IR 05000397/1992008
| ML17289A524 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/23/1992 |
| From: | Louis Carson, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17289A523 | List: |
| References | |
| 50-397-92-08, 50-397-92-8, NUDOCS 9205110098 | |
| Download: ML17289A524 (27) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION
REGION V
Repor t No.:
50-397/92-08 License No.:
NPF-21 Licensee:
Washington Public Power Supply System P.O.
Box 968 3000 George Washington May Richland, MA 99352 Facility name:
Washington Nuclear Project No.
2 (WNP-2)
Inspection at:
WNP-2 Site, Benton County, Washington Inspection conducted:
March 23 through 27, 1992 and subsequent telephone discussions on March 31, April 13, and April 15, 1992 Inspected by.
arson
,
eac or a )a
>on pec>a
>s a
e
>gne Approved by:
Q.P.
as,
)e
,
eac or a
>o og>ca Protection Branch
~Summa@:
~AI td:
a e
sgne This routine unannounced inspection covered the licensee's radiation protection activities in preparation for refueling outage seven (R-7).
This inspection included ALARA planning,'occupational exposure control, training health physics (HP) contractors, general employees training (GET), chemical decontamination, radwaste management, and spent fuel pool cooling operations.
Inspection procedures 83722, 83727, 83729, and 86700 were used.
Results:
No violations or deviations were identified.
For the R-7 outage, licensee management contracted a sufficient number of HP technicians to supplement their HP staff, and modified its HP organization to supervise HP contractors and control worker exposures.
However, management was slow to recognize that resources allocated to the HP work control liaison and ALARA planning group to prepare for the R-7 outage were inadequate.
The licensee's residual heat removal (RHR) system to the spent fuel pool (SFP) cooling assist mode test was successful in providing the required flowrate to the SFP.
However, radiological safety concerns were identified regarding the SFP water clarity and being able to position fuel rods while RHR )s operating.
I 9205110098 920423 PDR ADOCK 05000397
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1.
Persons Contacted:
Licensee:
DETAILS
"L. Harrold, Assistant Plant Manager
- A. Hosier, WNP-2 Licensing Manager
- A. Alexander, Health Physics (HP)/Chemistry Support Supervisor
"L. Bradford, HP/ALARA Planning Supervisor
"S. Flood, System Engineer
"S.
Kim, ALARA Engineer
- D. Thompson, Special Projects
- RE Davis, Materials Engineer
- C. Madden, equality Assurance Engineer
- P. MacBeth, Radwaste Supervisor
"L. Morrison, Chemistry Supervisor
"D. Wer1au, HP, Chemistry and General Employee Training Manager
"C. McGilton, Operational Assurance Manager
- J.
Rhoads WNP-2 Operations Event Assessment Manager
"H. Reis, Plant Engineering/Compliance Supervisor (*)
Denotes personnel who were present at the exit meeting held on March 27, 1992.
In addition to those individuals listed above, the inspector met and held discussions with other licensee personnel.
NRC:
M. Royack, Reactor Safety Branch Engineer, Region V
2.
ALARA 8 Occu ational Ex osure Durin Extended Outa es 83729)
Some of the licensee's R-7 outage plans, preparations, and goals were previously discussed in NRC Inspection Report No. 91-45.
The inspector examined the licensee's current state of readiness for the 75 day R-7 outage which starts on April 17, 1992.
a.
Audits 8 A
raisals The inspector examined the following equality Assurance Surveillance Reports (gASR), Technical Assessments (TAs) and Problem Evaluation Reports (PERs) to look for deficiencies that effect the R-7 outage:
gASR 292-012 (DRAFT):
This gASR evaluated critical radiological tasks associated with a reactor water cleanup system septa/filter element replacement.
The inspector noted that this gASR identified deficiencies in the Maintenance Work Request (HWR) to Radiation Work Permit program interface.
This gASR, also, revealed that the licensee's program did not have a
method to clearly delineate when radioactive materials become radwaste.
Therefore, radioactive materials and radwaste stored in the facility since 1985 were not accounted for by the radwaste departmen hl
QASR 292-029 (DRAFT):
This QASR evaluated the licensee's ability to effectively implement the corrective actions listed in PER 292-083.
