IR 05000335/1988004
| ML17221A745 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/18/1988 |
| From: | Gloersen W, Kahle J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17221A744 | List: |
| References | |
| 50-335-88-04, 50-335-88-4, 50-389-88-04, 50-389-88-4, NUDOCS 8805030013 | |
| Download: ML17221A745 (21) | |
Text
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t UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTAST., N.W.
ATLANTA,GEORGIA 30323 yq'ig tlPR 2 Report NoseI 50-335/88-04, 50-389/88-04 Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33102 Docket NoseI 50-335, 50-328 Facility Name:
St. Lucie License Nos e I DPR-67, NPF-16 Inspection Conducted:
March 18, 1988 v
t Inspector:
o sen Accompanying Personnel:
E.
Branagan (NRR)
r Approved by: ~
C~M
.
Ka e, Section ie D'$ision of Radiation Safety and Safeguards ate Signe Da e S gned SUMMARY Scope:
This special, unannounced inspection was conducted in the areas of liquid and gaseous radwaste systems, offsite dose calculation methodology, and effluent monitoring and sampling.
Results:
No violations or deviations were identified.
8805030013 88042i PDR ADOCK 05000335
REPORT DETAILS Persons Contacted Licensee Employees
- A. Bailey, guality Assurance Supervisor
- J. Baysinger, Senior guality Assurance Analyst
- H. Buchanan, Health Physics Supervisor
- R. Cox, Effluents Supervisor D. Faulkner, Primary and Effluent Monitor Supervisor
- R. Frechette, Chemistry Supervisor
- K. Harris, Vice President
- R. Jennings, Technical Advisor
- C. Pell, Technical Supervisor
- R. Sipes, Services Manager Other licensee employees contacted included engineers, technicians, operators, and office personnel.
NRC Resident Inspector
"P. Bibb
- Attended exit interview Exit Interview The inspection scope and findings were summarized on March 18, 1988, with those persons indicated in Paragraph 1.
The inspector described the areas examined and discussed in detail the inspection findings listed below.
No dissenting comments were received from the licensee.
The licensee did not identify as proprietary any of the material provided to or reviewed by the inspector during this inspection.
Licensee Action on Previous Enforcement Matters (Closed)
50-335/87-11-01 (VIO):
Inadequate procedures for operation of waste gas system.
The inspector reviewed the licensee's corrective actions pertaining to the violation identified above.
The violation occurred because the Unit 1 System Operating Procedure 1-0530020 was not updated as required to be consistent with the current mode of operation.
The inspector reviewed the revision to Operating Procedure 1-0530020, Waste Gas System Operation, dated August 14, 1987, and noted that the appropriate changes had been made.
The licensee indicated that responsible plant personnel have been made aware of the violation and the need to address various changes in operating methodology.
This violation is considered close.
Audits (84723, 84724)
Technical Specification 6.5.2.8 requires audits to be performed under the cognizance of the Company Nuclear Review Board (CNRB).
The audits are required to encompass:
the conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months; the radiological environmental monitoring program and the results thereof at least once per 12 months; and the Offsite Dose Calculation Manual and implementing procedures at least once per 24 months.
The inspector reviewed the following audits:
QSL-OPS-87-538, PSL-1 and 2, Technical Specification 3/4.7 "Plant Systems,"
August 25, 1987 QSL-OPS-87-553, Radiological Effluents and Honitoring, September 22, 1987 QSL-OPS-87-557, Facility Staff Training and Qualification, February 26, 1988 QSL-OPS-88-583, January Performance Monitoring Audit (PASS),
January 11, 1988 The inspector reviewed the audits and noted that no significant findings were identified in the audits.
No violations or deviations were identified.
5.
Effluent Reports (84723, 84724)
Technical Specification 6.9.1.7 requires the licensee to submit within 60 days of January 1 and July 1 of each year, routine Radioactive Effluent Release Reports covering operation of the unit during the previous six months of operation.
The reports are required to include a summary of the quantities of radioactive material released from the unit as outlined in Regulatory Guide 1.21.
Additionally, reports that are submitted 60 days after January 1 of each year are required to include an assessment of radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.
