IR 05000213/1990002
| ML20012D330 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 03/14/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20012D327 | List: |
| References | |
| 50-213-90-02, 50-213-90-2, NUDOCS 9003270198 | |
| Download: ML20012D330 (22) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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Report No.
50-213/90-02 r
License No.
DPR-61 Licensee:
Connecticut Yankee Atomic Power Company
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P. O. Box 270 Hartford, CT 06141-0270
Facility:
Haddam Neck Plant Location:
Haddam Neck, Connecticut Dates:
January 9 to February 15, 1990 Reporting Inspector:
John T. Shediosky, Senior Resident Inspector Inspectors:
Andra A. Asars, Resident Inspector Walter J. Pasciak, Chief, Facility Radiation Protection Section John T. Shediosky, Senior Resident Inspector Approved by:
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Donald R. Haverkamp, C#ief Date Reactor Projects Section 4A
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Division of Reactor Projects Inspection Summary: Inspection on January 9 - February 15, 1990 (Inspection
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Report No. 50-213/90-02 Areas Inspected:
Routine safety inspection by the resident inspectors. Areas reviewed included plant operations, radiological controls during thermal shield removal, reactor vessel thermal shield removal, inspection and reconstitution.
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of reactor fuel, Appendix R service water performance testing, security, potential service water filter clogging following a LOCA, Plant Operations
. Review Committee activities, written reports, analysis of moderator dilution
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events and procurement of emergency diesel generator fuel oil.
i Results: This was the fourth routine resident inspection during the 1989/1990 Refueling Outage.
Significant progress was made in the design change to the reactor interals; the thermal shield was removed and stored in the refueling cavity. Preparations are being made for the secondary cutting of the thermal shield for shipment to the burial site (Section 4.1.1).
Inspection and reconsititution of the reactor fuel was completed (Section 4.1.2).
The 9003270198 900314 l
PDR ADOCK 05000213
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-licensee identified two unsatisfactory conditions in the service water system; inadequate flow to the emergency diesel generator heat exchangers-(Section 4.2.1) and a potential for service water filters to clog following a LOCA (Section 6.1).
Licensee analysis of moderator dilution events was reviewed (Section 8).
Procurement practices for emergency diesel generator fuel oil were verified (Section 9).
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TABLE OF CONTENTS Page r
1.
Summary of Facility Activities (71707)*..............
2.
Plant Operations (71707, 71710, and 93702)..............
I 2.1 Operational Safety Verification
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2.2 Engineered Safety Features System Walkdown..........
3.
Radiological Controls (71707 and 83750)..............
3.1 Routine Radiological Controls Verification..........
3.2 Radiological Controls During Removal of the Thermal Shield..
4.
Maintenance and Surveillance (61726, 62703, and 71707)
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4.1 Maintenance Observation
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4.1.1 Reactor Vessel Thermal Shield Removal.........
4.1 2 Inspection and Reconstitution of Reactor Fuel.....
4.2 Surveillance Observation...................
4.2.1 Appendix R Service Water Performance Test.
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5.
Security (71707)
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6.
Engineering and Technical Support (71707, 37700, and 37828)....
l 6.1 Potential Service Water Filter Clogging Following a LOCA...
7.
Safety Assessment and Quality Verification (40500, 71707, 92700, and 90712)............................
7.1 Plant Operations Review Committee
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7.2 Review of Written Reports
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8.
Analysis of Moderator Dilution Events (TI 2515/94).........
9.
Procurement of Emergency Diesel Generator Fuel Oil (TI 2515/93)
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10. Exit Interview (30703)
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- The NRC Inspection Manual inspection procedure or temporary instruction (TI) that was used as inspection guidance is listed for each applicable report section.
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I DETAILS
1.
Summary of Facility Activities i
The Fifteenth Refueling Outage continued during this inspection period.
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Major work activities included fuel inspection and reconstitution and
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removal of the reactor core support barrel thermal shield. All reactor
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fuel remained in the spent fuel pool.
Dr. Ivan Catton of the Advisory Committee on Reactor Safeguards visited
the facility on January 30, 1990. A tour of the plant was made with members of station management and the resident inspectors. Areas toured included the switchgear rooms, control room, containment, and emergency diesel generator rooms.
2.
Plant Operations 2.1 Operational Safety Verification The inspectors observed plant operation and verified that the plant was operated safely and in accordance with licensee procedures and regulatory requirements.
