IR 05000335/1995019
| ML17228B343 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 11/15/1995 |
| From: | Robert Carrion, Decker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17228B342 | List: |
| References | |
| 50-335-95-19, 50-389-95-19, NUDOCS 9512080171 | |
| Download: ML17228B343 (28) | |
Text
~8 AEgy (4 P
~4
Cy C0
+~
~0
++*++
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900 ATLANTA,GEORGIA 303234199 Report Nos:
50-335/95-19 and 50-389/95-19 Licensee:
Florida Power and Light Company 9250 West Flagler Street Miami, FL 33102 Docket Nos.:
50-335 and 50-389 License Nos.:
DPR-67 and NPF-16 Facility Name:
St.
Lucie 1 and
Inspection Conducted:
October 16 - 20, 1995 Inspector:
R.
. Carrion, Radiation Specialist
/Pi&~'
Date Signed Approved b
ecker, Acting Branch Chief Plant upport Branch Division of Reactor Safety SUMMARY Il i/
D te Signed Scope:
This routine, announced inspection was conducted in the ar'eas of the organization of the Chemistry Department and Radwaste Group, plant water chemistry, the Radiological Environmental Monitoring Program (REMP), the Post Accident Sampling System (PASS), the Control Room Emergency Ventilation System, radioactive materials handling and transportation documentation, volume reduction of solid radwaste, the status of the Refueling Water Tank (RWT) leak migration, contaminated sludge disposal, and the leaking Unit 2 Spent Fuel Pool (SFP)
Ion Exchanger.
Results:
The licensee's organization of its Chemistry Department and Radwaste Group satisfied Technical Specification (TS) requirements.
(Paragraph 2)
The licensee's plant water chemistry program continued to be effectively implemented.
(Paragraph 3)
The licensee had an effective program in place to analyze radiological effluents, direct radiation, etc.
due to plant operations, as evidenced by the Radiological Environmental Operating Report.
(Paragraph 4)
Enclosure 9512080171 951115 PDR ADOCK 05000335 Q
The licensee's PASS was capable of fulfillingits intended sampling function and the technicians had maintained.their capability to operate the system.
(Paragraph 5)
The Control Room Emergency Ventilation System was adequate for its intended function and was being maintained in compliance with the applicable TSs.
(Paragraph 6)
The licensee's radwaste processing and shipping was conducted in a competent, professional manner and the radwaste shipping documentation was thorough and in compliance with the applicable regulations.
(Paragraph 7)
The licensee continued to make good progress in the reduction of its solid radwaste.
(Paragraph 8)
The licensee continued to monitor isotope migration due to the RWT leak.
(Paragraph 9)
The licensee had acted prudently on the issue of contaminated sewage sludge disposal.
(Paragraph 10)
The licensee had taken a proactive position in the resolution of the leaking Unit 2 Spent Fuel Pool (SFP)
Ion Exchanger.
(Paragraph ll)
Enclosure
REPORT DETAILS 1.
Persons Contacted Licensee Employees
- K. E. Beichel, Acting Chemistry Supervisor
- E. J.
Benken, Licensing Engineer
- H. F. Buchanan, Health Physics (HP) Supervisor
- R.
E.
Cox, Chemistry Effluents Supervisor D.
H. Faulkner, Primary Chemistry Supervisor R. J. Frechette, Chemistry Supervisor
- D. Haithcox, HP (Radwaste)
- D. A. Sager, Site Vice President
- R. B. Somers, HP (Radwaste)
- J. A. West, Operations Manager D. Wooldridge, Chemistry Instructor Other licensee employees contacted during this inspection'ncluded technicians and administrative personnel.
Nuclear Regulatory Commission (NRC)
H. S. Hiller, Resident Inspector
- R. C. Prevatte, Senior Resident Inspector S. Sandin, Resident Inspector
- Attended exit interview Acronyms and Initialisms used throughout this report are listed in the last paragraph.
2.
Organization (84750 and 86750)
Technical Specification (TS) 6.2 describes the licensee's organization.
The inspector reviewed the licensee's organization, staffing levels, and lines of authority as they related to the Chemistry Department and Radioactive Waste Group to verify that the licensee had not made organizational changes which would adversely affect the ability to control radiation exposures or radioactive material.
There had been no structural changes in the Chemistry Department since the previous inspection.
However, there had been some discussion about removing the responsibility of the radiation monitors from the Chemistry.
Department to the Instrumentation and Controls (ISC) Department, along with two technicians.
Otherwise, the Chemistry Department had remained unchanged from the last time it was reviewed.
There had been no changes in the Radwaste Group since the last time this area was reviewed.
(Refer to Inspection Report (IR) 50-335, 389/95-06, Paragraph 2.)
Enclosure
The inspector concluded that the licensee's organization in the areas of Chemistry and Radioactive Waste satisfied the requirements of the,TS.
/
No violations or deviations were identified.
3.
Plant Water Chemistry (84750)
During the inspection, Unit 1 was operating in its twelfth fuel cycle and Unit 2 was in its eighth refueling outage, which was scheduled to end in late November.
