ML20203A014
| ML20203A014 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/14/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20202J818 | List: |
| References | |
| 50-155-97-15, NUDOCS 9802230240 | |
| Download: ML20203A014 (24) | |
See also: IR 05000155/1997015
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U.S. NUCLEAR REGULATORY COMMIS810N
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REGION lil
! Docket No: 50 155
License No: DPR 06
Report No: 50155/97015(DNMS)
Licensee: Consumers Energy Company
Facility: Big Rock Point Nuclear Power Plant
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Location: 10269 U.S. 31 North
t Charlevoix, MI 49720
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Dates: December 5,1997 February 2,1998
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Inspectors: R. J. Leomon, Senior Resident inspector
C. E. Brown, Resident inspector -
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R. B. Landsman, Project Engineer
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J. E. House, Senior Radiation Specialist
P. W. Harris, License Project Manager
Approved by: Bruce L Jorgensen, Chief
Decommissioning Branch
Division of Nuclear Materials Safety
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9802230240 980214
PDR ADOCK 05000155
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EXECUTIVE SUMMARY
Big Rock Nuclear Power Plant
NRC Inspection Report 50155/97015
This routine decommissioning inspection covered aspects of facility management and control,
decommissioning support activities, spent fuel safety, and radiological safety.
-e Overall, the licensee's performance during the chemical decontamination process was
good. However, there were two instances of poor performance in the health physics
area.
Facility Manno 9 ment and Control
e The inspector concluded that changing the Site Emergency Plan to reduce the minimum
shift crew composition by one control room operator and one auxiliary operator was within
the current TS requirement for a plant in cold shutdown.
- Good cooperation and teamwork between the contractor and plant staN resulted in a high
quality, workable procedure to be used in the chemical decontamination process,
o Overell, the chemical decontamination of the primary system was a success. Reduced
dose rates should enable workers to dismantle the plan' faster witn lower accumulated
radiation doses, and with less risk of altbome contamination when major
piping / equipment systems are breached.
- The plant safety review committee evaluated twenty plant systems classified as non-Q -
and determined they are not needed for a pemianently defueled plant. These systems
were designated as available for decommission!ng,
o Administrative Procedure D1.11 " Safety Evaluation," was revised such that the
requirements of 10 CFR 50.82 were incorporated into the procedure. The licensee's use
of Procedure D1.11 should help ensure that the decommissioning fund will always
contain enough money to go to SAFESTOR or to complete the total decommissioning
job.
Qt&gmmiulonino Support ActlMel
e Installation of a HEPA filter in the reactor building ventilation was completed in December
1997, completing a PSDAR commitment and bringing the ventilation system in
compliance with the FGEIS. The ventilation system now has filtration for containment
exhaust air.
e The time for the SFP to heat-up from 80' F to 150' F was verified by actual testing to be
greater than the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> assumed in the licensee's submittal to the NRC.
- The materialintegrity of structures, systems, and components necessary for safe
storage of spent fuel and conduct of safe decommissioning was being maintained, and
plant housekeeping and fire protection wem generally good.
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Spent Fuel Safety
e The inspector concluded that the licensee can readily accommodate various conditions
that may challenge integrity of the fuel.
- A concern was identified relating to the potential for an individual inadvertently pulling up
a high dose item suspended on a cable in the SFP high exposure. Plant management
had the SFP covers put on the pool as a temporary solution.
RadiologiseL6.aMy
- The inspectors identified one violation of Technical Specifications conceming a
decontamination worker in a HRA not monitoring his dosimetry as required and not noting
that his ELD was on pausr. This inciden' indicated that worker attention to radiation
protection practices (especially in the area of monitoring dosimetry), remained a problem
for the station personnel.
- The chem decor project included appropriate dose minimization planning, and effectively
incorporated ALARA principles.
- The licensee's radiation prctection program for the chem decon prncess was also well
executed. Communication between licensee personnel and the vendor performing the
chem decon were very good.
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Egoori Detalls
EVinalgrYJofEant Activitin
The reactor coolant, shutdown cooling, and cleanup systems were chemically decontaminated.
Licensed reactor and senior reactor operators completed requalification training to keep their
NRC licenses active. Tests on spent fuel pool (SFP) heatup rates were conducted, twenty-plant
systems were classified as available for decommissioning, and the emergency plan exercise was
deferred until March 1998.
1. Facility Management and Control
1.1 Genersj
The in,spectors conducted frequent reviews of ongoing plant activities and attended
licenten meetings and reviews addressing these activities, in order to assess overall
facility management and controls. Specific events and findings are detailed in the
sections below,
l.2 QID#alleligBlaQRS9 ment. aniCMLQontrols at Permanently Shyt Down RfAcigIn
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l.2.1 QantIgj
The inspectors reviewed the licensee's systems for overall management and control of
the decommissioning process. Specific processes which were selectively examined
included the programs for identification and resolution of safety concems and the
commitment tracking programs and procedures. The inspectors also reviewed and
evaluated the licensee's organization, staffing, qualifications and training, including those
for the contracted workforce, to verify that licensing commitments were being met.
Reference was made to the requirements detailed in the technical specifications (TS),
and the post shutdown decommisaloning activities report (PSDAR). The inspectors also
selectively examined and evaluated the licensee's planning and scheduling to determine
their effectiveness. The effectiveness of the licensee's review of regulatory information.
