IR 05000309/1998003
| ML20154A922 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 09/28/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154A908 | List: |
| References | |
| 50-309-98-03, 50-309-98-3, NUDOCS 9810050043 | |
| Download: ML20154A922 (13) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-309 License No:
DPR-36 Report No:
50-309/98-03 Licensee:
Maine Yankee Atomic Power Company (MYAPC)
Facility:
Maine Yankee Atomic Power Station
- Location:
Bailey Point Wiscasset, Maine Dates:
May 3,1998 to August 1,1998 Inspectors:
Mark C. Roberts, Sr. Health Physicist William J. Raymond, Senior Resident inspector
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Todd J. Jackson, Health Physicist Steve W. Shaffer, Health Physicist
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Approved by:
Mark C. Roberts Chief, Decommissioning and Laboratory Branch Division of Nuclear Materials Safety
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9810050043 980'r28
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PDR ADOCK 05000309 G
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EXECUTIVE SUMMARY Maine Yankee Power Station NRC Inspection Report No. 50-309/98-03 This integrated inspection included aspects of licensee operations, engineering, and plant support. The report covers a three-month period of inspection, including results of announced inspections by three regional inspectors and a resident inspector from the Haddam Neck site.
Conduct of Ooerations
Status and condition of the plant were reviewed. Major activities during the report period included spent fuel pool modifications, radiological control area and security boundary modifications, and removal of asbestos insulation from plant components.
Soent Fuel Safetv:
Licensee performance was good to implement the nuclear island modifications per the administrative controls. Engineering calculations and technical evaluations supported the
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modifications. An open item will track NRC review of the conclusions regarding the analysis of the effects of a heatup event on the spent fuel pool structure.
Plant Suooort and Radioloalcal Controls:
Weaknesses were identified in the control of licensed materials and locked high radiation areas. Licensee corrective actions for both issues were appropriate and thorough.. The NRC exercised enforcement discretion in not citing these matters. (NCV 50-309/98-03 02)
Liquid effluents for January-April 1998 were reviewed and found to be within limits. An aggressive chemical decontamination of the reactor coolant system was performed during the period reviewed.
' An onsite area of contamination was reviewed and found to be adequately controlled by the licensee. The contamination had resulted from valve leaks in 1988. Records describing the area were available, and the remediation of the contamination will be addressed during the decommissioning of the site.
Use of a bag monitor to survey trash during the late 1980s was reviewed to determine if significant radioactivity had been released in trash classified as " clean". The licensee investigated the use of the bag monitor, including instrument capabilities, records, and conducted interviews of personnel. The licensee's investigation was found to be thorough and complete. The inspector concluded that there is no basis to believe that significant radioactive material was released in trash during the period in question.
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TABLE OF CONTENTS
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EXECUTIVE SUMMARY
................................................ ll TAB LE OF CONTENTS................................................. iii
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- Report Details.......................................................
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Ope rati o n s......................................................
O1 Conduct of Operation s............................................
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01.1 General Comments (71707)..................................
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II. Spent Fuel Safety..................................................
1 E1 Engineering Support of Decommissioning Activities......................
E1.1 Facility Modifications........................................
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Ill. Plant Support and Radiological Controls.................................
R1 Radiological Protection Controls....................................
R1.1 Control of Radiation Areas and Radioactive Materials................
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R1.2 Llauld Effluent Monitorina....................................
R1.3 Control of a Known Contaminated Outside Area....................
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i-R8 Miscellaneous Radiation Protection issues.............................
R8.1 Previous Controls for Release of Trash from the Site................
IV. Man ageme nt Meetings.............................................
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'X1 Exit Meeting Summary...........................................
PARTIAL LIST OF PERSONS CONTACTED.................................
- INSPECTION PROCEDURES USED.......................................
d ITEMS OPEN, CLOSED AND DISCUSSED..................................
LI ST O F AC R O N YM S U S E D............................................
Attach me nt I........................................................
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Report Details l
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Operations j
Conduct of Operations
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01.1 General Comments (71707)
The inspectors conducted reviews of ongoing plant operations. Inspectors attended the licensee's plan-of-the-day meetings and observed the coordination of decommissioning activities. Operator conduct in the control room was observed to be
- appropriate, inspectors toured site buildings and observed work activities in progress.
Site work by the licensee during this period focused on modifications related to the
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Spent Fuel Pool Island (SFPI), moving radiological control area (RCA) boundaries, security modifications, and asbestos insulation removal. No safety concems or violations were identified.