This PER described concerns that management directed HP technicians to release areas from radiological control contrary to procedures.
On March 26, 1992, QA noted that corrective actions agreed to by the licensee's Management Review Committee were not implemented by HP.
QA issued Quality Finding Report (QFR) 292-029-2 to document HP's failure to implement corrective actions properly.
- PER 292-083:
addressed above PER 292-176:
This PER identified a problem with cigarette butts being found on several occasions inside the radiologically controlled area (RCA) on the Radwaste Building roof.
On March 18, 1992, the plant manager issued a letter to plant staff which restated the licensee's policy on smoking within the RCA, and that disciplinary action will be taken on the perpetrators.
PER 292-081:
This PER investigated a storm drain that
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otentially was a unmonitored radiological pathway.
The icensee'
eff1 uents speci al ist wrote a 1 etter on thi s concer n on March 2, 1992, and was continuing to investigate this issue.
PER 292-192: This PER documented an unnecessary exposure that occurred when a radiographer exposed the wrong film.
A repeated radiograph resulted in an additional 20 millirem (mrem) dose to the radiographer.
The licensee's controls on radiographers were discussed in NRC Inspection Report 50-397/91-31.
- PER 292-207: This PER, written on March 10, 1992, said the licensee's proposed dose reduction fixes for the control rod drive SCRAM discharge volume (CRD SDV) were temporary, untimely, and may be compromising from a regulatory position.
The radiation levels in the CRD SDV area exceeded one rem/hour, which required the area to be posted as a High Radiation Area.
The inspector reviewed several proposals for reducing the CRD SDV radiation fields, and noted that this was part of the licensee's hot spot reduction program.
Recommendations to reduce the CRD SDV dose rates included system redesign, chemical decontamination, and hydrolazing.
Hydrolazlng the CRD SDV will be performed during the R-7 outage.
PER 292-245:
This PER, written on March 18, 1992, investigated and corrected the resident inspector's finding that the CRD SDV high radiation area was posted, but the flashing warning lights were not flashing.
The licensee's corrective actions included the use of redundant flashing lights, modifying the lights with
venting ports, and procuring a different light design by April 17,,
1992.
PER 292-247:
This PER, written on Harch 20, 1992, evaluated a
spill incident that occurred during the routine bucket verification test of the equipment drywell leak detection system.
The licensee will determine if the bucket test is necessary, and how often.
PER 292-254:
This PER, written on March 25, 1991, identified that a 10 milliCurie nickel-63 test source had a 35,000 disintegrations per minute leak.
The HP department confiscated and bagged the source.
The PER noted that this incident will be in their annual report.
TA 92-001:
This was gA's technical assessment (TA) of the MNP-2 chemistry program's analytical capability, and ability to accomplish technical goals for primary and secondary plant systems.
This TA will occur during the R-7 outage in the following areas:
Chemical decontamination activities Liquid radwaste processing and recycling to the condensate storage tank Fuel pool and reactor cavity water control and processing Equipment lay-up processes used during outages The HP/chemistry manager provided the inspector monthly trending data on the number of PERs generated by plant personnel since 1990.
The projected number of PERs for 1992 is 60; 44 PERs were written, so far, in 1992, as compared to 44 in all of 1991.
According to the HP/chemistry manager, personnel were encouraged to write PERs.
The gASRs, PERs, gFRs, and TAs indicated that the gA and HP/chemistry increased its vigilance in overseeing radiological areas, which should continue during the R-7 outage.
b.
Plannin 8 Pre aration for the R-7 Refuelin Outa e
The inspector examined the current status of HP and ALARA planning.
The inspector attended the last scheduled Senior Site ALARA Committee meeting before the R-7 outage.
The inspector discussed R-7 outage activities with outage management, HP/ALARA planning, HP operations, plant chemistry, refueling floor operations, and plant engineering.
ALARA Goals The R-7 outage ALARA exposure and personnel contamination goals are 600 person-rem and 188 skin/clothing contaminations respectively.