The next Unit 1 refueling outage was scheduled for early Spring 1996.
a.
Primary Water Chemistry 1.
TS-Required Parameters The inspector reviewed the plant chemistry controls and operational controls affecting primary plant water chemistry since the last inspection in this area.
TS 3.4.7 specifies that the concentrations of dissolved oxygen (DO), chloride, and fluoride in the Reactor Coolant System (RCS)
be maintained below 0. 10 parts per million (ppm), 0. 15 ppm, and 0. 10 ppm, respectively.
TS 3.4.8 specifies that the specific activity of the primary coolant be limited to less than or equal to 1.0 microcuries/gram (pCi/g) dose equivalent iodine (DEI).
2.
Pursuant to these requirements, the inspector reviewed daily summaries for both units which correlated reactor power output to chloride, fluoride, and dissolved oxygen concentrations, and specific activity of the reactor coolant.
For both Units 1 and 2, the arbitrarily-chosen period of June 1 through July 31, 1995, was reviewed and the parameters were determined to have been maintained well below TS limits.
Typical values for DO, chloride, and fluoride were less than five parts per billion (ppb),
seven ppb, and ten ppb, respectively, for Unit 1 and less than five ppb, four ppb, and less than four ppb, respectively, for Unit 2.
Typical DEI values at steady-state conditions ranged from 4.0E-2 pCi/g to.2.7E-3 pCi/g for Unit 1 and from 7.2E-3 pCi/g to 3.4E-3 pCi/g for Unit 2.
Unit 1 was suspected of haviAg some leaking fuel, in a third-cycle bundle, which would be replaced during the next refueling outage.
Unit 2 had shown no evidence of leaking fuel during the last fuel cycle.
Early Boration During the current inspection, the licensee was in the final stages of its early boration regimen (acid-reducing chemistry)
combined with hydrogen peroxide injection (acid-oxidizing chemistry)
as it shutdown Unit 2 for the refueling outage.
The process solubilizes Co-'58 and Co-60, which are removed via the demineralizers
~
This is done to reduce the source term and potential exposure to the outage workers.
Enclosure
The inspector reviewed a graph relating isotopic concentrations to reactor power/temperature for the duration of the shutdown to date.
The greatest isotopic concentrations were of Co-58 and Co-60.
The ratio of the Co-58 to Co-60 remained relatively constant throughout the process at about 200 to 1, with a maximum of 1.5 pCi/ml for the Co-58 concentration.
Additional isotopes identified included Cr-51, Cs-134, and DEI.
The licensee had not yet calculated the number of curies removed from the system during the course of the inspection.
Based on the results of this graphical data, the inspector concluded that the licensee was proactive in trying to reduce dose rates by removing significant quantities of activity via its early 'boration/hydrogen peroxide shutdown program.
The inspector concluded that the Primary Plant Water Chemistry was maintained well within the TS requirements.
b.
Secondary Water Chemistry TS 6.8.4.c requires the licensee to establish, implement, maintain, and audit a Secondary Water Chemistry Program to inhibit steam generator (SG) tube degradation.
The inspector discussed the licensee's program to inhibit SG.tube degradation with cognizant licensee personnel.
The licensee was evaluating a technique known in the industry as molar ratio control.
Its object is to avoid the production of either acidic or caustic solutions in limited-flow areas where impurities concentrate in SGs.
(The technical basis for this approach is the realization that most
.
corrosion reactions tend to be slower under neutral conditions than under either acidic or caustic conditions.)
This technique was recommended in the EPRI PWR Secondary Water Chemistry'Guidelines, Rev.
3, as part of a plant-specific secondary water chemistry control program.
Furthermore, it recommends monitoring for a'simple molar ratio of sodium to chloride as well as for a more complex ratio of sodium to a weighted sum of chloride plus sulfate.
The ratio could be adjusted to mitigate Intergranular Attack/Stress Corrosion Cracking (IGA/SCC).
The licensee's SG vendor recommends that the licensee's focus of operational chemistry be on the simple sodium-to-chloride ratio, based upon the following reasons:
1) evidence suggests that sulfate adsorbs onto the general surface areas within the SGs rather than concentrating (along with other impurities) within crevices and low-flow areas, and 2) in the vendor's proprietary calculations, based on high-temperature thermodynamic data, sulfate is rarely present in concentrated crevice chemistry solutions at concentrations comparable to those of chloride.
Because the industry is still in the process of collecting data and developing specific recommendations to be included in a ratio control program, the position of the Chemistry Department was to continue to monitor and trend the sodium/chloride ratio until the industry and site data base was sufficiently developed to form the basis for an implementation recommendation.
Previous testing indicated that the SGs be operated at a r'ecommended ratio of between Enclosure
I
0.3 to 0.7.
Furthermore, if corrective action is needed to return the ratio to the recommended range, it can usually be accomplished by adjusting blowdown flow or by adjusting the percentage of cation to anion resin in the condensate polisher vessels such that appropriate changes in the quantities of sodium and/or chloride transported to the SGs are made.