Including information notices applicable to the facility, was selectively examined. Specific
events and findings are detailed in the section below.
l.2.2 Sigtfina and_ Qu_a_liftcations
a. S_qpng
The inspecter reviewed the process for licensed reactor and senior reactor operators
maintaining license qualifications, and for qualifying operators as certified fuel handlers,
b. Observations and Findinas
On December 16,1997, licensed reactor and senior reactor operators completed
requalification training and passed required testing to meet NRC license requirements;
thus, maintaining their NRC licenses. Ir. January 1998, the licensee started certified fuel
handling training, which takes about seven weeks to complete.
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c. Qpnclusion
The licensee was effectively completing operstor license qualificatione and certified fuel
handlers' training.
1.2.3 EnitIgencyflan_SteILend_Schedvit
a. Ssopp
The inspector followed developments relating to emergency plan exercise rescheduling,
and attended a Plant Review Committee (PRC) meeting where changes to the Site
Emergency Plan, Volume 9, Chapter 5 Section 1.1.1, * Normal Organization" were
discussed. The inspector observed discussion of the changes and the safety evaluation,
b. Observations and Findinas
After discussions with the NRC, the State of Michigan, and FEMA, on December 30,
1997, the licensee was granted a scheduler exemption to defer the 1997 emergency
exercise until March 1998. The new emergency preparedness scenario will represent the
permanently defueled plant.
On January 27,1998, the PRC discussed changes to the Site Emergency Plan that would
change the normal organization (shift complement) from one shift supervisor (SS), two
control room operators, and two auxiliary operators, down to a shift complement of one
SS, one control room operator, and one auxillary operator.
The inspector verified TS 6.2.2, Table 6.21, ' Minimum Shift Crew Composition for a
Shutdown or Cold Shutdown Plant," which requires one SS, one control room operator,
and one auxiliary operator.
The PRC concluded that this change in staffing level would not result in an unreviewed
safety question (USQ), and the change is planned for March 1998,
c. CgaplV1100
Changes to the Site Emergency Plan to reduce the minimum shift crew composition by
one control room operator and one auxillary operator were within the current TS
requirements for a plant in cold shutdown.
l.3 ELalcly31ylewAJfal0D_ChBBatkapdfdOAlfications at Permamently Shut Down ReactoIs
(37801)
1.3.1 Chemical Decontamination Procedure
a. Scope
in December 1997, during PRC meetings, the inspector observed the safety review
process applied in discussions and interaction between licensee and contractor personnel
when Procedure O PCS 01," Primary Coolant System Operations During Chemical
Decontamination," was developed and approved.
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b. Observations and Findinns
The inspector verified that a quorum was present and evaluated ongoing discussion for
the determination that the proposed procedure or changes to the procedure were not
USQs The inspector reviewed the completed procedure prior to the start of the chemical
decor.tamination process.
Items that could possibly be a safety concem were well discussed and resolved by
making procedure changes. The changed procedure' , as retumed to another PRC for
further discussion before determining that a USQ did not exist.
c. Conc!unlen
Good cooperation and teamwork between licensee and contractor personnel resulted in a
high quality, workable procedure to be used in the chemical decontamination process.
l.3.2 Chemical Decontaminatl2D
a. Scope
The inspector examined an extensive chemical decontamination (chem decon) which the
licensee performed on the reactor vessel, steam drum, coolant pumps, primary piping,
and portions of the reactor water cleanup system heat exchangers. The purpose of the
chem decon was to remove contamination from the inside surfaces of reactor system
components in order to accomplish a major reduction in dose rates in the vicinity of those
components through which reactor coolant flowed. The process was designed to provide
a cost effective way to reduce radiation exposure and to concentrate radweste for
disposal, thus reducing radwaste disposal costs,
b. Qb.tifythons and Findinas
The overall process proceeded smoothly with few problems. Over 400 curies of
radioactive material were removed from system components and dose rates were
reduced by a factor or 10 or greater. Isotopic activity and curie removal were estimated
by gamma spectroscopy.
The licensee's quality assurance program for the gamma spectroscopy instrumentation
was adequate. A review of instrument control charts indicated that the detector systems
were operating under statistical control.
As this was the first time that this process had been used on an intact plant, there was
minimal industry experience available to assist the licensee in job planning and execution.
From reviewing the work schedule and the results of the overall chem decon process, it
was evident that licensee and vendor planning were thorough and that licensee vendor
communications were very effective. There was minimal schedule slippage, and the
transition from system heat up to chemical injection appeared to have been relatively
trouble free. During the process the licensee identified two system leaks which were
isolated and contained.
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c. Conclusion
Reduced dose rates should enable workers to dismantle the plant faster with lower
accumulated radiation doses. There should also be fewer opportunities for airbome
contamination when major pipinglequipment systems are breached since most
contamination has been removed from interior surfaces of components,
l.4 D9_ggmmill10DlDafttformance_and Stalys Review at Permanentiv_ Shut Down ReggigIs
(71801)
1.4.1 Challenoes burino the Chemical Decontamination Process
a. Scone
The inspector observed activities related to the chemical decontamination process and
noted what challenges arose and how they were addressed.
b. Observations _end Findinas
The challenges and actions to successfully complete the chemical decontamination
process were as listed below:
Items during system setup and startup:
- An air valve, connected to the reactor head during the system setup, leaked
through an open ended pipe. A small quantity of radioactive water spilled into the
containment when the reactor water level was raised. A blank flange was
installed on the pipe and the water level was lowered to stop the leak. - The leaky
valve was then replaced.
e The licensee moved up the scheduled flow test of main flow pump P 12 and
performed the test without lead shielding blankets around the filter housing.