Regular tours were conducted of the follow-ing plant areas:
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control room security access point
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primary auxiliary building protected area fence
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radiological control point intake structure
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electrical switchgear rooms diesel generator rooms
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aux 111ery feedwater pump room turbine building
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l Control room instruments and plant computer indications were observed
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for correlation between channels and for conformance with technical specification (TS) requirements.
Operability of engineered safety l
features, other safety related systems and onsite and offsite power sources were verified. The inspectors observed various alarm condi-tions and confirmed that operator response was in accordance with plant operating procedures.
Routine operations surveillance testing was also observed. Compliance with TS and implementation of appro-priate action statements for equipment out'of service was inspected.
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Plant radiation monitoring system indications and plant stack traces
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were reviewed for unexpected changes.
Logs and records were reviewed to determine if entries were accurate and identified equipment status or deficiencies. These records included operating logs, turnover sheets, system safety tags, and the jumper and lifted lead book.
Plant housekeeping controls were monitored, including control and storage of flammable material and other potential safety hazards.
The-inspectors also examined the condition of various fire protection,- meteorological, and seismic monitoring systems. Control
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l room and shif t manning were compared to regulatory requirements and e
l portions of shift turnovers were obsernd.
Control room access was properly controlled and a professional atmosphere maintained.
In addition to 179 hours0.00207 days <br />0.0497 hours <br />2.959656e-4 weeks <br />6.81095e-5 months <br /> of inspection during normal utility working
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hours, the review of plant operations was routinely conducted duri.ng
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portions of backshifts (evening shif ts) and deep backshif ts (weekend and midnight shifts).
Extended coverage was provided for 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> during backshifts and 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> during deep backshif ts. Operators
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were alert and displayed no signs of inattention to duty or fatigue.
2.2 Engineered Safety Features System Walkdown In addition to routine observations made during regular plant tours, the inspectors conducted walkdowns of the accessible portions of selected safety-related systems.
The inspectors verified system operability through reviews of valve lineups, control room system prints, equipment conditions, instrument calibrations, surveillance test frequencies and results, and control room indications. During this inspection period, walkdowns of the following systems were performed:
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Residual Heat Removal System
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No unacceptable conditions were identified.
3.
Radiological Controls 3.1 Routine Radiological Controls Verification During routine tours of the accessible plant areas, the inspectors
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observed the implementation of selected portions of the licensee's radiological controls program. The utilization and compliance with radiation work permits (RWPs) were reviewed to ensure detailed descriptions of radiological conditions were provided and that personnel adhered to RWP requirements. The inspectors observed controls of access to various radiologically controlled areas and use of personnel monitors and frisking methods upon exit from these areas. Posting and control of radiation areas, contaminated areas and hot spots, and labelling and control of containers holding radioactive materials were verified to be in accordance with licensee procedures.
During this inspection period, radiological controls for the following activities were observed.
t fuel inspection and reconstitution
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core support barrel thermal shield removal
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Health physics technician control and monitoring of these activities were determined to be adequate,
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3.2 Radiological Controls During Removal of the Thermal Shield The radiological control aspects associated with cutting the thermal
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shield from the reactor core support barrel were reviewed on a daily basis by the resident inspectors and on January 24, 1990 by the Chief,
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Facilities Radiation Protection Section, of the NRC Region I office.
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The thermal shield is being cut into three segments by means of metal disintegrating machining (MDM). All cutting is performed under water in the reactor cavity. The operational aspects of the process ate covered by vendor procedure VP 523, MP 2.7.1 CYW-19 Thermal Shield
Removal Phase 2 CYW, and the radiological controls aspects are covered under normal station procedures. The cutting tool consists of three carbon-tipped electrodes which cut by vibrating against the thermal shield while a current is passed through them.
The cutting process results in the generation of fines of about five microns or less and the evolution of small amounts of gas, principally hydrogen gas.
A box surrounds the electrode tips; and, a water suction is main-tained to collect the fines as they are created. The suction hose is routed to a charcoal filter system, and then through a paper-type filter system. All these systems are under water. The water flow entraps some of the gas created in the cutting process.
The remaining gas bubbles to the surface of the pool above the cutting location.
Since the workers in the area do not wear respirators, the principle
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radiological concern during the process is the potential for radioactive particulate material being brought to the water surface entrapped in the rising gas bubbles. Once airborne, the workers in the area could then inhale the radioactive particulate matter.
The licensee has made measurements of the radionuclide concentration as the gas bubbles penetrate the water surface. The concentrations that were measured were negligible (E-10 microcuries/ml) and indicate that no special radiological controls precautions need be taken for l
the workers in the area.