The licensee was collecting data at each planned shutdown to provide ratio comparisons for daily chemistry.
This data was expected to be instrumental in the future development of the licensee's molar ratio management philosophy.
The inspector concluded that the licensee was taking a cautious approach to the implementation of molar ratio control. until additional hard data/industrial experience could maximize its chances of successful implementation.
Otherwise, the licensee had implemented an effective Secondary Water Chemistry Program.
No violations or deviations were identified.
4.
Radiological Environmental Monitoring Program (REHP)
(84750)
Sections 3/4. 12. 1 of the licensee's Off-site Dose Calculation Manual (ODCH) specify that the licensee shall conduct a Radiological Environmental Monitoring Program in accordance with TS 6.8.4.g.
1 (as specified in ODCH Table 3. 12-1) to monitor radiation and radionuclides in the environs of the plant and define how the program shall be conducted.
The REMP shall provide representative measurements of radioactivity in the highest potential exposure pathways and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.
Accumulation of radioactivity in the environment can thereby be measured and trends can be assessed to determine whether the radioactivity resulted from plant operations.
The data may also be used to project the potential dose to offsite populations based on the cumulative measurements of any plant-originated radioactivity, as well as to detect unanticipated pathways for the transport of radionuclides through the environment.
The St. Lucie Nuclear Plant Environmental Monitoring Program is designed to detect the effects, if any, of plant operation on environmental radiation levels by monitoring airborne, waterborne, ingestion, and direct radiation pathways in the area surrounding the plant site.
It also verifies that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways.
Indicator sampling stations are located where detection of the radiological effects of the plant's operation would be most likely, where the samples collected should provide a significant indication of potential dose to man, and where an adequate comparison of predicted radiological levels might be made with measured levels.
Control stations are located where radiological levels are not expected to be significantly influenced by plant operation, i.e., at background locations.
An environmental impact assessment of plant operation is made'from the radiological measurements of the sampling stations.
Enclosure
The RENP was conducted at the St.
Lucie Plant by the Office of Radiological Control, Florida Department of Health and Rehabilitative Services (DHRS).,
Samples were collected and analyzed by DHRS personnel at the DHRS Environmental Radiation Control Laboratory in Orlando, Florida.'he REMP was discussed with the Florida,DHRS Health Physicist.
a ~
b.
Observation of Sample Collection The inspector accompanied the Health Physicist on part of her normal weekly rounds to collect samples to observe collection technique and to check the physical condition and operability of the sampling stations.
Air samples were taken at two plant indicator stations, H-08 and H-14, the plant control station, H-12, and one state air'onitoring station, H-09.
All air sampling stations included thermoluminescent dosimeters (TLDs) for the detection of direct gamma radiation.
'The DHRS Health Physicist pointed out additional TLD locations as well. The air sampling stations were located such that there would be no interference from tall weeds/vegetation in taking representative samples.
The inspector noted that the sampling units were well-maintained, within calibration, in good working order, and that there was no evidence of vandalism.
The inspector observed that the samples were properly collected and that the technician'used good HP techniques to avoid sample contamination and conducted her work in an efficient, competent manner.
1994 Annual Radiological Environmental Operating Report TS 6.9. 1.8 requires that the Annual Report be submitted prior to Hay
. of the following year.
TS 6.9. 1.8 also states format and content requirements for the Report..
The inspector reviewed the Annual Radiological Environmental Operating Report for calendar year 1994 to verify compliance with the TSs.
The Report had been submitted in compliance with TS 6.9.1.8 on April 19, 1995, and the format and contents were as. prescribed by the TS.
There were no changes to the environmental monitoring network during 1994.
For the 1994 calendar year, 814 samples were taken and analyzed.
Analytical results were divided into four categories based on exposure pathways:
Airborne, waterborne, ingestion, and direct radiation.
Each of the pathways was described as follows:
The airborne exposure included airborne iodine and airborne particulate samples.
No fission products or other man-made isotopes in the airborne particulate media were detected in 1994.
Overall, 1994 airborne results were very similar to those of previous years and preoperational levels.
The waterborne exposure pathway included surface water samples and shoreline sediment samples.
Tritium was the only man-made isotope detected in the surface water or shoreline sediment at collection station H-15.
Activity levels in the surface water samples were consistent with past measurements.
Tritium was reported in two (of sixty-four) samples.
The highest measured Enclosure
tritium activity was less than one percent of the reporting level defined in Table 3. 12-2 of the ODCH.
The shoreline sediment contained no detectable levels of any man-made isotopes, but the naturally-occurring K-40, Ra-226, and Th-232 were detected.
The ingestion exposure pathway included fish and crustacea, and.
broad-leaf vegetation samples.
Only naturally-occurring K-40 was detected at normal environmental levels for fish and crustacea.
Vegetation samples yielded concentrations of radioisotopes which were similar to those of the control sampling stations and of the preoperational period.
Only naturally-occurring K-40, Pb-210, and Be-7 were detected.
The environmental direct radiation exposure pathway was measured by use of TLDs.