During the flow test, a crud burst occurred which the licensee did not anticipate.
The crud apparently collected in the filter while the system was idle for several
months. The local area dose rate increased from that of a radiation area to a rate
meeting the requirements for a high radiation area. The containment was
evacuated by the health physics (HP) technicians until the dose in the area was
understood,
e The crud burst during the pump startup put crud into piping dead legs, causing
hot spots.
- A contractor was using duct tape to label the valves on the control panel for the
air operated valves. Some of the tape was coming off of the control panel. The
inspector considered using duct tape for labeling a poor practice After the
inspector raised this concem, the duct tape was replaced with permanent marker
labels.
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Items during operation of the system:
e Operations did not identify that a bypass line in the non regenerative heat
exchanger that contained a sample valve with a leaky, carbon steel threaded
reducer. Carbon steel was strongly attacked by the hot flush chemical which
flowed through the threaded leak path and ate away the reducer threads. This
caused a breach in the piping and a 7 spm leak of reactor coolant. The chemical
decontamination process was shut down for several hours, until the leaking valve
was found and closed.
- Main flow pump P 12 sealleakage increased and the pump had to be replaced
with the spore pump. This was a planned contingency.
- More resins were used than expected and more had to be ordered. More room
had to be made in the waste disposal system for the storage of the extra used
resin.
e The emergency diesel generator (EDG), which powered the P 12 pump, ran
continuously for several days. The EDG developed both a lubricating oilleak and
a fuel oil leak, and the turbo charger was alsn using oil. After seven days of
continuous running, the EDG was shut down and the attemate EDG was wired in
and run. The oilleaks were repaired and the turbo-charger was replaced. The
EDG was then retumed to powering the P 12 pump. Use of the attemate EDG
was a planned contingency.
The inspector observed the following strengths during the chemical decontamination
process:
- Contract management worked well with licensee management and a Good flow of
communication was evident. The shift supervisor and control room were also
kept informed of all activities that contractors were performing.
- The original estimate was that the chemical decontamination process would
remove 300 curies of activity. The chemical decontamination process removed
400 curies during the het flush and about 8 curies during the cold flush. The
filters also removed 12 curies of activity.
- Overall, the decontamination factor was at least 10 inside the piping, which was
the objective of the decommissioning process,
c. Conclusion
This was the first time that this chemical decontamination process was used in the field at
a nuclear power plant. The number of curies of activity removed from the piping was
about 100 curies more than expected. However, this extra activity used more resins than
expected and the radwaste problem associated with the extra resin had to be dealt with.
The inspector identified several challenges and some strengths during this chemical
decontamination process. Overall, the chemical decontamination was a success.
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1.5 Plant Procedures M2700)
1.5.1 Systems Available for becommissionina
a. Scope
The inspector reviewed the licensee's process for safely determining that a plant system
was no longer needed to support the safe operation of the spent fuel pool or the spent
fuel pool cooling system. The inspector also attended a Plant Safety Review Committee
(PRC) meeting where changes to the Quality List, Chapter 6, " Classification of Systems'
were reviewed. This screening process reclassified systems as Non-Q for a permanently
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defueled plant and therefore available for decommissioning.
b. Observations and Findinat
on January 28,1998, the inspector observed members of the PRC meeting concoming
the classifications of systems and the determination that systems were no longer needed
for a permanently defueled plant. The list of syt ns determined to be available for
decommissioning was as follows:
. Air Ejectors and Mechanical Vacuum Pump
.- Altemate Shutdown System
. Containment Gamma Mcnitoring System
. Control Rod Drive System
. Emergency Condenser System
. Feedwater System
. Extraction Water Drains
. Liquid Poison System
. Main Steam System
. Neutron Monitoring System
. Primary Coolant System
. Post incident System
. Reactor Cleanup System
. Reactor Depressurizing System
. Reactor Steam Drum
. Reactor Vessel, General
. Reactor VesselIntemals
. Turbine Generator System
. Off Gas System
c. Conclusion
The PRC determined that twenty plant systems are classified as non-Q and are not
needed for a permanently defueled plant. They also determined that these systems are
available for decommissioning. The inspector did not have any concerns with the PRC's
determination that these systems are not needed for the safe operation of the spent fuel
pool or spent fuel pool cooling system.
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1.5.2 Screening Requirements of 10 CFR 50.82 Added to Safety Evaluation Pmeedure
a. Scans .