Surveys of the outside of the thermal shield indicate generally uniform dose rates (about 1000 R/hr with a
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maximum of 8000 R/hr), The uniformity of the survey results suggest no significant radiological changes would be anticipated throughout the cutting operation.
l One thing that could change the situation is the loss of water j
flow during the cutting operation procedure.
VP 523 states in
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Sections 9.4.5 and 9.4.9 that the operation is to be terminated if L
suction is lost. This procedural requirement should ensure that appropriate radiological controls are maintained during an equipment malfunction.
l Based upon the information provided the inspector, adequate radiological controls were in place for aspects of the thermal shield removal operation that involved the MDM cutting process.
4.
Maintenance and Surveillance
4.1 Maintenance Observation The inspectors observed various maintenance and problem invest',gation activities for compliance with procedures, plant technical specifica-tions, and applicable codes and standards.
The inspectors also verified the appropriate quality services department (QSD)
involvement,- safety tags, equipment alignment and use of jumpers,
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radiological and fire prevention controls, personnel qualifications,
post-maintenance testing, and reportability.
Portions of activities that were reviewed included:
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fuel inspection and reconstitution
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core support barrel thermal shield removal
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Emergency Diesel Generator preventive maintenance
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"A" Auxiliary Feedwater pump reassembly
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4.1.1 Reactor Vessel Thermal Shield Removal General:
The reactor vessel thermal shield was removed from the core support barrel during this inspection period. This work was performed because of significant damage having occurred to the
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thermal shield support structure. This problem was previously discussed in NRC Inspection Reports 50-213/89-16, 89-20, and 89-24, Sections 4.1.1.
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Process Requirements:
The removal was made as a modification which was performed in accordance with plant design change record (PDCR) No. 987, Connecticut Yankee Thermal Shield Removal Program, Revision 0.
As the final engineering analysis could not be finished prior to beginning the work, a series of early construction approv-als were authorized for each phase of the evolution.
Each was
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supported by engineering analysis which addressed the safety of the work process and its effect on plant structures and components.
Those construction approvals that were reviewed i
to date, were identified as Revisions 0 through 5, dated
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December 11, 1989, January 4,11, and 19,1990, and February
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6, and 16, 1990, respectively. Although the final analysis of
reactor operation without the reactor vessel thermal shield has not been completed, this process allows the work to proceed in an evaluated manner.
However, any risk for actions i.
taken prior to completion of the analysis is accepted by the licensee.
The inspectors, who followed the discussions of the analysis
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within the Plant Operations Review Committee (PORC) meetings, observed a high degree of attention applied to the individual process items within work packages.
Detailed levels of
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planning, engineering analysis, scheduling and coordination were apparent within strong management systems. The system of early construction approval appears to be effective.
These construction approvals are supported by the following limited 10 CFR Part 50.59 safety evaluations:
Phase I Activities, review dated December 11, 1989 Phase II Activities:
Thermal Shield Hanger and Core Barrel Stabilizer Base Plate Installation, review dated January 3, 1990 Stabilizer Wire Rope Design, review dated January 3,1990
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Evaluation of Equipment and Process Selection for Thermal Shielci Removal, not dated.
Technical and Safety Evaluation fc. Thermal Shield Removal,
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dated January 17, 1990 (.
Civil Structural Safety Evaluations of Reactor Cavity Floor i
Slab (EL 22'-0") and Reactor Containment Charging Floor Slab (EL 48'-6") and Protection of Reactor Cavity Seals,
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dated January 18, 1990 Civil Structural Safety Evaluation of Floor Anchorage Assemblies, dated January 18, 1990 Nuclear Materials and Chemistry Evaluation, not dated.
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Evaluation of Technical Issues Resulting from Thermal Shield Removal Prior to Initiation of Primary Cutting, dated January 12, 1990, with:
Westinghouse Status Report of Technical Issues (MED-RPV-2629), dated January 11, 1990 Evaluation of Neutron Irradiation Effects due'to Thermal Shield Removal (PSE-CE-90-013), Dated January 5, 1990 l
Estimated Impact of Thermal Shield Removal on Safety Analysis (NE-90-SAB-006), dated January 8,1990 Debris and Cavity Water Control, review dated January 12, 1990, with report from Particle Data Laboratories, Elmhurst, IL, dated November 16, 1989, Westinghouse Evaluation of Debris (NSD-SE-RE-005-90), dated January 5, 1990, and Fuel assembly and Control Rod Guide Tube Debris Evaluation (NE-90-R-017), dated January 8, 1990 Limited 10CFR Part 50.59 Review, The Use of Lif ting Devic-es, dated February 2, 1990, with civil Structural Technical Review of 120 Segment Handling Tool, dated February 2, 1990 The above listed evaluations provided the basis for the licensee to risk release of the construction work of removal of the reactor vessel thermal shield.