TLD results for 1994 remained consistent with
'hose of previous years, i.e. essentially unchanged since the preoperational period.
The report showed that the program was conducted in accordance with requirements and that supplemental sampling and analyses were performed.
The radiological environmental data indicated that the levels of radiation and concentrations of radioactive materials in environmental samples (representing the highest potential exposure pathways to members of the public) were not increasing.
Therefore, plant operations had no significant radiological impact on the environment or public health and safety.
The maximum radiation dose from airborne, waterborne, ingestion, or direct exposure pathways attributed to plant operations in 1994 to any offsite member of the.
public was well within the criteria established by 40 CFR 190.
c.
Analytical Comparison of 1994 Report The NRC contracts with the Radiological and Environmental Sciences Laboratory (RESL) to analyze samples split between the State of Florida and the NRC.
The NRC compares the RESL results to those of the State of Florida for analysis confirmation.
The inspector randomly selected the analytical results for eight gross beta air particulate filter split samples from Sample Station H-14 (specifically, the four samples collected in February and the four samples collected in Hay)
and a shoreline sediment sample from Sample Station H-15, collected August 18, 1994, for'comparison of the results.
After adjusting for the different units used by the different laboratories to report the results, the inspector determined
'hat the reported results compared favorably.
Typical values for.
gross beta in the air particulates were reported by the licensee to be
, 0.015 pCi/m'.
The value of Cs-137 was less than 12 picocuries per kilogram (pCi/kg)
and the value of K-40 was 1217 pCi/kg.
The inspector concluded that the results of analyses of environmental samples by the State of Florida compared favorably to those of RESL, which served as an independent verification of the results.
Enclosure
'0
The inspector concluded that the licensee had an effective program in place to monitor radiological effluents, direct radiation, etc.
due to plant operations and that the Report was in compliance with the TSs.
In 1994, plant operations caused minimum impact to the environment and virtually no dose to the general public from those effluents.
No violations or deviations were identified.
5.
Post Accident Sampling System (PASS)
(84750)
Section II.B.3 of NUREG-0737 requires that the licensee be able to obtain a sample of the reactor coolant and containment atmosphere.
Furthermore, the sample must be promptly obtained and analyzed (within three hours total) under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18 3/4 rem to the whole body and/or extremities, respectively.
TS 6.8.4.e requires that a program be established, implemented, and maintained to ensure the capability to obtain and analyze, under accident conditions, reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples.
The PASS should provide these capabilities and should enable the licensee to obtain information critical to the efforts to assess and control the course and effects of an accident.
a.
PASS Operability The inspector reviewed the two most recent PASS operability log sheets for both units and discussed the results with the Primary Chemistry Supervisor.
The operability tests had been performed within the required six-month time limits, on December 14, 1994 and April 6, 1995 for Unit
and December 8,
1994 and Nay 22, 1995 for Unit 2.
A comparison of nine parameters (boron; dissolved hydrogen; gross activity; I-131, I-132, I-133, I-134, and I-135 activities; and DEI).
of the daily-analyzed RCS sample to the readings taken from the PASS satisfied the acceptance criteria of both units.
b.
PASS Panel Walkdown The inspector walked down the PASS Panel of both Units to observe their state of maintenance and operability.
The panels were
"mimicked" to help the operator understand the flow paths available by various valve configurations.
The Unit 1 system was simpler than that of Unit 2, but both'ere found to be well-maintained and no Deficiency Tags, Work Orders, etc.
were observed on either panel.
c.
Training The inspector reviewed the training of the chemistry technicians as it related to the PASS.
From discussions with the Chemistry Instructor, due to the relative infrequency of system operation, PASS requalification training was required on a biennial basis for the system of each unit.
Until 1993, technicians had been required to Enclosure
requalify on both systems each year.
Since then, the technicians had been required to requalify on each system every other year.
The requalification process consisted of the technicians reading a
self-study module, which referred to the appropriate chemistry procedure and system operation manual, for the respective system and demonstrating their understanding of that system by taking a sample and, in the case of the Unit 2 system, passing a written examination.
The inspector reviewed Training Module 2101114,
"Unit 1 PASS," for objectives, performance guidance, and general informational guidance and determined that it was thorough and included general information, enabling objectives, performance guidance, a self-test, etc.
The inspector also reviewed Instructor.Guide 2103114,
"Unit 1 PASS,"
and determined that it was consistent with the training module; i.e., the oral questions to be asked of the technician were specifically taken from the training module.
The guide also specified actions to be taken by the instructor during and after the technician's performance demonstration, depending upon whether or not the demonstration was satisfactory.
The inspector also reviewed a written examination for the Unit 2 PASS.
It consisted of multiple-choice, true-false, fill-in-the-blank, and short-answer questions as well as calculations.
The test was computer-generated from a bank of questions to assure its
~ integrity.
The test was determined to be,a good test of the theory, operational c'apabilities, and system valve configurations.
The inspector reviewed personnel records of Chemistry Department technicians and found them to be in order.