The inspector reviewed the Semiannual Radiological Effluent Release Report for the period July 1, 1987, through December 31, 1987.
The review included an examination of the liquid and gaseous effluent release data as well as dose estimate data.
Selected data from this report and previous reports are presented in Table 1.
The inspector noted that quantities of radioactive noble gases released during 1984 through 1986 were significantly higher than other Region II operating pressurized water reactors (PWRs).
Due to the licensee's extensive efforts in reducing radioactive gaseous effluent releases which included the identification, removal, and replacement of individual defective fuel rods in both units (referred to as "reconstitution"),
noble gas releases during 1987 were
nearly three times lower than the previous year (see Table 1). It should be noted that both units were operating at nearly full capacity during 1987 (approximately 290 days for Unit 1; 296 days for Unit 2; each unit was in a 60 day outage).
Another source of noble gas was from the leaking power-operated relief valves (PORVs).
Since February 1988, both PORVs on each unit have been isolated during full power operations.
Isolation of the PORVs at full power has resulted in a reduction in gaseous effluent levels.
Radioactive liquid releases during 1984 through 1986 were also higher than most of the other Region II P>JRs.
In 1987, liquid releases of fission and activation products were reduced by a factor of more than four relative to 1986 releases.
The lower releases of radioactive liquid effluents can be attributed to the licensee's fuel reconstitution program mentioned above and implementation of a
liquid radwaste reduction program (see Paragraph 6).
Lower releases in both liquid and gaseous effluents resulted in lower estimated offsite doses.
The estimated dose due to ingestion of fish and shellfish by an adult via the liquid pathway during 1987 was 6.95 E-02 mrem.
Doses in previous years via this pathway were still well below the offsite environmental dose limits specified in
CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operations.
Estimated total doses for 1987 due to noble gas and gaseous particulates and iodine exposures via the inhalation pathway, the grass-cow-milk ingestion pathway and the ground plane exposure pathway were 5.08 E-03 mrem to the whole body and 1.80 E+0 mrem to the thyroid.
Again, doses via these pathways in previous years were below the limits specified in 40 CFR 190.
The inspector also reviewed monthly plant reports to upper level management that highlighted programs and indicators of performance in areas such as:
plant operations, maintenance, radiological protection, radiological effluents and solid waste, quality assurance, and chemistry.
The following reports were reviewed:
( 1) Performance Monitoring Manage-ment Information Report for December 1987 (dated February 3, 1988)
and; (2) Monthly Preliminary Release Reports (dated February 1988).
Additionally, the inspector selectively reviewed both liquid and gaseous release permits for the period covering January 1988 to the time of this inspection.
The review included an examination of the gamma isotopic analytical data, results of monthly composites for tritium and gross alpha, results of quarterly composites for Sr-89, Sr-90, and Fe-55, and effluent monitor source checks before each release.
The inspector noted that Operations Procedure 1-0510022 only had a place for a checkoff indicating that a
source check had been performed.
The inspector discussed with the licensee the benefits of recording the effluent monitor
'indication before and after the detector was exposed to a check source in order to verify monitor operability.
The licensee acknowledged the inspector's comments.
No violations or deviations were identifie s.
7.
Liquid Radioactive Waste Systems (84723)
The inspector reviewed the liquid radioactive waste system operation.
In 1987, 1986 and 1985 total plant releases of liquid radioactive waste containing fission and activation products to the environment were 1. 14 curies, 4.96 curies, and 5.51 curies, respectively.
Beginning approximately February 1986, the licensee had implemented a guality Improvement Program (gIP) designed to reduce the curie content of liquids released from the plant and consequently to reduce the dose impact of radiological liquid effluents.
The liquid radwaste reduction program involved many technical improvements and administrative changes.
Some of the improvements and administrative changes implemented thus far included:
(1) recirculation of the aerated waste storage tank (AWST) through the waste ion exchangers; (2) removal of sludge from the liquid waste storage tanks every 12-18 months; (3) diversion of high curie content input sources such as spent resin cask dewatering; (4) use of charcoal pre-filtration in the waste ion exchangers to extend demineralizer capacity; (5) administrative control of laundry'astes so that these wastes can be processed separately and kept segregated from the AWST and the waste ion exchangers (high conductivity laundry wastes can shorten the life of the waste ion exchangers);
(6) placement of liquid radwaste reduction
"awareness" signs on floor drains; and (7) implementation of the health physics and maintenance leak identification program.