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The inspot x reviewed 10 CFR 60.82 ' Termination of Lloonse,' and Administrative l
Procedure D1.11, ' Safety Evaluations,' to determine that Procedure D1.11 contained i
applicable requirements of 10 CFR 50.82. !
b. . Observations and Findings ;
The plant safety evaluation Procedure D1.1i, Revision 0, was revised to incorporate 10 i
CFR 60.82 screening requirements. Future plant conf 6guration changes will require - l
reviews against the artierta of 10 CFR 60.69.6c and 59.82. The procedure revision
incorporates the three fundamental screening requirements of 60.82 as follows: licensees
- shall not perform any decommissioning activities that 1) foreclose release of the site for
possible unrestricted use; 2) resuN in significant environmental impacts not previously
reviewed; or 3) resuN in there no longer being reasonable assurance that adequate funds
will be available for decommissioning.
c. Conclusion
The applicable requirements of 10 CFR 60.82 were incorporated into Procedure D1.11 I
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- Safety Evaluation". The licensee's use of Procedure D1.11 should help ensure that '
enough money to go to SAFESTOR or to complete the total decommissioning job will
always be available, l
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1.6 Onsite Follow-uo. Written Reoorts of Non-routine Events at Power Reactor Facilities
(92700)
1.6.1 (Open) VIO 50165/96002-07: Failum to ensure combustible resterials are controlled _
according to T8. The inspector reviewed the proposed corrective actions for this - l
l violation, which relate mainly to outage work. There was no refueling outage in 1997. ;
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. The licensee will be performing similar dismantlement activities within the next few
[ months. The inspector will then determine if the corrocove actions were effective. This ,
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o ~ 1.6.2 _ (Closed) IFl 50155/96006-03: - Sufficient core spray flow. The post incident system relief )
- valves were set at 155 psig and would lift each time the diesel fire pump ran An ,
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engineering evaluation was performed, and the relief valve portion of the system was
L upgraded from 150 psig to 200 psig and the relief valves were replaced with 200 psig '
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valves. The licensee performed a safety evaluation which determined that the system is
!.- not required for plant safety in a permanently defueled plant. This inspector follow-up
i- _ ltem is closed.
l.6.3 (Closed) IFl 50155/96007-02: Delayed dispatch of operational suppori center (OSC)
emergency response team during the 1996 emergency exercise. The OSC response
- . team will no longer be exercised due to the plant being permanently defueled. - Therefore, ,
'" this item is no longer relevant and is closed.
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II. Decommissioning Support Activities
11. 1 M8101tampe and Sp.tyMlance_ALPJtmlanently Ghutdown. Reactors (02tu)1)
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II.1.1 Qgnetsj
The inspection evaluated maintenance and surveillance for structures, systems, and
components (SSCs) potentially affecting the safe storage of spent fuel and reliable
operation of radiation monitoring and effluent control equipment. Direct observations,
reviews, and interviews of licensee personnel were conducted to assess whether
maintenance and surveillance were performed in accordance with regulatory
requirements and resulted in the safe storage of spent fuel and reliable operation of
radiation monitoring and efnuent control equipment. This included the proper
implementation of TS and 10 CFR 50, Appendix B, requirements. An examination of
planned or completed maintenance and surveillance activities was conducted to assess
the maintenance and surveillance process from its inception to its completion.
11. 1 . 2 Installrdon and Testina of Hiah Efficiency Particulate Air (HEPA) Filters
a. tigg.g
The inspector reviewed the PSDAR, work packages, and safety evaluation; held
discussions with the project engineer; and walked down the filter system related to the
installation and testing of the HEPA filters,
b. Observations and Findinas
The Big Rock Point Plant's original design lacked HEPA filters in the reactor building
exhaust. In December 1997, the plant installed HEPA filters for use in the unlikely event
of potential accidents. This installation of HEPA filters brings the plant fully within the
bounds of the requirements of the Final Generic Environment impact Statement (FGEIS).
The HEPA filters were installed prior to the performance of the primary system chemical
decontamination. When the efficiency of the filters was tested, the results were not as
designed. The design efficiency was 99.95 percent efficient and the final test efficiency
was 99.70 percent efficient. The licensee performed an engineering analysis that
accepted the test results as adequate for the performance of the chemical
decontamination. The licensee plans to repair the bypass flow in the filter housing to
bring the filter efficiency up to the design value,
During startup testing of the dampers and the filter flow path, on two occasions the
damper repositioned. One situation was resolved by placing a lock washer under the
wing nut to hold the damper in positon. The operating crew Wa' '.hecking th' "ghtness
of the wing nut twice per shift in accordance with a January 30,1o98, de M *. The
other preblem was corrected by putting a tack weld on the "U" clamps ars .,tp i damper
shaft to the damper plate to keep the plate from rotating on the damper shaft. i his work
was accomplished by work order VAS-12711277.
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One flow problem remained unresolved. Flow induced vibrations on damper VHV324 will
require an up stream flow straightener to be installed to correct the vibration situation.
This is an inspector follow-up item (50-155/97015 01(DNMS)).
c. Conclusion
Installation of a HEPA filter in the reactor building ventilation was completed in December
1997, satisfying a PSDAR commitment and bringing the ventilation system up to
compliance with the FGEIS. The ventilation system now has filtration for containment
exhaust air.
11. 1 . 3 SpanLEnel Pool HagippRgle
a. Hopp3
The inspector reviewed the results of TV 59,' Fuel Pool Heatup Tests," to assess
whether decay heat in the pool has declined at the expected rate,
b. Obscrvations and Findinas
On January 28,1998, TV 59, * Fuel Pool Heatup Tests," was performed. Flow was
stopped for 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br />, and the SFP heated up from 50' F to 60' F; this is a heatup rate of
0.526' F/hr. The calculated time for the SFP to heat from 80' F to 150' F was 133
hours. The time to heat up from 50' F to 150' F was calculated as 190 hours0.0022 days <br />0.0528 hours <br />3.141534e-4 weeks <br />7.2295e-5 months <br />. The SFP
is currently averaging 50' F.