The licensee expects to spend 415 person-rem on seven main outage jobs, which are as follows:
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JOB In service inspection (ISI)
Main steam relief valve setpoint test CRD removal/replacement Recirculation pumps discharge valves ALARA Shielding Low power & intermediate range monitors Snubber Optimization PERSON-REM 112 130
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415 These ALARA goals were approved by the Site Sr.
ALARA Committee.
R-7 Outa e Sco e ALARA/Radiation Work Permits The inspector examined the work scope of the R-7 outage to assess its impact on the HP/chemistry department and the ALARA planning roup.
The inspector had discussions with the HP/Chemistry Manager, LARA planning staff, the site work control group and R-7 outage management personnel.
The R-7 outage management group had not finalized the total number of outage activities.
The cut off date for adding work activities to the outage was March 31, 1992.
At the time of this inspection the planned R-7 outage scope was as follows:
56,700 person-hours and 991 Maintenance Work Requests (MWRs)
39,516 person-hrs and
MWRs for drywell activities
9183 person-hrs and 222 MWRs for Reactor Building & Wetwell work 7991 person-hrs and 415 MWRs for Radwaste
& Turbine Building work Out of the 991 MWRs, 727 required that individuals sign on Radiation Work Permits (RWPs) in order to perform outage related task.
Other components of the R-7 outage work scope were preventive maintenance (PM) packages, surveillances (PPMs),
and mundane work task (MWTs),
which accounted for 380 activities requiring individuals to sign on RWPs.
The majority of the MWRs and work task (727 + 380 = 1107) can be performed under pre-existing general RWPs.
However, approximately 200 MWRs require specific RWPs to be written by HP and ALARA planning.
The inspector revealed that as of March 27, 1992, only 5 out of 200 R-7 RWPs were ready.
The inspector was concerned that the number of undeveloped RWPs at this phase of outage preparation had a radiological safety impact in the following areas:
The quality of pre-job ALARA reviews and RWPs to be generated for the outage in a short period.
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The amount of overtime that ALARA planners will work prior to starting the outage on April 17, 1992, and into the outage.
The HP/ALARA planning staff, outage management, and the inspector discussed why WNP-2 was behind in R-7 outage radiological paperwork preparation, and the reasons were as follows:
The R-6 outage was scheduled for 60 days April 1991 to June 1991, but it lasted 90 days longer, unti) October 1991.
WNP-2 had at least six forced mini outages between October 1991 and March 1992; Forty five percent of 130 post-job ALARA reviews and lessons learned analysis from the R-6 outage were not performed yet, and those reviews were needed for generating R-7 RWPs, Plant rocedures manual (PPM) 11.2.1.2, ALARA Program mplementation,"
and PPH 11.2.2.6
"ALARA Outage Planning,"
requires completed post-job ALARA reviews.
However, there is no specified time to complete such task.
HP/ALARA plannirg had six HPTs and five ALARA engineers assigned 'to develop R-7 outage RWPs and pre-job ALARA reviews.
One HP/work control liaison was responsible for identifying all R-7 outage work requiring ALARA Mork Scope Sheets (AWSS), and RWPs.
- The HP/ALARA planning and HP/work control resources were not dedicated to R-7 outage activities; they still had daily routine work responsibilities.
The ALARA planning group was scheduled to be moved three times between the time of this inspection and April 20, 1992.
ALARA planning and HP did not have most of the HWR AWSS packages needed to develop RWPs.
The licensee planned to use
HP technicians to help write RWP instructions for HHR jobs.
The ALARA planning engineers were going to supervise RWP development.
However, ALARA planning wanted completed R-6 outage p~st ALARA job reviews in order to incorporate lessons learned into R-7 outage RWPs.
During the Sr.
ALARA Committee meeting on Narch 25, 1992, R-7 outage management announced that 20 to 30 percent of MMRs were ready to work.
HP and ALARA planning explained that HWR AMSSs were not submitted for writing RWPs in accordance with the PPH 11.2.1.2 and PPM 1. 11. 18, "Radiation Work Permit."
The inspector noted that licensee procedures did not address timely submittals of AWSSs for outage situations.
PPH 1. 11.8, Section 5. 1. 1, provided passive guidance for submitting AMSSs at least two working days before starting work, but the procedure allowed a waiver of the two day guidance to support urgent work.