They were used by the inspectors in conjunction with the work procedures to provide part of the basis for their reviews.
The first submittal of information to the NRC was made by letter dated February 16, 1990. That document contains a request to amend the facility operating license Technical Specifications which concerns the remaining reactor vessel material surveillance capsules and duration for which the reactor vessel heatup and cooldown curves remain valid. Also I
presented by the document is a description of the project and a description of the debris control and cleanup program, along with a technical justification for the reactor vessel irradia-tion surveillance program. Additional submittals are expected to address system thermal-hydraulic considerations.
Procedural control of this modification was provided by vendor i
procedures VP-520, MP 1.7.1 CYW-18 Auxiliary Lower Internals Storage Stand Installation, dated January 11, 1990, and VP-523, MP 2.7.1 CYW-19 Thermal Shield Removal Phase 2 CYW, dated January 18, 1990.
These procedures incorporated the results of the above referenced engineering evaluations.
Each procedure was reviewed and approved by the PORC.
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I Processes Completed During Inspection Period:
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The inspectors verified that the procedures were sufficiently l
detailed to direct the complex evolutions. A brief summary of the evolutions completed during this inspection period is as follows:
Plates were installed on the reactor containment charging floor slab to serve as anchors for the core barrel stabi-
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lizer cables.
Protective plates were installed over the refueling cavity to reactor vessel pool seals.
Mock-up training was performed for remote installation of the MDM equipment on a full-size, simulated 120' section of
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thermal shield.
The refueling bridge was temporarily relocated beyond the west end of the reactor cavity and turned 180'.
A debris barrier was installed on the auxiliary lowe-internals storage stand (ALISS) while it was on the con-
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tainment charging floor.
The core support barrel was removed from the reactor vessel and suspended from the containment polar crane.
In an intricate lift, while the core barrel remained suspended from the polar crane main hook, the ALISS was placed on to the reactor vessel flange using the auxiliary hook, and the debris cover was installed to cover over the top of the reactor vessel.
The cover, in the form of an inverted top hat, extended into the vessel.
The core barrel was set onto the ALISS.
A second debris barrier was installed within the top of the core support barrel.
Stabilizer cables were installed between the upper flange of the core support barrel and the base plates on the containment charging floor.
Temporary bolts were removed from the thermal shield at the support blocks and were installed through the blocks to the core rupport barrel.
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use of the MDM equipment,
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Hydrogen peroxide was added to the reactor cavity water to a concentration of 10 ppm, to promote oxidation of soluble byproducts of the MDM process.
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L The MDM mast was installed onto the thermal shield and three vertical cuts made such that the shield was divided
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into three 120' pieces.
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Three lifting devices were delivered to the site, one for I'
each of the 120* thermal shield sections.
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Lift rig mock-up training was performed.
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rigging a full-size section of thermal shield from the core support barrel to the floor of the reactor refueling cavity. The thermal shield was then moved to the horizontal position and rigged onto the stand for secondary cutting. The evolution was performed several times to allow training of the rigging crews and to allow support personnel including radiation protection technicians to observe and comment. Modifications were made to the secondary cutting stand as a result of the mock up training
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evolutions.
The three segments of the thermal shield were individually lifted from the core support barrel and placed into temporary storage locations on the reactor refueling cavity floor.
The remote inspections for debris were made in reactor coolant system (RCS) piping by closed circuit television (CCTV).
At the end of the inspection period, the status of the reactor core support barrel thermal shield assembly was as follows:
The reactor core barrel was supported on top of the reactor vessel flange through the ALISS.
The stabilization restraint cables were put in place.
The three reactor vessel thermal shield segments were located on the floor of the reactor refueling cavity.
Stabilization was provided by their individual lifting rigs.
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The debris barriers and the reactor refueling cavity floor were being cleaned by underwat(r vacuum.
The cavity water l
also was being recirculated through 0.3 micron filter units.
The MDM cutting equipment was removed from the containment.
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A tank to contain the secondary cutting stand within the reactor cavity was being assembled.
Preliminary inspections by CCTV were made for damage of the outside surface of the reactor core support barrel.
Preliminary inspections by CCTV were made for debris in the reactor coolant loops up to the reactor coolant loop stop valves.- (These valves have remained shut during the inspection period).