Improvements being considered included installation of conductivity trip monitors upstream from the AWST and installation of
"bag" or roughing pre-filters downstream from the
"equipment drain and chemical drain tanks.
At the same time, a parallel effort to reduce gaseous releases by improving the-fuel integrity had the added benefit of reducing the curie input to the liquid radwaste system.
No violations or deviations were identified.
Gaseous Radioactive Waste Systems (84724)
The inspector reviewed the licensee's programs, procedures, and equipment provided for the collection, monitoring, analysis, processing, treatment, and release to the environment of radioactive or potentially radioactive gases.
The waste gas processing system at St. Lucie Units 1 and 2 was provided for the maintenance and control of the primary coolant cover gas system.
In the original design of the waste gas processing system, waste gas was "bled" from the cover gas system to a surge tank, from which the gas was periodically pumped to a series of three waste gas decay tanks.
The system was designed to retain or "holdup" waste gas for an average of 30 days, after which the gas, from which relatively short-lived noble gases had decayed to a stable non-radioactive gaseous form, would be discharged to the environment at a substantially reduced radioactivity leve A number of adverse factors were encountered at the St. Lucie facility which made the use of the design basis operating mode of the waste gas processing system impracticable.
The following factors were noted:
The design volume capacity of the installed waste gas decay tanks (WGDT) was inadequate to accommodate the volume of waste gas actually generated.
Each WGDT was approximately 144 cubic feet in volume and had a design operating pressure of about 150 psig.
Applying the method of NUREG-0017,
"Calculation of Releases of Radioactive Haterials in Gaseous and Liquid Effluents from Pressurized Mater Reactors (PWR-GALE-CODE)" the inspector calculated a 10-day fill time and a
10-day holdup period before each release.
In the FSAR, the licensee calculated average total fill and holdup time to be 30 days.
In practice, the licensee found the fill and holdup times to be on the order of two to four days, primarily as a result of higher than anticipated generation, of radioactive gases and system air and gas inleakage.
Both Units
and 2 experienced a higher than normal degree of fuel'ailure.
The fuel failure resulted in higher than normal concentrations of gaseous fission products and radioiodine in the waste gas system discharges and in releases from other plant discharge paths.
Concentration of these gases in pressurized waste gas decay tanks and related systems resulted in increased occupational exposures to plant workers.
The presence of higher than normal defective fuel at St. Lucie also resulted in higher-than-normal releases of noble gas and iodine fission products from other gaseous discharge points within the facilities.
The auxiliary building normal ventilation exhaust system and the containment purge exhaust became sources of effluents comparable to, and often larger than releases from the waste gas system.
Since 1984, the licensee had discontinued use of the waste gas decay tank system during steady-state power operations in favor of direct continuous discharge to the environment from the waste gas surge tank to the plant vent.
It was noted that the WGDTs were still operational.
The inspector reviewed this mode of operation relative to compliance with the plant Technical Specifications and
CFR 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to meet the criterion
"As Low As Reasonably Achievable" for Radioactive Materials in Light-Water-Cooled Nuclear Power Reactor Effluents.
The inspector and licensee representatives discussed calculations of air doses to unrestricted areas resulting from plant operation.
The licensee estimated that the use of the WGDTs in 1987 would have lowered the population dose to the general public from about 0.4 person-rem to the whole-body to about 0. 13 person-rem.
These dose estimates were based on releases from Units
and 2 combined.
This reduction in dose to the general public would have increased the occupational dose by at least 0.2 person-rem mainly due to the additional maintenance of the MGDTs and waste gas compressors.
This
occupational dose estimate was based on specific radiation work permits issued during the last year that the WGDTs were in operation.
It is possible that some exposure was not specifically listed for WGDTs, thus, the occupational exposure estimate was most likely underestimated.
The bases for these dose estimates were discussed with licensee representatives, and these estimates appeared reasonable.
