perform
all or part of the above analyses.
The PASS provides in-line monitoring for noble gas activity, CO and moisture
in the helium coolant, as well as for radioactivity in the reactor building
stack gas. The PASS also provides the capability to collect grab samples
of the coolant and of the reactor building atmosphere that can be trans-
ported to the radio-chemical laboratory for CO, CO2, H2, CH4, N2 and
radionuclide analyses. These species are the indicators of core damage
in a gas-cooled reactor, and their relative magnitudes indicate core
temperature, fuel particle failure, air ingress or water ingress. We
find that the licensee partially meets Criterion (2) by establishing an
on site radiological and chemical analysis capability. However, the
licensee should provide a procedure, consistent with the clarification
of NUREG-0737, Item II. B.3, Post Accident Sampling System, transmitted
to the licensee on July 9, 1982, to estimate the extent of core damage
based on radionuclide concentrations and taking into consideration other
physical parameters such as core temperature data. Guidance for the
procedure to estimate core damage for a water-cooled reactor is attached
(Attachment 1). The procedure for estimating core damage should be
consistent with those portions of these recommendations which are
applicable to a gas-cooled reactor.
- 4 -
Criterion (3):
Reactor coolant and containment atmosphere sampling during post
accident conditions shall not require an isolated auxiliary system
(e. g. , the letdown system or the reactor water cleanup system) to
be placed in operation in order to use the sampling system.
Helium coolant and reactor building atmosphere sampling during post accident
conditions do not require an insolated auxiliary system to be placed in
operation in order to perform the sampling function. Sample recirculation
and collection during normal and post-accident conditions utilize the same
systems.
Criterion (3) is acceptable because PASS sampling is performed without
requiring operation of an isolated auxiliary system and the PASS valves
are accessible after an accident.
Criterion (4):
Pressurized reactor coolant samples are not required if the
licensee can quantify the amount of dissolved gases with
unpressurized reactor coolant samples. The measurement of
either total dissolved gases or H2 gas in reactor coolant
samples is considered adequate. Measuring the 02 concen-
tration is recommended, but is not mandatory.
Only negligible quantities of hydrogen and other combustible gases would
be liberated by the overheating of the graphite core. However, water
leaking into the helium coolant would react with the graphite core,
generating H2 and CO. In-line monitoring is provided for CO and moisture.
Also grab samples of the coolant are taken which may be analyzed for H2,
CO, CO2' CH4 and N2. We find that these provisions meet Criterion (4)
of Item II. B.3 of NUREG-0737 and are, therefore, acceptable.
- 5 -
Criterion (5):
The time for a chloride analysis to be performed is dependent upon
two factors: (a) if the plant' s coolant water is seawater or brack-
ish water and (b) if there is only a single barrier between primary
containment systems and the cooling water. Under both of the above
conditions the licensee shall provide for a chloride analysis within
24 hours of the sample being taken. For all other cases, the
licensee shall provide for the analysis to be completed within 4
days. The chloride analysis does not have to be done onsite.
This criterion is not pertinent to a gas-cooled reactor, because the
potential for chloride-containing materials entering the reactor system
is nil .
Criterion (6):
The design basis for plant equipment for reactor coolant and contain-
ment atmosphere sampling and analysis must assume that it is possible
to obtain and analyze a sample without radiation exposures to any
individual exceeding the criteria of GDC-19 (Appendix A, 10 CFR Part
50) (i . e. , 5 rem whole body, 75 rem extremities). (Note that the
design and operational review criterion was changed from the
operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC-19
criterion (October 30, 1979 letter from H. R. Denton to all licensees. ))
The licensee has performed an activity release and shielding analysis to
ensure that operator exposure while obtaining and analyzing a PASS sample
is within the acceptable limits. The analysis is based on the severe
accident identified in Section 14.10 and Appendix D of the FSAR as DBA#1.
From our review of this analysis, we conclude that PASS personnel radi-
ation exposures from reactor coolant and containment atmosphere sampling
and analysis will be within 5 rem whole body and 75 rem extremities, which
meet the requirements of GDC-19 and Criterion (6) and are, therefore,
acceptable.
- 6 -
Criterion (7):
The analysis of primary coolant samples for boron is required for
PWRs. (Note that Rev. 2 of Regulatory Guide 1.97 specifies the
need for primary coolant boron analysis capability at BWR plants. )
This criterion is not applicable to a gas-cooled reactor, because there
is no volatile boron in the reactor system.
Criterion (8):
If in-line monitoring is used for any sampling and analytical
capability specified herein, the licensee shall provide backup
sampling through grab samples, and shall demonstrate the
capability of analyzing the samples. Established planning for
analysis at offsite facilities is acceptable. Equipment
provided for backup sampling shall be capable of providing at
least one sample per week until the accident condition no longer
exists.