Underwater surveys were made of the thermal shield segments to characterize its radioactive waste curie content.
Debris Control:
Repair activities during the 1987 outage apparently created the source of metallic debris which caused damage to the-stainless steel fuel clad.
Debris control at that time consisted of the placement of barriers and collection devices in the work area, along with cleaning and inspections _
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Because of the potential for additional fuel or reactor component damage, a high degree of attention currently is beir.g applied to debris control.
Analysis has characterized the size of debris generated by the two principal machining techniques for the removal and disposal of the thermal shield.
Each process generates a unique type of debris. The MDM primary cutting debris is a i
very fine powder, with a size range of 0.38 to 579.0 microns
and a median size of 0.578 microns. The plasma are secondary cutting debris is of a granular consistency; some slag and heavier bead size is produced.
I'.s size ranges from 38 to 2360 microns.
Using this information, plans were formulated to control migration of this material into the reactor coolant system.
Additionally, its effects on system components will be evaluated.
The licensee has established an overall project objective to control debris during the thermal shield cutting and removal
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phase to meet or better the particle size limits specified in ANSI 45.2.1 - 1973, for cleanliness level B.
This is defined i
as "...no larger than 1/32 inch in any dimension (which is j
approximately 800 microns) with the exception of fine hairline i
siivers of less than 1/32 inch thickness and up to 1/16 inch j
long."
Process controls have been in place to draw the MDM cutting debris from an enclosure surrounding the cutting heads through
a filtration system.
In addition, recirculating filtration
systems have been maintaining reactor refueling cavity water filtered to 0.3 micron.
Because of the heavier type of debris
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created by the high speed plasma are torch, that work will be isolated within a steel enclosure.
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Prior to'the MDM cutting evolution, hydrogen peroxide was added to the reactor cavity water to promote oxidation of the soluble ferrous hydroxide into an insoluble ferric hydroxide.
The ferrous hydroxide is a product of the free iron ions generated at the cutting electrode and is not filterable.
Hydrogen peroxide was added to maintain a concentration of 8 to 12 ppm.
In an attempt to isolate sections of the reactor refueling cavity, debris barriers wera installed to cover the reactor
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vessel at the flange. A polyester felt material was placed under and inside the core support barrel to control the j
migration of debris into the reactor vessel.
The reactor
l coolant system loop isolation valves have remained shut during the entire removal process. Barriers made of herculite were used to cover the reactor upper internals package (stored at I
the west end of the refueling cavity) and to isolate the fuel
transfer canal.
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Following the primary cutting and removal of the thermal
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shield sections, extensive cleanup was performed using an underwater vacuum with a particulate filter.
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l The licensee is presently formulating the detailed plans for a
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series of system cleanup flushes, cleaning and inspection.
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This is planned first for the RCS support systems, then for the RCS itself. These will eventually include the reactor vessel and its internal components. Of particular concern, is the possibility that debris, similar to that which caused fuel damage, may' remain within the RCS, its support systems and
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piping such as the pressurizer, or in outlying support system py
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The refueling water storage tank has been drained and will be inspected and cleaned to provide a source of clean water for l
flushing. This entire program is expected to be quite exten-sive.
Inspection Findings:
The inspectors reviewed the licensee's actions throughout the inspection period. Observations were made frequently at the work locations in the reactor containment and during the mock-up training. The inspectors attended most of the PORC meetings in which reviews were made of the engineering analysis supporting the work and the work procedures.
Discussions were held frequently with plant management, supervisors involved with the thermal shield project,
engineering and licensing personnel, as well as those
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responsible for performing the work activities.
I Of particular interest to the inspectors were radiation protection measures and monitoring of the work activities, along with any activity which may have affected worker safety.
This includes safe rigging practices of heavy loads, worker interface with radiation protection and station procedure requirements and monitoring the effectiveness of management oversight. The point of concern was that the licensee was dealing with large pieces of radioactive material. The
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thermal shield was a 4.2" thick, 157" high, 137.5" inside
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diameter stainless steel cylinder, weighing about 85,000 pounds.
Each of the three 120* pieces weighs about 28,000 pounds. Detailed survey of the inside surface of the thermal shield indicated radiation levels up to 45,000 Rem per hour.
The inspectors observed a high regard for procedural require-ments of both the work procedures and radiation protection L
procedures. Work practices were professional with appropriate levels of supervision.
Proper interface was observed between the workers and the health physics technicians. The work progress was monitored to conform with these requirements, i
The licensee maintained work practices designed to control L
potential " hot particles" from within the reactor cavity.