The inspector concluded that the PASS was capable of fulfillingits intended sampling function and that the technicians had maintained their capability to operate the system.
No violations or deviations were identified.
6.
Control Room Emergency Ventilation System (84750)
Per
CFR 50, Appendix A, Criterion 19', licensees'shall assure that adequate radiation protection be provided to permit access to and occupancy of the control room under accident conditions and for the duration of the accident.
Specifically, operability of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room remains habitable for operations personnel during and following all credible accident conditions such that the radiation exposure to personnel occupying the control room is limited to 5 rems or less whole body, or its equivalent.
Section 6.4 of the Final Safety Analysis Report (FSAR) discusses Habitability Systems in general, including the Control Room Air.
Conditioning System.
Section 9.4. 1 of the FSAR discusses the Control Room Air Conditioning System in particular, including the design bases, system description (under both normal and emergency operations),
safety evaluation, inspection and testing requirements, and instrument requirements.
Enclosure
TS 3.7.7 defines operability, requirements for the control room emergency air cleanup systems under the various design scenarios.
TS 4.7.7 sets the surveillance requirements for the systems.
The inspector reviewed the last two surveillance results for the charcoal adsorbers and High Efficiency Particulate Air (HEPA) filters to verify compliance with TS requirements.
The first surveillance reviewed was conducted on February 8, 1994 and the filters satisfied the acceptance criteri'a and no irregularities were noted.
The most recent surveillance reviewed was conducted on October 3, 1995, and the results of the laboratory analysis of the charcoal adsorber efficiency had not yet been completed and returned to the licensee by the vendor at the time of this inspection.
No irregularities were noted in the rest of the surveillance.
Furthermore, the inspector noted that the surveillances had been conducted within the TS-required frequency.'he inspector also reviewed the periodic test for positive pressure in the Control Room of the Unit 2 system to verify compliance with TS requirements.
The first was conducted on September 27, 1993, and the most recent
'on Parch 27, 1995.
Both periodic tests satisfied the acceptance criteria and no irregularities were noted.
The inspector noted that the surveillances had been conducted within the TS-required frequency.
The inspector reviewed Figure 9.4-2,
"Control Room A/C System PAID Diagram," for Unit 2.
The inspector walked down the system, from the air intake to the Control Room, to air exhaust, noting the major components, such as isolation dampers, filter banks, and fans as well as detectors for smoke, radiation, etc.
All components were well maintained, with no sign of physical degradation.
The inspector discussed system operation under.
both normal and emergency conditions with cognizant licensee personnel.
Based on the scope of this review, the inspector concluded that the System was adequate for its intended function and that it was being maintained in compliance with the applicable TSs.
No violations or deviations were identified.
Radioactive Material Processing and Transportation (86750)
CFR 71.5 (a) requires that each licensee who transfers licensed material outside of the confines of its plant or other place of use, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the regulations appropriate to the mode of transport of the Department of Transportation (DOT) in 49 CFR, Parts 170 through 189.
CFR 71 Subpart H establishes the quality assurance (gA) program requirements applicable to transportation of radioactive materials.
CFR 20.2006 and Appendix F to
CFR 20 specify the requirements for control of transfers of radioactive waste intended for disposal at a land disposal facility and for establishing a manifest tracking system for those transfers.
CFR 61.55 and 61.56 establish the requirements for classification and characterization of radioactive waste shipped to a near-surface disposal site.
Enclosur'e
Pursuant to these requirements, the inspector reviewed the licensee's activities affiliated with these requirements, to determine whether the
. licensee effectively packages, stores, and ships radioactive solid materials.
The licensee's program for the packaging and transportation of radioactive materials, including solid radwaste, was conducted by the Radioactive Waste Group within the Health Physics Department.
Radwaste was processed and packaged (including the preparation of shipping documentation)
by 'the Radwaste Group, with the assistance of Radiation Protection Men (RPM)
on loan from the Health Physics Operations Department to complete specific
.
tasks, such as loading a shipment or compacting contaminated material.
a.
Quality Assurance (QA) Program
CFR 71. 101(c) requires the licensee to obtain NRC approval of the QA program prior to the use of any package for shipment of licensed mat'erial subject to
CFR 71 Subpart,H.
The inspector determined that the licensee's QA program had been approved by the Nuclear Regulatory Commission (NRC) Office of Nuclear Material Safety and Safeguards upon its issuing of "Quality Assurance Program Approval for Radioactive Material Packages No. 0169," Revision 5 on August 10, 1994.
b.
NRC Certificate of Compliance Packaging
CFR 71. 12(c)(1) requires the licensee to maintain copies of certificates of compliance (C of C) for NRC approved packages used for transport of radioactive material.
The inspector verified that the licensee possessed a current copy of the C of C for Package No. 0345, a shielded cask (USA/9176/A)
routinely used for shipments of radioactive material of low specific activity (LSA).
c.
Registration of Use of NRC Approved Packages
CFR 71. 12(c)(3) requires the licensee to submit to the NRC, prior to the first use of an NRC approved package, the licensee's name and license number and the package identification number specified in the package approval.