ALARA planning estimated that
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AMSSs were submitted in accordance with procedure PPH 1.11.8, Section 5. 1. 1, only 20 percent of the time.
Also, fifty percent of the time the AWSSs were submitted within 1 working day, and thirty percent of the AWSSs were submitted within half of a working day.
The inspector evaluated the impact that an urgent HMR had on the ALARA planning program.
The urgent work involved stopping some main steam valve leaks.
On March 25, 1992, licensee management gave the ALARA planning group a one day notice to develop a specific RWP.
The ALARA engineer stopped working on R-7 outage preparation in order to plan for the main steam leak job.
The AMSS arrived at the ALARA planning group on Harch 26, 1992, and the valve leak work was performed on the 27th.
This main steam valve work expended a
cumulative radiation dose of 430 millirem.
On Harch 31, 1992, a telephone conference call was held between licensee plant management,'NRC resident inspectors, and NRC Region V
staff on the state of the licensee's R-7 outage preparedness.
The licensee's plant manager described their increased HWR, ALARA and RWP efforts as follows:
HP and ALARA planning was put on 60 hour6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> work weeks.
Twenty-five contract HPTs were assigned to assist ALARA planning.
The HWRs and RWPs required for the first month of the R-7 outage were identified.
This included working the R-6 outage post job ALARA reviews that were needed in support of the R-7 outage.
A total of 425 HWRs, preventive maintenance
& PPH packages with the AWSSs were sent to ALARA planning, since the on site inspection.
Plant management agreed to contact the NRC Regional Office on April 13, 1992, to give an update of MNP-2's R-7 outage preparation.
The inspectors did not have any further concerns in this matter.
R-7 Outa e
Su lies Facilities
& E ui ment (83727
& 83729 The inspector reviewed the licensee s program for assuring supplies and equipment were available to support the R-7 outage.
The licensee assigned a Sr.
HPT the task, and he explained that most HP supplies, equipment, and services were controlled through a site warehouse minimum/maximum inventory program.
However, some special orders of equipment were made which included the following:
Radiation detectors:
30 ionization chambers, R02s
& R02As;
Teletectors; 20 pancake probes; 5 Geiger counters; 1 R07 underwater radiation detector; and 2 frisking booths, (IPH-8s)
Air Sampling Equipment: six mobile cart air samplers, filter heads, and cartridges
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100 multi-dosimeter badge suits Other miscellaneous equipment 8 supplies included:
50 lead blankets; plastic boots for CRD work; HEPA filters The Sr.
HPT was overseeing an R-7 pre-outage project to convert the bathroom on the Refueling Floor to a frisking booth room equipped with an IPM-8.
Also, an IPM-8 frisking booth will be installed on the Turbine Generator Floor.
The inspector had no concerns with this aspect of the licensee's R-7 outage planning.
Trainin
& ualifications NRC Inspection Report 50-397/91-45 addressed enhancements to the licensee's General Employees Training (GET) and HP Training Program.
Contract HPT training 'was in progress at the time of this inspection.
The inspector discussed training plans and schedules with the HP/GET training manager.
Some of the required special training included:
Decontamination training Lessons learned training from the April 1991 resin sp'ill Hot particle control mock-up training The inspector discussed details of the hot particle training with the instructor and a
HP craft supervisor.
The licensee s training schedule and plans covered several groups of contractors, including HP contractors.
The inspector had no concerns in this area.
Control of Radioactive Materials Radiation Fields and on amlna
>on The inspector examined plans for controllinq radiation fields, contamination, and radioactive material dunnq the R-7 outage.
The licensee was preparing to chemically decontam)nate the reactor recirculation (RRC) system discharge piping, reactor pressure vessel (RPV) nozzle flushes, and hot spot program system flushes.
The licensee appointed an R-7 outage contamination control supervisor to monitor and correct adverse work situations.
The HP/chemistry department wrote letters to plant departments and procedures for reinforcing material controls during the R-7 outage.