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Good controls were in place over the work activities.
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example, after start of MDM primary cutting of the thermal l
shield, minor problems were caused by the gases created in the inetal disintegration process.
These gases churned the charcoal within the first of two devices filtering the effluent of the MDM cutting equipment. The charcoal fines from this filter led to rapid clogging of the secondary particulate filters. Work was stopped when clogged particu-
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late filters resulted in the release of a minor amount of gas from the MDM equipment. This gas carried with it some of the radioactive material disintegrated from the thermal shield (This is discussed in section 3.2 of this report). Water clarity also became a concern and work was stopped until the filtration system was modified. By simply enlarging the vent
of the charcoal filter vessel, sufficient gas separation occurred to eliminate the agitation of the charcoal within the
. filter. This stopped the production of fines and provided for effective water filtration.
The water within the reactor cavity was recirculated through progressively finer filters.
Final clarity was accomplished with 0.3 micron elements.
Additional time was spent in the cleanup of the reactor refueling cavity floor. Again, these activities reflect a high degree of concern for the control of radioactive material and worker-safety.
Mock-up training was performed using a specially fabricated piece of steel made to simulate one of the 120' sections of thermal shield. An area was set up on the turbine operating floor in which the reactor cavity volume was modeled using staging.
Personnel were trained in remote rigging of the MDM machine onto the thermal shield.
It was also used in extensive rigging crew training for moving the thermal shield sections. These training evolutions led to making small changes to the support stand for secondary thermal shield cutting. Although a minor interference would have occurred with the original design of the cutting stand, complex actions would have been necessary to correct them once in the reactor refueling cavity.
Finding the problem while on the j
turbine deck allowed for a simple corrective action.
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The inspectors noted that the mock-up training evolutions were attended by a significant cross section of personnel responsible for performing, supervising, planning and managing the work. They included, but were not limited to quality j
assurance, health physics, maintenance, construction and contract support personnel.
There were no unacceptable conditions identified.
4.1.2 Inspection and Reconstitution of Reactor Fuel The licensee completed the inspection and reconstitution of reactor fuel on February 7, 1990. This work was performed to l.
reclaim fuel assemblies which had suffered damage due to l
metallic debris within the RCS.
The fuel damage was l
previously discussed in NRC Inspection Reports 50-213/89-16,
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89-20, and 89-24, Sections 4.1.2.
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Ninety-two assemblies were reconstituted for reuse in operat-ing cycle 16. Additionally, six assemblies from which donor rods were used to reconstitute these assemblies, were
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themselves rebuilt by consolidating acceptable fuel rods.
These six assemblies will be used in future operating cycles.
The licensee supplemented the procedures in place for fuel
rod eddy current test inspection, rod handling, and assembly
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L reconstitution with an additional process for inspection of the fuel rod surface by ultrasonic examination.
Vendor procedure VP-521, Babcock & Wilcox Procedure No. F0-038 Operation Procedure for the Ultrasonic Inspection of Fuel Rods at Connecticut Yankee, Revision 0, dated January 16, 1990, directed the process for examination of the fuel rod clad surface for defects. This technique was used in conjunction with eddy current testing to better identify and characterize the size of surface defects, preliminary results from the inspection process revealed that within the reload assemblies 311 leaking fuel rods were
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identified, and 179 fuel rods were found with defects greater than twenty percent of the fuel. clad wall thickness and rejected.
In all, for various reasons, 527 fuel rods were rejected in the inspection and reconstitution process.
Examination and handling of approximately 3479 fuel rod was
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performed to accomplish this task.
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The inspectors observed inspection and reconstitution activities throughout the inspection period.
Personnel adhered to procedures and radiation protection requirements; and controls to provide material accountability were in place.
There were no unacceptable conditions identified.
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4.2 Surveillance Observation
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The inspectors witnessed selected surveillance tests to determine
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whether properly approved procedures were in use, technical specifi-cation frequency and action statement requirements were satisfied, necessary equipment tagging was performed, test instrumentation was in calibration and properly used, testing was performed by qualified personnel, and test results satisfied acceptance criteria or were properly dispositioned.
Portions of the following activities were reviewed:
ST 11.7-14, Appendix "R" Service Water Performance Test
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performed on January 30, 1990.
SUR 5.1-17B, Emergency Diesel Generator EG-2B Manual Starting
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and Loading Test
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No unacceptable conditions were identified.