The inspector reviewed an NRC-issued letter dated May 20, 1993, which indicated that the licensee was a registered user of the above cask used for shipments of Low Specific Activity (LSA) material.
The inspector also verified that the licensee was a registered user.for other types of containers and maintained C of Cs for those containers.
d.
Radioactive Material Shipping Documentation Packages Per
CFR 172.200,. the licensee is required to prepare.shipping papers describing hazardous materials offered for transport in the Enclosure
manner specified in 49 CFR 172 Subpart C.
Also, per
CFR 20.2006, the licensee is required to prepare shipping manifests for each shipment of radioactive waste to a licensed land disposal facility-such that they meet the requirements of Appendix F to
CFR '20.
Per
CFR 20.2006(d)
and Section III.A.1 of Appendix F to 10 CFR 20, the licensee is required to prepare all radioactive waste shipped to a
licensed land disposal facility or waste collector such that the waste is classified according to 10 CFR 61.55 and meets the waste characteristics requirements in 10 CFR 61.56.
Per
CFR 173.425(b)(9)
and 173.441(c),
the licensee is required to provide specific written instructions for maintenance of the exclusive use shipment controls to the carrier of packages of radioactive material consigned as exclusive use.
Those instructions were required to be included with the shipping paper information.
Per
CFR 71.47,
CFR 71.87(i)
and (j), 49 CFR 173.441,
CFR 173.443 and
CFR 173.475(i), the limits for external radiation levels and for removable surface contamination levels of packages offered for shipment are delineated.
The inspector reviewed selected portions of the following Health Physics Procedures:
HP-40, Rev.
39,
"Shipment of Radiological Materials," approved July 13, 1995; HP-47, Rev.
16, "Classification of Radioactive Haterial," approved January 26, 1995; and HP-48, Rev. 3,
"Activity Determined from Dose Rate from Containers,"
approved December 7,
1989, as related to the shipping of radioactive materials and associated documentation.
The procedures collectively.addressed the requirements.
The inspector determined that the licensee's procedures for shipping radioactive materials included provisions for.
performing the required surveys and for assuring that the radiation and contamination limits were met for each package offered for shipment.
The inspector also determined that they included provisions for performing the required surveys and for assuring that the radiation and contamination limits were met for each package offered for shipment.
In addition, the inspector determined that the procedures included provisions for providing drivers of exclusive use shipments with the required instructions.
To verify that the procedures were being followed, the inspector reviewed three arbitrarily-chosen documentation packages:
Radioactive Haterial Shipment (RNS)
No. 95-25, a LSA, Type A,shipment, destined for processing (incineration and/or compaction)
before final disposal; RNS No. 95-25, an Excepted Package containing a Pressurizer Code Safety Valve for testing in a laboratory; and RNS No. 95-47, dewatered bead resin destined for the disposal facility.
The documentation packages contained thorough documentation about the respective shipments and the above-referenced items.
The radiation and contamination survey results were within the
CFR requirements and the shipping documents were being maintained as required.
The documentation packages includhd a copy of the instructions provided to the drivers, with respect to the exclusive use status of their Enclosure
shipment and emergency information.
The inspector also determined from the reviewed shipping records that the licensee classified and characterized waste shipments through the use of the WASTRAK computer software.
The inspector concluded that the shipping papers for the selected shipments of radioactive materials had been prepared in accordance with the above referenced procedures and satisfied regulatory requirements.
e.
Observation of Work in, Progress The inspector observed some of the activities associated with the receipt of equipment to be used during the current Unit 2 outage, including its unloading from the trailer and storage in the Blowdown Treatment Building to evaluate the effectiveness of training, adequacy of procedures, etc.
The work proceeded well; each member of the work detail handled his responsibilities in an efficient, professional manner.
The personnel were conscious of potential radiological material hazards, as evidenced by the use of radiation dose meters, smear samples, etc.
The work was done in accordance with Health Physics Procedures HPP-80, Rev.
1, "Receipt of Radioactive Material" and HPP-41, Rev.
1,
"Movement of Material and Equipment."
Control of Material in the Radiation Control Area (RCA)
The inspector discussed the process used to remove material from the RCA with the Radwaste Supervisor.
The licensee removes material from the RCA by HPP-41, which specifies the criteria by which material may be released from the RCA.
Per procedure, in order to release material (or a vehicle) from the RCA, that material shall have no detectable activity, as determined by direct frisks or smear surveys done by HP personnel.
Material determined to be clean then may be released from the RCA with authorization from HP.
All material surveyed and released through designated control points is entered into the
"Unconditional Release Log" and the log entry is signed by HP personnel authorizing the release.
The log is later inspected by HP supervisory personnel for proper signatures.
.
The inspector went to one of the designated control points (Gate 3)
and discussed the procedure with the technici an on duty at the time.
The inspector was shown the area where material from ghe RCA was placed prior to"being frisked/surveyed, where and how the frisks were done, and where the clean material was placed subsequent to frisking.
Each area was properly posted and identified the material held within.
Periodically, the clean material in the holding area was collected, removed from the RCA, and disposed of as normal trash.