Technical Specification 3. 11.2.4 requires that appropriate portions of the gaseous radwaste treatment system be used to reduce radioactive materials in gaseous waste prior to discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site, to unrestricted areas, when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation.
The inspector reviewed licensee calculations that demonstrated that the releases per unit per 31 day period resulted in air doses that were less than the action levels
.
specified in Technical Specification 3. 11.2.4.
Technical Specification 4. 11.2.4.2 requires that the gaseous radwaste treatment system and ventilation exhaust treatment system shall be demonstrated operable by operating the gaseous radwaste treatment system equipment arid ventilation exhaust treatment system equipment for at least 30 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.
The inspector noted that in 1987, the licensee released
WGDT volumes for Unit 1 and
WGDT volumes for Unit 2.
The use of the WGDT system satisfied the Technical Specification noted above.
The licensee elected to use the WGDT system during reactor power change evolutions.
Licensee representatives acknowledged that their review of the waste gas system operation had identified the capacity and number of waste gas decay tanks provided in the original design as inadequate to permit operation in accordance with the design concept of operation.
The licensee initiated a
guality Improvement Program to determine methods for improving the design of the gaseous waste treatment system.
The following engineering solutions were discussed:
(1) regular draining of water from the quench tanks, gas surge tanks, and waste gas decay tanks (to prevent water intrusion to the waste gas compressor);
(2) design modifications of the waste gas compressor so that the component would be less susceptible to water intrusion; and (3) minimization of oxygen inleakage (explosion hazard)
by replacing the diaphragm type valves (a source of oxygen inleakage)
with ball valves at the inlet and outlet of each WGDT.
It should be noted that the licensee was planning to use the WGDT system by December 1988 on a regular basis.
In addition to the engineering studies noted, above, the licensee has improved the integrity of the fuel by new fuel design and "reconstitution" of fuel assemblies.
Unit 1 was refueled in Harch and April of 1987.
After a preliminary run at full power, the unit was shut down and all the fuel was examined using ultrasonic nondestructive testing methods, and underwater video camera techniques.
The defective fuel rods were identified and removed from the fuel bundles and replaced with pre-tested rods.
The licensee referred to the process of replacing individual fuel
rods as "reconstitution."
The benefits of the reconstitution process were noted in the significantly reduced monthly noble gas effluent levels from May 1987 - January 1988, which were a factor of ten lower than during the previous fuel cycle.
Prior to refueling, Unit 1 had a level of defective fuel that resulted in abnormally high noble gas releases from the plant.
Unit 2 was refueled in October and November 1987, and the fuel was 1005 ultrasonically tested and reconstituted.
Gaseous effluent releases during the first few months after the outage were lower, however, the time span was too brief to note any trends.
The licensee was evaluating the benefits desired from tuel reconstitution for consideration in subsequent refueling outages.
Several other factors were being studied or implemented with regard to fuel design improvement to reduce fretting of fuel pins and thus reduce the likelihood of fuel pin failures.
The licensee was considering using new fuel that was fabricated using an approximately three inch-long solid zircalloy cap at the lower end of each fuel rod.
I-uel pellet conditioning prior to loading was used to reduce the moisture content, which was suspected to have been a factor in earlier fuel failure mechanisms.
Additionally, the licensee, during power changes, would either decrease or increase reactor power at a slow ramp which was expected to result in a lower incidence of fuel failure.
.
No violations or deviations were identified.
Offsite Dose Calculational Manual (ODCM)
The inspector reviewed the St. Lucie Plant Chemistry Operating Procedure No. C-200, Revision 8, Offsite Dose Calculational Manual," dated May 28, 1987.
The ODCM contained two types of dose calculation procedures for Units 1 and 2:
(1)
a simplified procedure that was used for the principal dose contributing nuclide(s);
and (2)
a more detailed procedure that calculated doses for each nuclide.
Both procedures were used to calculate doses to the maximally exposed member of the general public, not to the general population.
The methodology for calculating doses from exposure to radioactive gaseous effluents (ODCM, Ch. 2, 3, 4, and relevant portions of Appendices A and C)
was reviewed in detail since gaseous effluent releases from the St. Lucie Plant have been larger than those from other plants. in Region II.