The importance of the length of time for the spent fuel pool (SFP) to heat up to 150' is to
have significant time for the licensee to restore SFP cooling if SFP cooling is lost. The
licensee is committed to restore SFP cooling before SFP temperature exceeds 150'so
that the concrete structure of the SFP does not degrade.
The results of TV 59 to date are as follows:
- November 22,1997 0 60' F/hr
- December 20,1997 0.b6' F/hr
- January 28,1998 0.53' F/hr
c, Concj9319D
The time for the SFP to heat up from 80' F to 150' F was verified by actual testing to be
more than the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> assumed in the licensee's submittal to the NRC.
11. 1 . 4 Steam Drum level Indication in the Control Reg
a. S.9222
The inspector observed that the steam drum levelindication in the control room was not
functioning, raising a concern that the control room operators didn't have adequate level
indication to monitor for leaks from the reactor coolant system.
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b. Observations and Findinas
On January 27,1998, the steam drum level Indication in the control room was not
working. The level reference legs had not been refilled after chemical decontamination
flushes and there was a slow leak from the reference leg. The chart recorder level had
been drifting up as the levelin the steam drum had been slowly dropping because the
reference leg was draining. Over a 4 day week end, the operators were without control
room steam drurr,levelindication. The reactor was empty of fuel, but the steam drum
was full of radioactive water. The operators had no chart trend indication or permanent
chart recording of the reactor coolant system level. However, operators were logging a
local pressure and RCS pressure as attemate level monitoring indications. The inspector
discussed this situation with the SS, maintenance manager, acting operations manager,
and the plant manager was assured that the control room steam drum level would be
restored.
On January 28,1998, the steam drum levellndication was filled and vented, and drum
level indication was restored in the control room. Steam drum level was again being
recorded on the level recorder. A daily order was written to communicate to all crews
how to maintain steam drum level and where to maintain the level.
c. Conclusion
The operating management became lax in maintaining steam drum level indication in the
control room after the reactor was defueled. After the inspector held discussions with
plant management, the steam drum levelindication was restored to service. The
operations and maintenance staffs may tend to neglect potentially important
instrumentation once TS no longer require its operation.
11. 2 Operational Safety Verification (71707)
11.2.1 General
The inspectors conducted frequent reviews of ongoing decommissioning activities.
Specific events and findings are detailed in sections below.
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11. 2 . 2 Lepk from CarA9nktpel Piece oLPicina Durina Chemical Decontaminall9n Procou
a. Sspp_t
The inspector reviewed the circumstances that led to the identification of a seven gallon
per minute (gpm) leak from the cleanup system.
b. Observations and Findinos
On January 1,1998, beginning at 11:00 p.m., the primary coolant system leakage
increased from 0.37 gpm to 1 gpm, then to 4 gpm. Water was found running out of the
cleanup heat exchanger room. Operations isolated the cleanup system from the primary
system.
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The operators entered the heat exchanger room, removed lagging, and determined that
sample valve VC 2 was leaking through. The operators closed VC 14 (sample line
isolation valve) stopping the leak. Just prior to isolation, the leakage reached seven gpm.
The caute of the leak was failure by operations to identify a bypass line in the non-
regenerative heat exchanger with a sample valve had a stainless steel to carbon steel
reducer. A known smallleak through the threads of the carbon steal reducer was not well
communicated to the chemical decontamination team. The leak caused a flow path for
the hot flush chemical, which strongly attacks carbon steel. The chemical hot flush ate
the carbon steel threads away, resulting in the seven gpm leak.
c. Q9D0l uf]9D
'
The inspector concluded that a leaking valve allowed flow through an unidentified carbon
steel reducer, which was eaten away and resulted in a seven gpm leak during the
chemical hot flush.
II.3 Qgeommissionino Performance and Stojus Review f71801)
11. 3 . 1 Q&att.ej
The status of decommissioning, and licensee and contracted workforce conduct of
decommissioning activities in accordance with licensed requirements and crmmitments,
were evaluated. Control and conduct of facility decommissioning was exart ed to verify
the license and TS requirements and commitments described in the Final Hazard
Summary Report (FHSR) and PSDAR, were being met and implemented. Specific events
and findings are detailed in the section below.
a. Scope
Plant tours were performed by the inspectors to evaluate the materialintegrity of
structures, systems, and components necessary for the safe storage of spent fuel and
conduct of safe decommissioning, and to evaluate plant housekeeping and fire
protection.
b. Observations and Findinog
,
I
After conducting plant tours the inspectors brought minor concems to the plant manager's
l
attention. The plant staff then later resolved these concems. The inspector also
observed that plant management was actively monitoring plant housekeeping and fire
protection. When issues arose they were resolved. I
1
c. Conclusions
The inspector concluded that the material integrity of structures, systems, and
components necessary for the safe storage of spent fuel and conduct of safe
decommissioning was being maintained. Plant nousekeeping and fire protection were
generally good.
14
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111. Spent Fuel Safety ,
i
Ill.1 Qeneral J
,
The inspectors evaluated the performance or condition of structures, systems or I
components associated with storage, control, and maintenance of sperit fuelin a safe
manner.