Material Controls The inspector reviewed the following licensee procedures and letters on controlling radioactive materials and non-radioactive materials:
PPM 1. 12.7,
"Extended Term Radioactive Material Storage,"
established requirements for extended term storage of radioactive materials in WNP-2 facilities, including the Warehouse and the spent fuel pool.
PPM 1. 12.7, also, requires that Maintenance Work Request instructions include material disposal.
The HP/chemistry manager wrote a letter to plant personnel dated March 25, 1992, re-emphasizing PPM 1. 12.7, and
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statinq that any orphan material not in compliance with the PPH by Apr>l 17, 1992, will be disposed of by the radwaste department.
PPH 1.3.9, "Plant Material Condition Inspection Program,"
was revised on March 2, 1992, to establish guidelines and goals for contamination control, radioactive and hazardous waste reduction, equipment design and material control.
The revisions to PPM 1.3.9 assigned more responsibility to individuals, and area coordlnators for implementing the program.
PPH 6. 1.1, "Spent Fuel and Equipment Inventory," was revised on March 4, 1992, and it required inventory list for items in the spent fuel pool and other equipment pools.
This procedure is implemented when the reactor pressure vessel head is removed and reinstalled.
The primary change in this PPM was to establish more stringent radiological controls to minimize potential unplanned exposures from disintegrating materials stored in the spent fuel pool.
The WNP-2 Dangerous Waste and Waste Minimization Reports were sent to the State of Washington on February 28, 1992.
The licensee reported they were evaluating the use of on non-regulated hazardous materials.
The inspector noted changes in the licensee's hazardous mixed waste program.
The licensee reported a reduction in the use of petroleum based solvents such as methylene chloride.
However, the inspector detected the strong odor of methylene chloride being used by workers in the Reactor Building, during a tour on March 24, 1992.
The inspector reported this observation to the HP craft supervisor who followed up on the report.
The licensee's programs in the area of material controls were adequate.
The inspector had no concerns in this area.
Chemical Decontamination The licensee projected saving 330 person-rem by performing the chemical decontamination on the RRC system discharge piping during the R-7 outage.
Some of the licensee's chemical decontamination plans were addressed in NRC Inspection report 50-397/91-45.
The licensee decided to only decontaminate the discharge piping of the RRC system, because decontaminating the RRC system suction piping would only save an additional 36 person-rem.
Also, the licensee's engineers determined that the decontamination process could adversely affect the Inconel 182 weld material on the RRC suction piping by accelerating cracking.
The weld crackinq concerns were addressed
)n the licensee's
CFR 50.59 Safety Evaluat)on.
The inspector reviewed the
CFR 50.59 Safety Evaluation on the chemical decontamination.
The evaluation addressed reactor material, chemical, radwaste and radiological safety.
The inspector reviewed procedure PPH 10.2.89, Reactor "Recirculation System Chemical Decontamination,"
schedules, drawings, and plans.
The
inspector and the system engineer discussed chemical decontamination lans and toured facilities where decontamination equipment will be ocated.
The ALARA planning department appointed the Sr.
ALARA engineer as the chemical decontamination coordinator responsible for all HP concerns.
The chemical decontamination coordinator and the system engineer visited a facility in Region III to observe a chemical decontamination in order to incorporate that experience into WNP-2s efforts.
The inspector, also, had discussions with the materials engineer, radwaste supervisor and the plant chemistry process engineer.
The licensee sent a
RRC system test flange to a contractor, and had the chemical decontamination process applied to it and analyzed.
The test results and the data indicated that the chemical decontamination process should succeed.
The inspector asked the licensee about some of the special radwaste considerations involved with the chemical decontamination'.
The radwaste supervisor explained that 60-100 Curies of solid radwaste would be disposed of.
The radwaste supervisor, also, explained that the State of Washington approved WNP-2's request to dispose of chelate chemical decontamination waste with a six to seven percent by weight content.
The inspector reviewed a letter from the State stipulating that chelated wastes must be disposed of in a polyethylene container, and placed in an approved concrete engineered barrier.
The inspector, also, noted that a
chemical process eng>neer performed an hazardous waste evaluation on the chemicals being used in the decontamination.
Other special considerations will be to control the chemical decontamination residue going to the liquid radwaste system.