Table
lists the effluents from St. Lucie Units'
and 2 over the period 1985 through 1987.
A spot check of the dose factors (e.g.,
ODCM Tables G-2, G-3, G-5, G-6, and G-7) for the principal dose contributing noble gases and radioiodines confirmed that the values, were acceptable.
The methodology for calculating doses was acceptable.
The methodology for determining the setpoint for release of noble gases was checked and confirmed to be 0.35 Ci/sec (assuming one release point).
As of December 17, 1987, 41% of this value was allotted to the plant vent, and this resulted in a high alarm setpoint of 0.144 Ci/sec for the plant vent for each unit.
The inspector confirmed that the correct setpoints I
were keyed into the control room moni tors for Units
and 2.
The Chemistry Department formally checked and documented that the setpoints in the control room were correct once per month.
Informal checks were performed more frequently.
The inspector performed hand calculations of the doses to an individual maximally exposed to noble gases released during the year 1987.
The hand calculations were within reasonable agreement of the licensee's reported values in the Semiannual Report for July 1, 1987, through December 31, 1987, which were based on a
more detailed procedure and used the meteorological data for the year 1987.
The results of the 1986 and 1987 Land Use Censuses were reviewed and nothing unusual was noted.
Several typographical errors (on pages 18, 19, 20, and 23) were noted; however, these errors were not significant.
No violations or deviations were identified.
9.
Radiological Effluent Monitoring and Sampling (84723, 84724)
The inspector reviewed the licensee's program for monitoring and sampling of radioactive liquid and gaseous process and effluent streams.
The inspector observed the operation of the process and effluent monitors during a tour of the facilities.
The monitors that were observed during the tour appeared to be functioning properly.
Calibration and maintenance records for the Unit 2 effluent monitors (performed October 1987)
and documentation of NBS traceability of calibration sources were reviewed and
'ppeared to be satisfactory.
The inspector walked through a simulated sampling of liquid effluents prior to release.
The procedures and technical methods appeared to be adequate.
Pre-release and post-release
~records for radioactive liquid effluents were reviewed and appeared to be in order.
The inspector also discussed with licensee representatives the maintenance and operational history of the various process and radiological effluent monitors.
The Chemistry Department was responsible for the calibration, source checks, functional checks, and daily observations of the radiological effluent and process monitors.
Since January 1988, the licensee had been trending the maintenance problems of the process and effluent monitors.
The data were reviewed by the Primary and Effluent Monitor Supervisor so that unnecessary Technical Specification Action Statements and Plant Work Orders (PWOs)
would be avoided.
Since the Instrumentation and Control Department responds to all plant PWOs, direct cost benefits would be derived by minimizing the unnecessary PWOs.
During the past 12 months, the license has had relatively few maintenance problems with the various process and effluent monitors.
The licensee indicated that central processing unit (CPU) failures were observed on the Steam Generator Blowdown Gaseous Effluent Monitor and the Emergency Core Cooling System (ECCS) monitor.
High humidity environments were considered to have been a factor in the CPU failures.
With regard to the Steam Generator Blowdown Gaseous Effluent Monitor, the licensee had in-place an
alternate sampling method to satisfy the Technical Specification requirements for radiation monitoring instrumentation.
During the plant tour, the inspector noted that the licensee had the capability to discharge radioactive liquid effluents at both the Unit 1 and Unit 2 discharge points.
However, since the initial startup of Unit 2, all radioactive liquid releases have been made from the Unit 1 side.
The Unit 2 liquid effluent monitor has been tagged out of service and inoperable since the initial Unit 2 startup.
The inspector noted that valve lineups and discharge lines from Unit 2 were aligned so that all the liquid waste would be diverted to the Unit 1 liquid waste processing system.
No violations or deviations were identified.
Air Cleaning Systems (84724)
Technical Specifications 3/4.7.7, 3/4.7.8, and 3/4.9.12 provide requirements for the testing of charcoal adsorber sample retention efficiency for methyl iodide and for in-place leak testing of HEPA filtration and charcoal adsorption sections of the Engineered Safety Feature (ESF) filtration systems.