4.2.1 Appendix R Service Water Performance Test During this inspection period, the licensee performed the first of four special tests which verify integration of the new switchgear building.
ST 11.7-14, Appendix R Service Water Performance Test, demonstrates the ability of service water (SW) to support the emergency diesel generators (EDGs) and the containment air recirculation and residual heat removal systems during the most restrictive Appendix R configuration.
The inspectors observed portions of the test performance on January 30, 1990.
The test results for SW flow to EDG-2B heat exchangers were-unsati sf actory. Cooling water flow was measured at 355 gpm rather then the required 400 gpm.
Evaluations of EDG-28 operability and reportability of this condition were initiated and in progress at the close of the inspection period.
The licensee removed a segment of piping at the EDG-2B heat exchanger SW outlet to inspect the condition of the piping. A corrosion layer has formed on the inside surface of the pipe, effectively reducing the pipe inside diameter to approximately 2.75 inches rather than the original inside diameter of 4.03 inches. At the end of this inspection period, the licensee was developing a method for cleaning of this and other affect-ed piping.
The inspectors observed portions of the test and inspec-
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tion of the EDG-2B heat exchanger outlet piping. This matter will be reviewed during routine inspections of engineering and maintenance activities and as part of the ongoing resolution of outstanding issues concerning the SW system.
5.
Security During routine inspection tours, the inspectors observed implementation of portions of the security plan. Areas observed included access point search equipment operation, condition of physical barriers, site access control, security force staffing, and response to system alarms and degraded conditions. These areas of program implementation were deter-mined to be adequate.
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IS 6.
Engineering and Technical Support l
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6.1 Potential Service Water Filter Clogging Following a LOCA Additional filtration of the service water (SW) system is provided by two self-cleaning filter units located on the upper level of the auxiliary building.
Service water is then distributed to the containment air recirculation fan coolers, spent fuel pool cooling
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heat exchangers, and several other auxiliary building components.
The filters are automatically cleaned by a self-contained backwash system which cycles filtered water back through the filter mechanism.
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During an engineering evaluation for procurement of SW filter re-placement parts, inconsistencies were noted in the qualification of some backwash component parts.
Further investigation identified the following discrepancies concerning these filters:
The Material Equipment and Parts List designated the filters
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as QA, however, components of the filter backwash are non-QA.
The backwash air-operated valve fails to the closed position
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on a loss of air, therefore backwash capabilities are lost and the probability of clogging is increased.
Backwash alone is not sufficient to prevent clogging of the
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filter during the worst case river water silt conditions.
The SW design flow analysis assumes backwash is available.
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Operator action is necessary to change filter trains when one filter clogs. However, in the event of a Loss of Coolant Accident, the area becomes uninhabitable due to high radiation fields.
In this case, clogging of the filters renders the containment air recircula-tion fan coolers incapable of reducing post-accident containment
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pressure.
The potential for clogging of these filters and inhibiting contain-ment pressure reduction capabilities was determined to be reportable L
in accordance with 10 CFR 50.72. The appropriate notifications were l-made on February 2, 1990.
The licensee is currently developing a corrective action plan involv-ing a design change to the SW system.
This change will be reviewed during routine inspection of design change activities and the resolu-tion of outstanding issues concerning the SW system.
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7.
Safety Assessment and Quality Verification 7.1 Plant Operations Review Committee The inspectors attended several plant operations review committee (PORC) meetings. Technical Specification 6.5 requirements for required member attendance were verified.
The meeting agendas
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included procedural changes, proposed changes to the Technical Specifications, plant design change records, and minutes from previous meetings.
PORC meetings were characterized by frank discussions and questioning of the proposed changes.
In particular, consideration was given to assure clarity and consistency among procedures.
Items for which adequate review time was not available were postponed to allow committee members time for further review
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and comment.
Dissenting opinions were encouraged and resolved to the satisfaction of the committee prior to approval.
The inspectors observed that PORC adequately monitors and evaluates plant performance and conducts a thorough self-assessment of plant activities and programs.