11. Spent Fuel Safety E1 Engineering Support of Decommissioning Activities
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E1.1 Facility Modifications a.
Inspection Scope (37801. 60851. 60854. 37700)
The inspector reviewed modification activities that supported spent fuel safety. Plant design change activities to establish the nuclear island were reviewed to verify that the modifications were implemented in accordance with licensee controls.
b.
Observations and Findinas The modifications associated with design change package (DCP) 97-42 were reviewed (through ECN 3), along with supporting engineering calculations. The list of references and materials reviewed is provided in Attachment 1. DCP 97-42 installed new equipment for the SFPI that eliminated reliance on the existing plant cooling water systems. The modifications included:
The use of the existing spent fuel pool (SFP) heat exchanger (E-25) and fuel
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pool cooling pumps (P-17A and B) with a new secondary decay heat removal (DHR) system that used exterior fan coils to disperse heat to the outside air.
The existing heat exchanger inlet and outlet hose connections were maintained to allow for the emergency supply of cooling water from the fire water system in the event of the loss of the normal cooling system. SFP purification and makeup were provided. Normal makeup of coolant is provided by the existing 150,000 gallon primary water storage tank (PWST) and primary water pump P-SFP2 installed in the spent fuel building (SFB). An underwater demineralizer was installed to replace ion exchanger I-4.
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d The DHR cooling system used a series of six parallel, fan-cooled, finned-coiled
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water to air coolers. The secondary cooling loop was designed to use water
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during the summer of 1998, with a change over to ethylene glycol-water in October 1998. The two redundant capacity DHR pumps (P-SFP1A and SFP18) each supply a nominal 1000 gallons per minute (gpm) flow. The fan coolers are staggered to operate as the heat rejection rate changes with air
temperature. The fan coolers were located outdoors in a diked area with a gabled roof. The six cooler units each have three fans (total of 18) operating at 1800 revolutions per minute (rpm). During startup testing, a high noise condition existed when all fans were operating. Licensee actions to consider noise abatement design changes continued at the end of the inspection period
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The SFPI electrical distribution system was provided by 480-volt power,
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stepped down from the tertiary winding of the exist 5g reserve station services
transformer X-16, and two outdoor unit substations with 2500 kVA transformers. One unit substation was designated Bus-SFP, which supplied motor control center (MCC) MCC-SFP1 (existing MCC-11B) and the new MCC-SFP2. A backup power supply is provided by a 250KW,480 voit emergency diesel generator. The diesel is manually aligned and is mounted on a trailer with a 1500 gallon fuel tank. The diesel is intended to power manually selected DHR loads, instrumentation and other miscellaneous loads for the nuclear island. Technical evaluation 14-98 was written to change the diesel
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i rating to 200 KW (240 amps with unity power factor), which was acceptable based on the required load of about 170 KW.
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The licensee made several system enhancements and addressed design
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i deficiencies: a siphon break was installed on SFP cooling lines; E25 was
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qualified to higher temperature to support SFP passive cooling, and a thermal 4-relief valve was added; and, calculations and procedure changes were made to
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address P-17A and B net positive suction head (NPSH) issues.
NRC review focused on the mechanical and electrical portions of the SFPI modifications (DHR makeup and cooling system, ventilation system). NRC review of other modifications continued at the end of the inspection, including the changes to the SFB ventilation system; installation of a new SFB ventilation system radiation monitor; new security system modifications, the testing of the programmable logic controller (PLC); and, the installation of the new control room.
The inspection verified the design control program was appropriately implemented, and that onsite fabrication activities were conducted in accordance with the associated work packages. The DCP was developed and implemented per the administrative controls in the Engineering Manual. Inspector observations of as-built installations and activities in progress, and a sampling review of work packages showed the
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modifications were well controlled and completed per the DCP packages. Plant
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operational procedures were changed as required (e.g., the NPSH limitations on the i
SFP cooling pump operation). The licensee revised the Offsite Dose Calculation
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Manual (ODCM) to incorporate the Fuel Building exhaust vent as a new gaseous
effluent pathway (reference ODCM Change No. 98-05).
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Functional testing of the new systems was completed per DCP 97-42 and functional test instructions were controlled using the work order process. The testing included flushing of the DHR system and performance testing on spent fuel cooling prior to the i
final tie-in; functional testing of the control circuits for the SFP fan cooler banks and j
the circulating pumps P-SFP1A and P-SFP1B; startup and performance testing of the
PLC for the control room monitoring, alarms, displays, and operator interface controls.