ESF filtration systems at St. Lucie include the Units 1 and 2 Control Room Emergency Habitability System, the ECCS Area Ventilation System, and the Unit 1 Fuel Building Ventilation air treatment system.
Non-ESF filtration systems at St. Lucie are the Containment Recirculation system, the Hydrogen Purge system, and the Unit 2 Fuel Handling Building Ventilation system.
The Technical Support Center had no separate ventilation system since it was part of the Unit 1 Control Room Ventilation system envelope.
The inspector reviewed selected test records for the licensee/vendor laboratory tests of the methyl iodide and elemental iodine retention efficiency of charcoal adsorber samples for the Unit 2 ESF Systems.
The laboratory test conditions.~ere 955 relative humidity at 130 C.
The inspector also reviewed test records of the licensee/vendor in-place leak tests of the ESF and non-ESF filtration systems (testing occurred during October 1987).-
No violations or deviations were identified.
Licensee Action on Previously Identified Inspector Followup Items (IFIs)
(92701)
(Closed)
IFI 50-335/87-11-02 and 50-389/87-10-01:
Perform a time-motion and exposure study'for collection, transport and analyses of post-accident particulate and iodine samples from the plant vent (reference:
NUREG-0737, Table II.F.1-2).
The inspector reviewed the licensee's response to this IFI in a well-documented memo from R. J. Frechette (Chemistry Department)
to the St.
Lucie Technical Staff dated March 11,
1988.
The review also included an examinati on of Chemistry Procedures C-110, Initial Set of Post Accident Samples and Guidelines for Establishing Post Accident Water and Gas Inventory Control, Revision 5, November 5, 1986; and C-72, Processing Gaseous Wastes, Revision 22, Parch 18, 1987.
The licensee's evaluation included an estimate of the dose that a chemistry technician could receive under accident conditions while obtaining a plant vent grab sample.
The route time references were based on actual transit times as measured by the licensee.
The inspector reviewed with the licensee the assumptions that were made in.performing the evaluation.
The licensee calculated whole body and extremity doses that a technician would receive during retrieval and analysis of a
particulate and iodine sample assuming a design basis loss of coolant accident and source terms as described in the FSAR.
Total doses to the whole body for the entire evolution, assuming that one technician collected, transported,,
and analyzed the sample, would have been approximately 560 mrem and the dose to the extremities would have been approximately 605 mrem.
The licensee concluded that (1) plant vent stack radioiodine samples could be acquired under accident conditions without exceeding the dose limits specified in GDC-19; (2)
a guideline of approximately 2.75 R/hr should be set as the upper limit exposure rate for the sample cartridge (measured at one inch); any sample catridge reaching this dose rate should immediately be removed from service; and (3)
a 1,000 ml plant vent stack (PYS)
sample should provide an adequate quantity of radioiodines to perform a reliable analysis of PVS effluent.
This item is considered close /
TABLE 1 St. Lucie Nuclear Station (two units)
Semiannual Effluent Release Summar No. Abnormal Releases a.
Liquid b.
Gaseous Year 1985 1986 1987 Full Power Days (both units)
Liquid Waste Released (gallons)
Activity Released (curies)
a.
Liquid 1.
Fission and Activation Products 2.
Tritium 3.
Gross Alpha b.
Gaseous 1.
Noble Gas 2.
Halogens 3.
Tritium 4.
Gross Alpha Dose Estimate (mrem)
a.
Liquid (fish 5 shellfish)
Whole-body b.
Gaseous 1.
Whole-body 2.
Thyroid 5.00 E+06 5.51 E+00 6.50 E+02 9.02 E-05 6.03 E+04 9.80 E-01 5.62 E+02 1.55 E-07 7.92 E-01 1.87 E-02 6.60 E+00 7.79 E+07 4.96 E+00 5.56 E+02 3.10 E-04 4.33 E+04 3.11 E-01 8.11 E+01 1.06 E-06 1.11 E+00 1.32 E-02 6.67 E+00 NA 3.36 E+07 1.14 E+00 6.76 E+02 2.20 E-05 1.48 E+04 9.45 E-02 1.45 E+02 6.39 E-07 6.95 E-02 5.08 E-03 1.80 E+00 NA = Not avai a
e