7.2 Review of Written Reports Periodic and special reports, licensee event reports (LERs), and safeguards event reports (SERs) were reviewed for clarity, validity, accuracy of the root cause and safety significance description, and adequacy of corrective action. The inspectors determined whether further information was required. The inspectors also verified that the reporting requirements of 10 CFR 50.73, 10 CFR 73.71, Station Administrative and Operating, and Security Procedures, and Technical Specification 6.9 had been met. The following reports were reviewed:
SER 90-S01 Safeguards Event Report LER 89-22 Pressurizer Safety Valves Setpoints Found High During Testing LER 89-23 Inoperable Fire Barrier Between Switchgear Room and Cable Spreading Area LER 89-24 Design Deficiency Identtfied in Safety Injection Block Switch Haddam Neck Plant New Switchgear Building Bimonthly Progress Report No. 20, dated January 31, 1990 l
Haddam Neck Plant Monthly Operating Report 89-12, covering the period December 1, 1989 to December 31, 1989
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l Haddam Neck Plant Monthly Operating keport 90-01, covering the period
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January 1, 1990 to January 31, 1990 Special Report concerning Radioactive Effluent Pursuant Actions, dated February 14, 1990
Haddam Neck Plant Report of the Results from the 1989 Steam Generator Tube Inspection, dated January 19, 1990. This report is required by Operating License Technical Specification 4.10.1.
Revision No. I to the Haddam Neck Plant Compliance Review for 10 CFR 50, Appendix R Safe Shutdown Design Basis Fire, dated January 18,
1990.
The principal reason for this revision was to incorporate the results or recently completed performance analysis of service water, component cooling water and ventilation systems into the analysis, t
No unacceptable conditions were identified.
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8.
Analysis of Moderator Dilution Events (TI 2515/94)
s In response to the identification of an unanalyzed moderator dilution event, the NRC issued a letter, dated September 26, 1977, informing licensees of the event and requesting an evaluation of the potential for this type of event. The Itcensee responded by letters dated January 13, 1978, June 4, 1980, and August 13, 1981.
These submittals described the review of all possible dilution paths to the reactor coolant system.
It was concluded that the boron dilution event analyzed in the Final Safety Analysis Report is the worst case of this type of event. Therefore, no corrective measures were necessary.
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NRC reviewed these submittals, found the analysis to be satisfactory and agreed with the licensee's conclusions. A Safety Evaluation Report was issued on November 16, 1981, which details the NRC acceptance and closure of this item.
9.
Procurement of Emergency Diesel Generator Fuel Oil
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- NRC Multi-Plant Action Item A-15 concerned the procurement practices of emergency diesel generator fuel oil in regard to 10 CfR Part 50, Appendix L
B requirements.
l The licensee considers that fuel oil is safety related. However, it is l
procured as a commercial grade consumable item and dedicated for use in a safety-related system in accordance with administrative control procedure ACP 1.2-4.2, Commercial Grade Procurement, Upgrade and Dedication, Revi-sion 3, dated November 21, 1989. This process involves sampling delivered
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batches of oil in accordance with chemical control procedure CHM 7.2-1, Diesel Oil (75/25 #2 Fuel Oil Blend) Sampling and Analysis, Revision 5,
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dated May 26, 1988, prior to filling the fuel oil storage tank.
Although the present operating license has no stated specifications for i
fuel oil, the new Technical Specifications, which is expected to be issued in the near future, contain the following requirements in section 4.8.1.1.2,d and e'
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By obtaining a sample of new fuel oil upon delivery to the site prior
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to addition to storage tanks:
1) Verifying in accordance with the tests specified in ASTM-0975-81; a) An API gravity of greater than or equal to 30 degrees but less
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than 40 degrees; b) A water and sediment volume percentage of less than or equal to 0.05% (500mg/L);
c) A Saybolt viscosity, SUS at 100*F of greater than or equal to 32.6, but less than or equal to 40.1; and d) A flash point equal to or greater than 125'F.
2) Verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with ASTM-D975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D2622-82, e.
At least once every 31 days by obtaining a sample of fuel oil from the underground fuel oil storage tanks in accordance with ASTM-D-270-75, and verifying that:
1) API gravity is greater than or equal to 30 degrees but less than or equal to 40 degrees; 2) Water and sediment volume percentage is less than or equal to 0.05% (500mg/L); and 3) Kinematic viscosity at 40*C is greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes (alterna-tively, Saybolt viscosity, SUS at 100 F of greater than or equal l
to 32.6, but less than or equal to 40.1).
Procurement as commercial grade with a dedicated upgrade procedure for use in safety-related equipment is considered to be a conservative and
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realistic approach when compared to attempting to obtain source documentation on a product which is only handled in bulk processes.
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'There were no unacceptable conditions identified.
10.
Exit Interview
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During this. inspection,. periodic meetings were held with station manage-ment to discuss-inspection observations and findings. At the close of the inspection period, an exit meeting was held to summarize the conclusions of;the. inspection.
No written material was given to the licensee and no proprietary information related to this inspection was identified.
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