Later, the inspector borrowed a detector from the licensee and surveyed two dumpsters outside the RCA which contained material which had been surveyed and released as clean.
No evidence of radiologicall'y-contaminated material was identified.
Enclosure
Based on observations in the work area, discussions with cognizant licensee personnel, review of the procedures used for this work, and an independent survey of released material, the inspector concluded that the licensee had implemented an effective program to control.the release of material from the RCA within regulatory limits.
g.
Training Program for Transportation of Radioactive Material Per
CFR 71.105(d)
the licensee is required to provide for
.indoctrination and training of personnel performing activities affecting quality as'ecessary to assure that suitable proficiency is achieved and maintained.
Also, per 49 CFR 172.702 any hazmat employer is required to ensure that each of it's hazmat employees are trained in accordance with the requirements prescribed in 49 CFR 172 Subpart H.
Furthermore,
CFR 172.704(c) requires each hazmat employer to provide each hazmat employee with initial training and recurrent training at least once every two years.
r The inspector reviewed training certificates for the HP supervisors responsible for shipping/transportation activities and determined that their required training was current and that the licensee's program for recurrent training of shipping personnel was being conducted on a
frequency of at least once every two years.
During conversations with licensee personnel involved in preparation of radioactive material for transport, the inspector noted that those personnel were well versed in the applicable NRC and DOT requirements and concluded, therefore, that the licensee had an effective training program in place for personnel involved in the preparation of radioactive material for transport.
Based on the above reviews, the inspector concluded that the licensee had implemented effective gA and management control programs for the handling, packaging, and transport of radioactive material (including the required paper documentation)
and that regulatory requirements
'were satisfied.
No violations or deviations were identified.
8.-
Solid Radwaste Volume Reduction (86750)
The licensee had continued its review into all aspects of solid radwaste in an effort to reduce its disposal volume.
(Refer to Inspection Reports 50-335, 389/95-06, Paragraph ll.b.)
Special emphasis had been placed on reducing its generation, especially of the use of plastic materials and the use of washable materials in place of disposables for everything from protective clothing and shoe covers to tool bags. Additionally, procedural changes and work processes were under review.
The licensee had also continued its evaluation of a water-dissolvable material for use in, protective clothing, shoe covers, mop heads, etc.
and had begun using this material onsite on a trial basis.
Furthermore, the licensee was evaluating a volume reduction technique, currently used to process, biohazards, which involved a proprietary incandescence technique.
Enclosure
The inspector concluded that the licensee was making a good effort to reduce its solid radwaste.
No violations or deviations were identified.
9.
Refueling Water Tank (RWT) Leak Status (84750)
The RWT leak was addressed in IRs 50-335, 389/93-17; 50-335, 389/93-27; 50-335, 389/94-16; and 50-335, 389/95-06 which described the circumstances surrounding the leak, measures taken by the licensee, planned actions to monitor the released material, and results of the monitoring program.
The licensee had treated the event as an Unplanned Release and had incorporated this information into the Semiannual Radioactive Effluent Release Report for the second half of 1993 and the Annual Radioactive Effluent Release Report for 1994.
The licensee had continued to monitor the migration of the tritium released through a system of twenty monitoring,wells and four recovery wells, originally established to monitor and recover diesel product from a nearby diesel fuel tank leak.
The inspector went to the RWT location to review the general area, including nearby structures and monitoring wells, and noted that part of the area was being used as a laydown area for Unit
'2 outage equipment and materials.
The inspector also reviewed results of analyses from quarterly well samples taken in June and October 1995.
For the three wells nearest the RWT (Monitoring Wells Nos.
4 and 18D and Recovery Well No. 2), the tritium concentrations had dropped relatively
~
continuously for Monitor Well No.
4 (to less than its Minimum Detectable Activity (MDA) in October)
since the initial leak.
Monitor Well No.
18D and Recovery Well No.
2 had shown erratic but generally decreasing concentrations (to 1.48E-4 pCi/ml in October and 2.33E-4 pCi/ml in June, respectively) of tritium.
In wells located more distant from the RWT, tritium activities were reported to be less than the HDA in all but two cases.
The results of Monitor Well Nos.
6 and 14 showed a generally gradually declining trend (to 5.31E-5 pCi/ml in October) for Monitor Well No.
6 and a relatively constant concentration (1.82E-5 pCi/ml in October'or Monitor Well No.
14.
The inspector concluded that the licensee was adequately monitoring the migration of the release and that the public health and safety was not adversely affected by this release.
No violations or deviations were identified.
10. Contaminated Sewage Sludge Disposal (84750)
The issue of contaminated sludge and its disposal was addressed in Paragraph 12 of IR 50-335, 389/93-17; Paragraph 11 of IR 50-335, 389/93-17; Paragraph 10 of IR 50-335, 389/94-16, and Paragraph 12 of IR 50-335, 389/95-06.
Since the last inspection (95-06), the licensee had continued shipping the material to the Gifford Regional Waste Treatment Facility for normal treatment and processing, as approved by the Florida Department of Health and Rehabilitative Services.