The inspector's questioned if the chemical decontamination procedure allowed sufficient water velocity to thoroughly remove residual particulate CRUD that might be left inside the RRC pipe walls.
The system engineer, materials engineer, ALARA engineering, radwaste supervisor, and plant chemistry decided to perform a visual inspection and take a sample of the RRC pipe wall after the chemical decontamination.
The system engineer wrote a letter dated March 27, 1992, recommending post decontamination CRUD removal corrective actions and precautions.
The inspector asked if HP and chemistry personnel were going to correlate RRC system piping dose rate reductions with the total radioactivity removed during the chemical decontamination.
HP and chemistry said the chemical decontamination contractor will collect, trend, and analyze the process.
HP will record radiation survey data from various locations on the RRC piping at different times during the process.
Chemistry personnel will collect and analyze RRC system samples in accordance with the chemical decontamination procedure.
The radwaste department will quantify the radioactivity collected and processed in accordance with the solid radwaste program.
However, the licensee has no intentions of collectively analyzing the data points, because the overall objective of this chemical decontamination effort is to lower radiation levels for the R-7 outage.
Long term benefits are only expected as a function of the RRC system recontamination rate, and the licensee's water chemistry controls progra ~
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Reactor Pressure Vessel S ra Nozzle Flushes The licensee contracted services to high volume, low pressure flush eighteen reactor pressure vessel (RPV) spray nozzles in order to save
person-rem during the outage.
This method of decontamination is suitable for reducing radiation levels and CRUD traps.
The only concern the inspector raised regarded the direction of the flush water flow 'and CRUD material.
The licensee engineers said that most of the CRUD will be flushed into the RPV, and any residual CRUD transferred to the RRC system should be carried out with the chemical decontamination process.
The inspector noted that the volume of radioactive material put into the RPV annulus by this process is insignificant in comparison to what is already there.
The licensee was planning to clean out the RPV annulus during the R-9 outage.
S stem Flushes System flushes will be performed on the residual heat removal system, the control rod drive scram discharge volume piping, and the liquid radwaste drain system.
-These systems were identified in the Hot Spot Dose Reduction Program at WNP-2.
During the Sr.
ALARA Committee meeting members stressed 'that plant engineers, operations, and HP/ALARA planning had to coordinate flushes 4uring the R-7 outage.
Committee members agreed that the spent fuel pool cooling system operations and local leak rate testing on valves presented opportunities to flush systems and lower radiation fields in the Reactor Building.
Material 8 Contamination Control Durin Valve Work The inspector examined other efforts towards lowering'radiation fields and controlling the spread of contamination.
Particularly, the inspector was concerned about efforts to reduce the spread of cobalt/stellite materials, during the R-7 outage.
The inspector found that the ALARA/HP group identified most of the plant's stellite bearing valves, and asked management to incorporate a stellite valve list into a computer base that would bring special attention to those valves.
The ALARA/HP group, also, wrote instructions using the latest industry guidance on cobalt reduction during valve maintenance work.
The inspector revealed that none of the ALARA/HP group's efforts were incorporated into plant procedures PPM 1.3.19, "Plant Material Condition Inspection," or PPM 10.1.13,
"Systems Cleanliness Control."
ALARA or HP did not take the initiative to incorporate their efforts into Radiation Work Permit instructions.
However, after the inspector bought this observation to the attention of licensee plant staff, the maintenance department volunteered to incorporate the cobalt reduction instructions into the applicable Maintenance Work Request.
The licensee's overall planning and preparation efforts for the R-7 outage were adequate.
However, management was slow to recognize that the limited resources allocated to the ALARA planning group, and HP work control impacted their ability to develop timely, detailed ALARA Work Scope Sheets for effective Radiation Work Permits.
No violations or deviations were identifie '
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HP Chemistr Radwaste R-7 Outa e Staffin
& Or anization Chan es
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ana b.
emen on ro s Staff Chan es The inspector noted the following changes to the licensee's HP/Chemistry staff, since the last inspection.
The assistant HP/chemistry manager was promoted to the HP/chemistry manager position in January 1992.