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The supporting calculations and technical evaluations reviewed for DCP 97-42 are listed in Attachment 1. The safety evaluations completed per 10 CFR 50.59
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supporting the DCP appeared to be completed per the administrative controls and
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were consistent with the UFSAR and design work to declassify plant systems. The j
calculations appeared to use acceptable methodologies and conclusions were
supported by the results. The analyses (Calculation 73-9) and testing (SFP Heatup
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Rate Test per Procedure 4-17-23) determined the pool thermal performance and
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provided the basis for the sizing and operational thermal requirements of the forced j
air coolers.
j Calculation MYC 2001 analyzed the pool for an assumed loss of cooling event in i
which the pool would boil. The licensee concluded that the stresses on the pool structure will remain within the ACI code limits under worst case load conditions,
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which the licensee defined to include deadweight, hydraulic and thermal loads. The
analysis predicted some cracking on the outside surface of the walls, but that the leak
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tightness and structural integrity would not be affected. The methodology appeared
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acceptable, however, an open item will track the NRC review of conclusions regarding j
the acceptability of a heatup event (boiling) on the pool structure (URI 98-03-01).
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Conclusions
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In general, licensee performance was very good to implement the nuclear island modifications per its commitments and administrative controls. Engineering calculations and technical evaluations supported the modifications. An open item will track NRC review of the analysis of the affects of a heatup event on the pool structure.
lli. Plant Support and Radioloalcal Controls R1 Radiological Protection Controls
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R1.1 Control of Radiation Areas anc' Radioactive Materials a.
Inspection Scope (83750)
The inspector toured the radiological control area (RCA), observed the posting and control of high radiation areas (HRA), reviewed the controls for special nuclear
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materials, and reviewed the licensee's corrective actions associated for a loss of locked HRA door controls, b.
Findinas and Observations
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Locked Hioh Radiation Area (LHRA) Controls
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f The licensee discovered at 9:30 a.m. on May 13,1998 that a door at the entrance to
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the LSA Building was closed but not locked. The licensee had controlled the entrance i
to the building as a " locked HRA" per Technical Specifications 5.12 (TS 5.8 in the defueled technical specifications) because dose rates inside the room were greater
than 1000 mrem /hr. The door had been unsecured for 15 minutes after a worker j
went through the door to enter the building, and then left by another exit. The door
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was not secured because the latch required lubrication and the worker had not verified it was latched prior to exiting the area.
The licensee took immediate actions to verify no unauthorized personnel were in the area, and that all TS LHRA doors were locked. The need to assure proper control of HRA doors was discussed with the worker and other plant personnel. The event was described in Condition Report 98-136. The licensee determined that a program to periodically lubricate and adjust LHRA doors had lapsed; an assignment was made to Plant Services to reinitiate the preventive maintenance program. Licensee corrective actions were timely and appropriate.
The inspector observed the LSA building door on May 21 and during the week of June 22, and noted that the lock was secured and that the door closing mechanism operated well. The inspector and a licensee HP technician conducted a survey of accessible areas in the room on May 21. No radiation dose rates In excess of 1000 mrem /hr were identified. There have been no similar occurrences at the site since 1995.
Control of Licensed Materials The licensee identified on May 20 that source materials were not locked as required by procedure. The licensee had stored the sources in an unlocked caged area, and relied on the lock door controls for the new fuel vault to control access to the sources.
However, the licensee had removed all new fuel from the vault and stopped locking the room in May 1998 to facilitate construction activities in the area.
The inspector toured the area on May 21 and noted that the sources had been retumed to a locked storage area. The inspector also observed source range monitor detector cables that were not locked. The detectors contain special nuclear matenal.
The licensee established locked contrc;s for all licensed materials in the room as of 5:30 pm on May 21, pending further review of regulatory requirements. The Plant Manager subsequently reported that, as a conservative measure, the licensed materials would be kept locked. The inspector reviewed regulatory requirements regarding the storage and control of licensed materials inside the restricted area; no violations were identified.
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Conclusions Licensee corrective actions for both the LHRA and licensed material issues were appropriate and thorough. This non-repetitive, licensee identified and corrected
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violation is being treated as a Non-Cited Violation, consistent with Section Vll.B of the enforcement policy (NCV 98-03 02).
R1.2. Liould Effluent Monitoring,
a. Inspection Scope (84101)
Effluent data were reviewed regarding radioactive material in liquid discharges during i
the period January - April 1998.
b.