Specifically, four Enclosure
l
additional shipments had been made (for a total of seven).
The last shipment was made on October 13, 1995 and contained approximately 4000 gallons of material.
Sampling done by the State prior to leaving the site determined the maximum concentration of Co-.60 to be 0. 114 pCi/g.
The licensee planned to continue to dispose of its contaminated sludge in this manner until a nearby county water treatment facility is completed.
The licensee planned to pipe the material directly from the plant to the treatment facility, starting in the first half of 1997.
The inspector concluded that the licensee had proceeded in a prudent manner on this issue.
'No violations or deviations were identified.
11. Status of Leaking Unit 2 Spent Fuel Pool (SFP)
Ion Exchanger (84750)
The inspector followed up on the status of the Unit 2 SFP ion exchanger which had developed a leak during a resin backflushing operation, permitting an estimated twenty-five cubic feet of resin to become distributed within various Unit 2 systems, primarily the SPF and the Unit 2 Refueling Water Tank (RWT) but also including the RCS and the Emergency Core Cooling Systems (ECCS), the High Pressure Safety Injection (HPSI)
and Low Pressure Safety Injection (LPSI) Systems.
(Refer to IRs 50-335, 389/95-06, Paragraph 7.)
The RWT and the ECCS had been cleaned of resin and their sulfate levels have been reduced to pre-event levels and the clean up of the SFP was continuing at the time of the previous inspection and was expected to continue for several weeks more.
The. sulfate concentration in the SFP had been reduced from about 6 ppm to about 1.8 ppm during the week of the previous inspection.
During the current inspection, the inspector discussed the progress in resolving the event with cognizant licensee 'representatives.
The SFP clean up had been completed, including a guality Assurance check of randomly-selected individual fuel assemblies for resin which may not have been recovered.
The engineering evaluation commissioned by the licensee to assess the effects of sulfate concentrations with respect to the corrosion potential of the SFP liner, fuel storage racks, spent fuel assemblies, etc.
as well as the effects of sulfates on RCS components based on the possibility of resin transport from the SFP to the RCS had been finalized.
Its recommendations had been incorporated into the Startup Procedure, via Chemistry Department'etter of Instruction No. LOI-CC-15, for the heat up and startup of Unit 2 upon the completion of the present refueling outage.
The Letter of Instruction specified sampling regimens,'old points, action levels, etc. to assure that any remaining resin material would be broken down and removed via the Chemical and Volume Control System (CVCS).
In addition, as part of lessons learned resulting from its root cause analysis, the licensee had revised Operating Procedure No. 2-0520020,
"Radioactive Resin Replacement,"
to flush the ion exchanger outlet strainers with flow in the normal direction of travel and to avoid the rapid opening and closing of the Primary Makeup Water (PHW)
and ion exchange discharge valves to preclude damage to the internals of the ion exchange vessel.
Enclosure
12.
The inspector concluded that the licensee had taken a proactive position in the resolution of this event.
I No violations or deviations were identified.
Exit Interview (84750 and 86750)
The inspection scope and results were summarized on October 20, 1995, with those persons indicated in Paragraph 1.
The inspectors described the areas inspected and discussed the inspection results, including likely informational content of the inspection report with regard to documents and/or processes reviewed during the inspection.
The licensee did not identify any such documents or processes as proprietary, with the exception of the process referenced in Paragraph 8.
Dissenting comments were not received from the licensee.
13.
Acronyms CFR CiCofC-CVCS DEI DHRS DO DOT ECCS EPRI FPL FSAR g
NRC ODCM PASS pCi PMW ppb ppm PSL and Initialisms Code of Federal Regulations curie Certificate of Compliance Chemical and Volume Control System Dose Equivalent Iodine Department of Health and Rehabilitative Control Dissolved Oxygen Department of Transportation Emergency Core Cooling Systems Electrical Power Research Institute Florida Power and Light Final Safety Analysis Report gram High Efficiency Particulate Air Health Physics High Pressure Safety Injection Instrumentation and,Controls Intergranular Attack Inspection Report kilogram liter Low Pressure Safety Injection Low;Specific Activity micro-Curie (1.0E-6 Ci)
Minimum Detectable Activity milli-liter Number Nuclear Regulatory Commission Off-site Dose Calculation Manual Post Accident Sampling System pico-Curie (1.0E-12 Ci)
Primary Makeup Water parts per billion parts per million Plant Saint Lucie Enclosure
- Pressurized Water Reactor 0A
'- 0uality Assurance RCA
- Radiation Control Area RCS
,- Reactor Coolant System REHP
- Radiological Environmental Monitoring Program RESL
- Radiological and Environmental Sciences Laboratory Rev
- Revision RMS
- Radioactive Material Shipment RPH
- Radiation Protection Han RWT
- Refueling Water Tank SCC
- Stress Corrosion Cracking SFP
- Spent Fuel Pool SG
, - Steam Generator TLD
- Thermoluminescent Dosimetry TS
- Technical Specification Enclosure
0