111. 2 Spent Fuel Pool Spletv at Permanentiv Shut Down Reactors @001)
a. Scope
The inspect!on evaluated spent fuel and fuel pool safety. Factors considered in the
evaluation included: siphon and drain protection; SFP instrumentation, alarms, and !
leakage detection; SFP chemistry and clean lIness control; criticality controls; and SFP ,
operation and power supply, t
b. Observations and Findinas
The only connection that can drain the fuel pool by siphoning is the cooling discherge
pipe, which retums water to the northwest comer of the fuel pool. The below surface
discharge was designed to minimize surface disturbance and improve visibility. To
prevent siphoning the water out of the pool, throo 7/16 inch holes were drilled just below
the water surface in the pipe. The licensee did not have any surveillance procedure to
document the condition of this siphon break. The licensee agreed to look into this issue.
Spent fuel pool instrumentation consists of level and temperature. Level is displayed in
the control room and recorded twice during a shift. Temperature indication is provided
locally at the S&Phooling water pump discharges, which is also recorded twice a shift.
These are adequaw to assure the safe storage of spent fuel. The licensee is planning to
provide level, temperature and new alarms for the new " control monitoring station".
Fuel pool leakage is monitored by eight tell tale drains from the area between the ;
'
stainless steelliner and concrete. The tell tale leakage is recorded each shift with zero
leakage detected historically. There are severalindependent sources of makeup water to ,
the pool.
The Inspector observed that water chemistry and cleanliness controls were adequate.
The licensee maintains a cover over the pool. Criticality concems do not exist at this time
because fuel will not be moved until approved storage casks can be obtained.
Electrical power to req'uired systems has been supplied from essential buses since initial
design, and there are redundant supplies of water. Either the electrical or diesel fire
pumps can supply service water to the c, cling water heat exchanges if normal, non-
essential power is lost to the service water pumps.
The power supplies to the equipment used to cool the SFP are as follows:
e the primary source is the 138 Kv line, through the station transformer to the 1 A
bus and 2A bus
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.
e the secondary source (following an automatic transfer) is the 46Kv line, througn
the station transformer, to the 1 A bus and 2A bus
e if neither the 138 KV nor the 46 KV line are available, the emergency diesst
generator automatically starts and energizes the 2B bus; after the 2B bus is
energized, the 1 A or 2A bus will be manually aligned to supply power to one of the
SFP pumps
e if for any reason the above three sources are unavailable, then the standby diesel
generator can be aligned to the 2B bus -
e the pumps required for cooling the SFP are spent fuel pit pump No.1 or No. 2,
reactor cooling water pumps No.1 or No. 2, and service water pumps No.1 or
No. 2. These pumps are all powered through the same primary and attemate
sources; the No.1 pumps are powered from the 1 A bus, and the No. 2 pumps are
powered from the 2B bus,
c. 9J2DgM!qu
The inspector concluded that the licensee can readily accommodate various conditions
that may challenge fuel pool level or cooling.
111. 3 Cold Weather Preparations (71714)
During a containment tour, the inspectors questioned the licensee on their contingency
plans for heating containment in the event that the main boiler is lost. The inspectors
were told that there was no contingency plans to supply additional heat to the
containment. The licensee stated that it would be good practice to have some sort of
program in place to maintain the temperature in the containment and/or respond if normal
heating is lost, and they agreed to waluate possible approaches to prevent the freezing
of systems which support the integrity of the fuel.
111. 4 Fuel Pool Activities (60710)
a. Scope
Review of the SFP cable log (which lists the weights and doses of items hanging from the
SFP rails) review of Field Monitoring Report FM-B-97129, and discussions with NPAD
inspectors, raised a concem that someone could inadvertently pull one of the high dose
items to the surface and receive a very high exposure,
b. Observatiqas and Findinas
Cables with high dose items hanging from them are hooked over the SFP rails. The
weight of the items varies from 12 pounds to 100-pounds. The dose of the items varies
from 3 R/hr to 1700 PJhr at one foot.
The potential for a high exposure from an individual pulling a high-dose item from the pool
was identified by NPAD and documented in FM B 97129. At the time of the inspection,
,
however, the licensee had not taken positive measures to prevent a person from
16
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.
Inadvertently pulling one of these items to the surface and receiving a very high exposure.
The inspector discussed this concem with plant management.
Management took immediate corrective action to place the SFP cover on the pool making
these cables inaccessible. On January 30,1998, a daily order was written stating the
covers need to be placed back over the SFP as soon as possible after any work is
performed that reduces removal of the covers. Positive controls for times when the SFP
covers must be removed and SFP activities resumed remain to be instituted. The
licensee's resolution of positive control over these cables in an inspector follow up item
(50155/97015 02(DNMS)).
c. GQncly119D
Positive controls are needed on high dose item suspended on cables in the SFP, to
prevent individuals from receiving a high exposure. The inspector communicated this
concem to plant management who had the covers put on the pool as a temporary
solution.