Observations and Findinas Liquid effluent monitoring data were reviewed for the period during which the licensee performed decontamination of the reactor coolant system (RCS). Liquid wastes to be discharged from the plant are collected in a radwaste test tank and sampled prior to release. Analysis of the test tank samples is used to determine tank discharge rate,
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and to determine in advance what will be the actual discharge concentration. This assures that planned discharges will be in compliance with discharge limits. The
service water radiation monitor is downstream of the test tank discharge line point of addition to the service water piping and provides confirmation that excessive radioactive materials are not present in the service water discharge.
The inspector reviewed the monthly totals of test tank batch sample analyses for
' January, February, March, and April,1998. During this period the licensee performed
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an aggressive chemical decontamination of the RCS and processed all the water used
during the decontamination, c. Conclusions Liquid discharges for the period were within regulatory limits. The data will be included in the licensee's effluent release report covering the period reviewed.
R1.3 Control of a Known Contaminated Outside Area a. Inspection Scope (83726)
The inspectors toured the site and examined the condition of the exterior of the plant.
b.
Observations and Findinas The inspectors observed an area outside of the RCA that was posted as a radioactive materials area. Upon discussions with the licensee it was determined that the area had soil contaminated primarily with cesium-137 in the picocurie per gram (pCi/g)
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range. The contamination is from plume migration or surface runoff from two previous
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events involving the refueling water storage tank (RWST). The licensee discovered the contamination outside the RCA during site characterization work.
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In February 1988, the licensee discovered a leak at a flange connection between the RWST siphon heater retum line and icolation valve CS-81. Upon discovery, the
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licensee contained the spilled material and repaired the leak. Surveys revealed that the ground in the area was contaminated, and the licensee excavated the contaminated area. During the excavation, a second leak was discovered at the base
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of the RWST siphon heater retum line isolation valve CS-81. The second leak, significantly smaller than the first leak, was promptly repaired. Soil excavation continued until approximately 800 cubic feet of soil had been removed, at which point
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the excavation was stopped due to concems that the foundation of the RWST was
being undermined. The excavated area was then backfilled with clean soil and repaved.
in a letter dated November 2,1988, the licensee requested approval from the NRC for in-place disposal of the remaining contamination under 10 CFR 20.302(a). On August 31,1989, the NRC granted the licensee's request. This area is included in the
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scope of work for the decommissioning of the facility. The licensee described the
plans for maintaining control of the contamination during decommissioning of the site.
The contaminated area will remain posted, and the licensee will cordon off and limit access to the area. The licensee will take soil samples on a regular basis in order to
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precautions necessary to prevent spread of the contamination in the limited use dirt j
road that runs through the contaminated area, c. Conclusions The licensee has fully documented the presence of the contamination in the area
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adjacent to the RWST, and the NRC had previously reviewed the circumstances of the spill and its disposition. This area will be included in the licensee's remediation efforts during the decommissioning of the Maine Yankee site. No new safety
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concems or violations were identified.
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R8 Miscellaneous Radiation Protection issues R8.1 Previous Controls for Release of Trash from the Site a. Inspection Scope (83726)
The inspector examined data and records regarding the historical use of a " bag monitor" to survey waste material prior to release from the site. The licensee had investigated the possibility that radioactive material was released in trash from the site
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during the 1980s in Condition Report (CR) 98-92.
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b.
Observations and Findinas The inspector reviewed the records and data related to the licensee's investigation of whether radioactive material could have been released from the site and disposed of in the Wiscasset landfill. The WCM-14 bag monitor was put into use in mid-1988, modifying the previous practice of passing bags of sorted trash through a personnel portal monitor. At the time, the bag monitor was implemented as an improvement to previous practice. The monitor was used as a final survey of materials which were not expected to be contaminated.
Some bag monitor source check records were found to be incomplete, and the licensee identified a period of time when it appeared that the bag monitor was used while the procedure for " Sorting of Primary Side Debris" specified use of the portal monitors. The licensee found other related instrument records to be complete and retrievable. The inspector noted that the use of the bag monitor from the time the bag
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monitor operating procedure was issued (April 1988) until the sorting procedure was revised (November 1988), was an improvement in the scanning methodokigy. This apparent administrative procedure problem would not have been expec'ed to increase the chance that contaminated packages would be released from the site, but would have been expected to reduce the chance of release because it was a monitmic improvement.
c. Conclusions The inspector found the licensee's investigation to be thorough and complete. The inspector observed thtst the licensee's records and procedures from the 1988-89 time frame appear consistent with contemporaneous NRC guidance and industry practice.