In-line monitoring of CO, moisture and noble gas activities is provided
for the helium coolant. Radioactivity in the reactor building stack is
continuously monitored. A backup helium coolant grab sample and a
reactor building atmosphere grab sample can be obtained for on site
and off site analysis. Samples can be transported to the University of
Colorado facilities for off site analysis in approximately one hour. We
find these provisions meets Criterion (8) and are, therefore, acceptable.
Criterion (9):
The licensee' s radiological and chemical sample analysis capability
shall include provisions to:
- 7 -
a) Identify and quantify the isotopes of the nuclide categories
discussed above to levels corresponding to the source terms
given in Regulatory Guide 1.3 or 1.4 and 1.7. Where necessary
and practicable, the ability to dilute samples to provide
capability for measurement and reduction of personnel exposure
should be provided. Sensitivity of onsite liquid sample
analysis capability should be such as to permit measurement
of nuclide concentration in the range from approximately
ip Ci/g to 10 Ci/g.
b) Restrict background levels of radiation in the radiological
and chemical analysis facility from sources such that the
sample analysis will provide results with an acceptably small
error (approximately a factor of 2). This can be accomplished
through the use of sufficient shielding around samples and out-
side sources, and by the use of a ventilation system design
which will control the presence of airborne radioactivity.
The radionuclides in both the helium coolant and the reactor building
atmosphere samples will be identified and quantified using the on site
gamma spectrometer. Radiation background levels will be restricted by
shielding. Ventilated radiological and chemical analysis facilities
are provided to obtain results within an acceptably small error
(approximately a factor of 2). We find that these provisions partially
meet Criterion (9). However, between approximately 5 hours and 7 days
after the onset of a loss-of-cooling accident, the activity of the
coolant sample will exceed the measurement capability of the gamma
spectrometer. The licensee should provide on site capability to measure
these higher activities by means such as sample dilution or collimation
of the sample beam.
Criterion (10):
Accuracy, range, and sensitivity shall be adequate to provide
pertinent data to the operator in order to describe the radio-
logical and chemical status of the reactor coolant systems.
- 8 -
The accuracy and sensitivity of the PASS instruments and analytical
procedures are consistent with the recommendations and the clarifica-
tions of NUREG-0737, Item II.B.3, Post-Accident Sampling Capability,
transmitted to the licensee on June 30, 1982. Therefore, they are
adequate for describing the radiological and chemical status of the
reactor core during normal operation and the first few hours of a
severe accident. The analytical methods and instrumentation are capable
of operation in the post-accident sampling environment. No additional
training of chemistry personnel is required because the same systems
are used for normal and post-accident sampling and analysis. However,
no provisions are made for the on-site measurement of radionuclide con-
centrations in the helium coolant in the post-accident period beyond
the first few hours.
We find that the licensee partially meets Criterion (10). The licensee
should provide additional information on the measurement of coolant
activities in the time period beyond the first few hours after the
onset of a severe accident.
Criterion (11):
In the design of the post accident sampling and analysis capability,
consideration should be given to the following items:
a) Provisions for purging sample lines, for reducing plateout in
sample lines, for minimizing sample loss or distortion, for
preventing blockage of sample lines by loose material in the
RCS or containment, for appropriate disposal of the samples,
and for flow restrictions to limit reactor coolant loss from
a rupture of the sample line. The post accident reactor coolant
and containment atmosphere samples should be representative of
the reactor coolant in the core area and the containment atmos-
phere following a transient or accident. The sample lines
should be as short as possible to minimize the volume of fluid
to be taken from containment. The residues of sample collection
should be returned to containment or to a closed system.
- 9 -
b) The ventilation exhaust from the sampling station should be
filtered with charcoal adsorbers and high-efficiency particulate
air (HEPA) filters.
The licesee has addressed provisions for recirculation and purging to
ensure that samples are representative. Ventilation exhaust from PASS
is filtered through charcoal adsorbers and HEPA filters. We determined that
these provisions meet Criterion (11) of Item II.6. 3 of NUREG-0737, and are,
therefore, acceptable.
Conclusion
We conclude that the post-accident sampling system partially meets the
criteria of Item II.8. 3 of NUREG-0737. Two of the eleven criteria
are not applicable to a gas-cooled reactor. The licensee' s proposed
methods to meet six of the remaining nine criteria are acceptable.
The three criteria which have not been fully resolved are:
Criterion (2): Provide a core damage estimate procedure to include radio-
nuclide concentrations and other physical parameters as
indicators or core damage.
Criterion (9) Develop procedures to identify and quantify radionuclides
and (10): in coolant samples for time periods longer than a few hours
after the onset of an accident.
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