IV. Radiological Safety
IV.1 QQgypational Radiation Expp;ure (83750)
IV,1.1 General
Numerous aspects of licensee processes to mini.nize occupational radiation exposure
were selectively examined in order to evaluate overall radiation safety and to provide for
early identification of potential problems. Areas examined included: audits; planning and
preparation; extemal exposure control; intemal exposure control; control of radioactive
materials and contamination; surveys and monitonngs; and maintaining occupational
exposure as low as reasonably achievable (ALARA). Specific events and findings are
detailed in the sections below.
IV.1.2 Dose Monit9dag _urina Chemicpl Decontamination Process (837@)
a. Scop _q
The inspectors attended the pre-job and At. ARA briefings reviewed the RWP (B982035)
and Pre-job briefings signature sheet for January 4,1998, and held discussions with the
HP members of the HP department and the plant management.
b. Observatl0D! Lend Findinas
The inspectors observed the pre-job and Al. ARA briefing for the Chemical Decon (High
Radiation Area) RWP 8982053. During this briefing monitoring dosimetry and ensuring
that electronic dosimeters (ELD) were not on " pause" were discussed at length and were
stressed.
Daily, the RWP and pre-job briefing were reviewed with the workers and signed by them.
The written pre-job briefing requires that the worker monitor dosimetry every 15 minutes
when in a high radiation area.
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On January 4,1998, s' chemical decontamination worker read and signed RWP B982035
and was in a HRA for abou! 55 minutes.-
When the worket tried to log the ELD out, it was discovered that the ELD was on " pause,"-
l which was reported to the licensee; When the licensee interviewed the worker it was
learned that he had not checked his dosimetry every 15 minutes as required. Later, the
licensee determined that the workers ELD was faulty.- The worker was assigned a dose
of 6 mrom for the time he was in the HRA No exposure limits were exceeded.
The inspectors discussed this incident with the radiation protection manager and the plant
. manager, both of whom indicated that this behavior was not consistent with management
expectations for health physics practices. Plant management had counseled the worker
and his supervisor. Plant E-mail messages were sent to plant workers stressing the - o
need to ensure that ELDS are r'ot on : pause", i
The licensee's immediate corrective actions included the following:
.- The HP log-in book at the sphere will include the ELD reading in and out of the
sphere.
. The chemical decon supeivisor will review team member's ELD readings prior to
entry to the HRA at the chemical decontamination equipment area.
. The HP supervisor will tour a minimum of 3 times a shift, specifically reviewing
workers compliance with reading the ELD every 15 minutes while in a HRA.
. The ELDS and TLDs must be in clear plastic bags that allow monitoring of the
ELD reading while in a HRA.
. Briefing will remind workers of the expectations and requirements of monitoring
the ELD's reading once or twice an hour while in a radiation area and once every
15 minutes when in a HRA, This is a procedure requirement.
The licensee has had previous problems with workers being in the RCA with ELDS on
" pause" and the corrective actions and pre-job briefings did not prevent this incident. ,
Technical Specification 6.11 requires, in part, that procedures for personnel radiation -
protection shall be adhered to for all operations involving personnel radiation exposure. -
The written Pre-Job Briefing for Radiation Work Request (RWP B982035), and Procedure
5.5, " Radiation Work Permit," (Revision _12), instruct workers to monitor dosimetry every-
15 minutes when in e high radiation area. As described above. on January 4,1998, a
chemir.at decontamination worker was in a high radiation area (HRA) for greater than 15
minutes and did not inonitor his dosimetry. The worker was in the HRA about 55
minutes. Tnis is considered a violation (50-155/97015-03(DNMS)).
t
c. Conclusions l
A violation of Technical Specifications, concoming a decontamination worker in a HRA
not monitoring his dosimetry as required, and not noting that his ELD was on " pause,"
18
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Indicated that worker attention to radiatio 1 protection practices (especially in the area of
monitoring dosimetry), remained a problem for station perroml.
IV.2 Maintainina occupational Exposures ALARA (83750)
a. Scgge
The inspector reviewed the ALARA program for the chemical decontamination operation.
b. 90servations And Findinos
The chem decon of the reactor system and associated piping was done to reduce dose to
workers during plant dismantlement and represented a major Al. ARA initiative. The
licensee instituted several practicet to minimize dose since the solubilization and
transport of large quantities of isotopes in the reactor piping system would create the
potential for elevated radiation areas during the chem decon process. The licensee's
ALARA plan for dose minimization during the chem decon project was based on the
proven health physics practices of time, distance and shielding.
Temporary shielding in the form of standing water shields and lead blankets was
employed. Water shields were used to surround the ion exchange columns, lead
blankets were used in work areas as well as in localized and transient high radiation
areas. Low dose waiting areas (LOWA) were established for workers in the RCA during
pre-work activity and for equipm nt staging. Remote monitoring equipment includad area
radiation monitors (ARM), cameras and teledosimetry,
c. Conclusion
The chem decon project included appropriate dose-minimization planning, and effective
ALARA principles were implemented.
IV.3 Radiation Protection. Plant Chemistry and Radwaste(83722)
a. Scope
The inspetor reviewed the re 11ation protection (RP) program for the chemical
decontamination operation.
b. Observations And Findinas
The licensee instituted positive control over access to the containment sphere
during the chem decon process. Only personnel that were needed in the work
areas of the sphere were admitted. Access was controlled by use of a control
point outside of the sphere, a logbook, an access authorization list, locked rooms,
physical barricades, and radiation work permits (RWPs). Area radiation monitors
(ARMS) were placed in locations where elevated and changing dose rates would
be exp'cted from the transport of isotopes and crud bursts during the process.