The HP/chemistry manager's qualifications were previously discussed in NRC Inspection Report 50-397/91-10 when he was promoted from the plant HP supervisor position.
The plant HP supervisor position was still vacant, however, the licensee was still searching for a candidate.
The HP/chemistry manager was considering naming an acting plant HP supervisor in order to enhance his staff during the R-7 outage.
The assistant HP/chemistry manager position has been eliminated.
A corporate senior health physicist resigned.
This individual, also assisted the gA operations event analysis (OEA) group in reso)ving radiological PERs requiring root cause analysis.
The OEA manager reassigned al.l radiological PERs to a staff member.
The Corporate Radiological Health Officer will be assigned as a Drywell coordinator during the R-7 outage.
The corporate radiological health group supporting plant HP and OEA was mentioned in NRC Inspection Reports 50-397/91-45, 50-397/91-34, 50-397/91-10.
R-7 Staffin
& Or anization The plant HP technician (HPT) staff, according to the HP/chemistry manager consisted of 24 Sr.
HPTs and 14 Jr.
HPTs.
There are currently four openings for Sr.
HPTs.
The plant HPT staff will be supplemented with 125 contractor HPTs (44 Jr.
HPTs
& 81 Sr.
HPTs)
for the R-7 outage.
This was an increase of 24 contract HPTs, because the plant manager approved a request from the HP manager to hire additional help.
The five ALARA engineers will be assigned areas of radiological responsibility for supporting the outage.
For example the Sr.
ALARA engineer was assigned to support all radiological functions relating to the chemical decontamination.
Plant HP supervision for the outage will have two persons on day and night shift.
All HP personnel will be scheduled to work sixty hours per week in support the R-7 outage.
The operations HP supervisor, HP craft supervisor, and the ALARA supervisor discussed the HP organizational structure that will be deployed to ensure supervisory control over contract HPs and radiological work.
The licensee provided the inspector a copy of the R-7 HP assignment list.
Groups of HPTs will be assigned to work in following areas:
Reactor Building Refueling Floor Turbine Building/ Balance of Plant The turbine generator/rotor work Drywell including the chemical decontamination
Oosimetry Laundry Each of the above areas will be assigned a day and night shift lead HPT with a complement of Sr.
and Jr.
HPTs.
There will be eight chemist assigned as HPTs in support of the chemical decontamination and refueling floor activities.
Two HPTs will specifically support the testing of motor operated valves and local leak rate testing.
There will be four HP supervisors, three HP technical support staff, and a
HP contractor coordinator for providing supervisory control of HP personnel and workers in the RCA.
The licensee's HP staff augmentation for supporting the R-7. outage was adequate.
The number of radiological professionals in plant HP staff, OEA, gA, and corporate radiological health warrant some licensee management attention.
-No violations or deviations were identified.
S ent Fuel Pool Activities 86700 The licensee was planning to off load all the fuel in the reactor core, and store it.the spent fuel pool (SFP), during the R-7 outage.
The inspector observed some of the SFP operations and reviewed SFP procedures being implemented in preparation of the R-7 outage.
Licensee requ>rements for the SFP cooling system were in Technical Specification
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TS) 3.9.8, TS 3.9.9, and the Updated Final Safety Analysis Report UFSAR) Chapter 9.1.3.
Safe Water Shield Above Active Fuel TS 3.9.8 and TS 3.9.9 requires the water levels above the top flange of the reactor vessel, and the top of SFP racks be at least 22 feet, during refueling operations.
According to a licensee letter dated March 24, 1992 several procedures must be revised by April 17, 1992 (R-7 Start date) in order to assure consistency with the TS requirement.
A iiua1ity Finding Report ((FR) 2-91-022 addressed the lack of procedures for situations that allowed bypassing the refueling bridge up limit stops.
Refueling bridge operational limits are found in TS 3.9.6.
The licensee was aware of the potential radiological safety consequences of this unacceptable condition, and they were resolving it.
The inspector had no further concerns in this a matter.
SFP and E ui ment Inventor Ouring a previous inspection, the inspector asked the HP/chemistry manager how they accounted for storing materials in the SFP; especially materials in the SFP not by design.