While it is not possible to reconstruct all records pertaining to releases, there is no basis to conclude that significant amounts of radioactive material were released from the site during the period when the bag monitor was used. In addition, the licensee and the NRC performed independent surveys of the Wiscasset landfi!! and found no measurable, plant-related radioactive material in the landfill (see NRC Inspection Reports 50-3C9/98-01 and 50-309/98-02).
IV. Manaaement Meetinos X1 Exit Meeting Summary The inspectors presented the inspection results to members of the licensee's staff on June 25,1998. The licensee acknowledged the findings presented.
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PARTIAL LIST OF PERSONS CONTACTED Licensee W. Odell, Operations Director R. Fraser, Engineering Director J. Sauger, Project Management and Construction W. Ball, Operations Manager G. Zinke, Director, Nuclear Licensing and Regulatory Affairs M. Ferri, Decommissioning Director Other P. Dostie, Maine, Nuclear Safety inspector INSPECTION PROCEDURES USED IP 37700:
Design Changes and Modifications IP 37801:
Safety Reviews, Design Changes, and Modifications at Permanently Shutdown Reactors IP 60851:
Design Control of ISFSI Components IP 60854:
Preoperational Testing of an ISFSI lP 71707:
Plant Operations IP 83726:
Control of Radioactive Materials and Contamination, Surveys and Monitoring IP 83750:
Occupational Radiation Exposure IP 84101:
Radioactive Waste Management ITEMS OPEN, CLOSLD AND DISCUSSED URI 98-03-01 Acceptability of a heatup event (boiling water) on the pool structure.
NCV 98-03-02 Weaknesses in the control of licensed materials and LHRAs.
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LIST OF ACRONYMS USED
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CFR Code of Federal Regulations CR.
Condition Report DCP-Design Change Package DHR
. Decay Heat Removal FTl Functional Test instructions gpm gallons per. minute
ISFSI Independent Spent Fuel Storage Installation LHRA'
' Locked High Radiation Area MCC Motor Control Center NCV Non-Cited Violation NPSH Net P3sitive Suction Head NRC--
Nuclear Regulatory Commission
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- Offsite Dose Calculation Manual.
pCl/g picoCuries per gram PDR Public Document Room PLC.
Programmable Logic Controller PWST Primary Water Storage Tank RCA RadiologicalControlled Area RCS Reactor Coolant System
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rpm revolt' ions per minute RWST
- Refue:ing Water Storage Tank SFB
' Spent Fuel Building SFP Spent Fuel Pool SFPI Spent Fuel Pool Island.
TS Technical Specification UFSAR Updated Final Safety Analysis Report -
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Unresolved item WO Work Order i
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Attachment i Spent Fuel Island Modifications The inspection of the design, modification and functional testing of the nuclear island included observation and/or review of design change packages (DCP), functional test instructions (FTis) and associated safety evaluations, and work orders (WO) on a sampling basis, including the following references:
' MODIFICATION PACKAGES DCP 97-42, Spent Fuel Pool isolation, ECN 1 - 3 TEST PACKAGES WO 97-3435-18, FTl of Spent Fuel Island PLC WO 97-3435-32, FTl of the Primary Cooling Pump P.
- ure Instrumentation WO 98-011-01, FTl of DHR System and Piping to the New Fan Coolers WO 98-0225-02, FTl of SFP Cooling Pump P-17A WO 98-0225-04, FTl of the Cooler Fan Banks and Circulation Pumps WO 98 027-00
- FTl of Fuel Building Heating and Ventilation System WO 98-057-01, FTl of the Completed DHR Piping System and FP Addition WO 98-332-00, SFPI Makeup Water System Test WO 98-545-00, FTl of the Backup Diesel Generator CALCULATIONS and EVALUATIONS MYC-2004, SFP T-H Calculation (11/97)
Calculation 78-97, SFP Cooling Pump Performance (NPSH)
Calculation 73-97, SFP Cooling system thermal analysis MYC-2005, SFP Passive Cooling (YNSD)
J MYC-2001, SFP structural Analysis for Passive Cooling Technical Evaluation 210-97, SFP Cooling Pump Operating Conditions Technical Evaluation 014-98, DG-SFP1 Maximum Operating Limit Change PROCEDURES ODCM Change No.98-05, Fuel Building Exhaust Vent as a New Effluent Pathway Procedure 0-06-4,10 CFR 50.59 Determinations Procedure 4-17-23, SFP Heatup Rate Test Procedure AOP 2-52, Loss of SFP Cooling, Revision 5,5/4/98 i
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