Some areas where ARMS were located included the T-31 primary ion exchangers,
particulate filters, and the secondary lon exchangers. Significantly elevated dose
19
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rates were expected from the T-31 primary ion exchangers, in addition to the
water shielding, these resin beds were located in an inaccessible area which was
posted and barricaded.
Change out of the particulate filters was perfo:Tned by hand instead of remote
equipment. The licensee and vendor evaluated dose rates with the filter housing
open and compared the time required for experienced technicians to remove the
filters against the time necessary to remove the filters by remote equipment. They
concluded that a manual filter change od would result in a lower total dose. The
initial filter change out resulted in a dose of only about 3g millirem to the technician
performing the job.
c. Conclusion
The licensee's radiation protection prograta for the chem decon process was well
planned and executed. Planning and communication, between licensee personnel
and the vendor performing the chem decon, were very good as was eviden.:ed by
the ALARA program, execution of the RP program during the process, and
adherence to the work schedule. .
V. Management Meeting
The inspectors presented the inspection results to members of licensee management at
the conclusion of the inspection on February 2,1998. The licensee acknowledged the
findings presented. The licensee stated that documents and processes associated with
- the chemical decontamination were considered proprietary; however, the inspector
described the expected contenta of the report and no use of proprietary information was
identified.
1
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
K Powers, General Manager
D. Hice, Plant Manager
K Pallegl, Radiation Protection and Environmental Manager
W. Trubliowicz, Scheduling and Project Control Manager
G. Withrow, Engineering Manager
D. Debner, Nuclear Fuel Projects Manager
C. Jurgens, Construction Manager
K Wooster, Emergency Planning Manager
M. VanAlst, Security Manager
L. Potter, Malntenance Supervisor
L. Darrah, Operations Supervisor
G. Hausler, Work Control Supervisor -
D. Lacroix, Training Supervisor
G. Rowell, Corrective Action Supervisor
M. Bourassa, Licensing Supervisor
G. Petijean, Senior Engineer (Altemate PRC Chairman)
E. Zionert, Human Resoure.es Director
T. Mosley, Senior Engineer Reactor Protection Projects
J. Corley, Nuclear Performance Assessment
L. Berry, Site Licensing Engineer
,
21
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INSPECTION PROCEDURES USED
36801- Organization, Wlanagement, and Cost Controls
37801 Safety Review, Design Changes, and Modifications
42700 Plant Procedures
60710 Fuel Handling Activities
60801 Spent Fuel Pool Safety at Permanently Shut Down Reactors
62801 Maintenance and Surveillance at Permanently Shutdown Reactors
71707 operational Safety Verification
71714 Cold Weather Preparations
71801 Decommissioning Performance and Status Review
-71801 - Decommissioning Performance and Status Reviews at Permanently Shut
Down Reactors
83722 Radiation Protection, Plant Chemistry and Radwaste
83750 Occupational Radiation Exposure
83750 Dose Monitoring During Chemical Decontamination Process
83750 Maintaining occupational Exposures ALARA
.92700. Onsite Follow-up, Written Reports of Non-routine Events at Power Reactor ~ ;
Facilities
22
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4
ITEMS OPENED AND CLOSED
QR10.td
.
50-155/97015-01 IFl - Licensee's resolution of flow induced vibration on damper
VHV 324
50-155/97015-02 IFl Licensee's resolution for positive control over cables
hanging in the SFP
50 155/97015-03 VIO Worker failed to monitor dosimetry while in a high radiation
arca
Reviewed but Remains Open
50-155/96002 VIO Control of combustible materials during outage related work
'
Closed
50-155/96006-03 IFl Sufficient core spray flow
- 50-155/9007-02 IN Delayed dispatch of operationa' support center emergency
response team
1
23
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_ _ _ _ _ _ _ ___- ____-_ ________ .
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LIST OF ACROK'c4S USED
ALARA As Low As Reasonably Achievable
ARM Area radiation monitors
CFR Code of Federal Regulations
DNMS Div!sion of Nuclear Materials Safety
EDG Emergency diesel generator
ELD Electronic dosimetry
FEMA Federal Emergency Management Agency
FGEIS Final Generic Environment Impact Statement
FHSR- Final Hazard Summary Report
FM Field Monitoring
FSAR Final Safety Analysis Report
HEPA High Lificiency Padiculate Activity H5
HP Health Physics
IFl inspector Follow-up item
LDWA Low dose waiting m 's
LTP License Terminatic lan
NPAD Nuclear Performar. Assessment Department
NRC Nuclear Reguls ory commission
"CM Offsite pose Calculation Manual
'
-Pct Public Document Room
PRC Piart 9afety Review Committee
PSDAR Post-Shutdown Decommissioning Activities Report
QA Quality Assurance -
Radwaste Radioactive Waste
RP Radiation protection
RWP Radiation Work Request (RWP-Bg82035)
SAFESTOR Placing and maintaining a nuclear facility in a condition to safety store fuel
SFP- Spent Fuel Pool
SS Shift supervisor
SSCs Structures, Systems, and Components ,
'
TLD Thermal Luminescent Dosimeter
TS Technical Specification
TV Test Variable
USC Unreviewed Safety Question
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24
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