The inspector referenced the concerns identified in NRC Information Notice (IN) 90-47
"Unplanned Radiation Exposures to Personnel Extremities Oue to Improper Handling of Potentially Highly Radioactive Source."
Ouring this inspection, the licensee's refueling floor supervisor performed a test on two types of rope materials (polypropylene 8 polypropylene with a carbon polyurethene coating).
These types of rope materials are used for tying off materials oyer the side of the SFP.
The test was being conducted under the licensee's mundane task program under the technical assistance of the
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HP/chemistry department.
The inspector observed portions of the test which required that two metal cylinders, with the rope material inside to be lowered into the SFP for irradiation by spent fuel.
The results of the test were not final at the time of this inspection, however, preliminary results suggested that the polyurethene coated rope was radiologically more su>table for use.
The inspector observed good HP practices and support during the thst.
The inspector reviewed licensee rocedure PPM 6.1. 1, "Spent Fuel and Equipment Pool Inventory "
and icensee practices were consistent with this PPM.
The inspector did not have any further concerns in this matter.
RHR S stem to SFP Coolin S stem Assist Mode Test In February 1992 the licensee concluded that the spent fuel pool cooling (SFPC)
system could not adequately cool a full reactor core off load during the R-7 outage without the assistance of the residual heat removal (RHR) system.
The licensee wrote test procedure PPM 8.3.227TP,
"RHR Assist Mode of Fuel Pool Cooling Test,"
and implemented the test on March 24, 1992.
The inspector observed the performance of the RHR to SFPC test.
The inspector observed HP support during the initial test setup and implementation.
Some of the valves operated for the test were located in high radiation areas.
,IP provided coverage at the SFP during the removal of the refueling platform equipment.
HP monitored radiation levels and airborne radioactivity during the RHR pump discharge test phase, and monitored the SFP for floating debris resulting from the test.
The test demonstrated that RHR and SFPC had a combined flowrate of 2,750 gallons/minute (gpm);
RHR 2200 gpm and SFPC 550 gpm.
There was no detectable increase in radiation levels or airborne activity on the refueling floor or the SFPC pump room.
The inspector took independent radiation measurements with NRC ionization chamber 9154 (calibration due date on April 7, 1992) during the SFPC operations.
The noise level at the flow restricting orifices, in the hall outside the SFPC pump was
decibels.
However, the SFP water quality diminished.
The visual opacity of the SFP water increased from 0.09 net turbidity units (NTUs) to 0.32 NTUs; the cobalt-60 concentration increased from 2.77E-4 microCuries/cubic centimeters (uCi/cc) to 5. 12E-4 uCi/cc.
The visual clarity of the SFP water with RHR system running would not allow safe movement of reactor fuel or spent fuel.
Licensee engineers were addressing this safety concern.
The engineers were considering vacuuming the bottom of the SFP, and taking advantage of the SFP turbulence with RHR running in order to better cleanup the SFP.
The inspector did observe that the SFP clarity was restored the next day.
The system engineers placed plastic and tape over the SFP vents in order to prevent water from getting inside.
The vent covering did not completely work.
The system engineer and the inspector observed water leaking from the SFP ventilation drains.
The HP/chemistry department wrote Performance Evaluation Report (PER 292-253)
on the leak because HP was re-assured during Plant Review Committee meeting that plastic covers would be adequate.
The system engineers will have to reset the SFP skimmer surge tank weirs to prevent the surge from the RHR system flow from filling the SFP level up to high.
The inspector had no further concerns in this are,
The licensee's engineers and HP technicians worked well together during these SFPC operations.
The test was successful, however, SFP clarity is required for safe fuel movement during the R-7 outage.
No violations or deviations were identified.
The licensee's programs were capable of accomplishing its safety objectives in support of the R-7 refueling outage.
however radiological professional staffing constraints has affected the licensee's ability to efficiently resolve radiological issues.
This observation was most evident considering the impact that constrained resources has had on ALARA planning since ending the R-6 outage.
No violations or deviations were identified.
5.
Exit The inspector met with the licensee representatives identified in Section 1 of this report on March 27, 1992.
The scope and findings of the inspection were discussed.
No violations or deviations were identifie l i
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