IR 05000213/1990008
| ML20055G128 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 07/11/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20055G113 | List: |
| References | |
| 50-213-90-08, 50-213-90-8, NUDOCS 9007200157 | |
| Download: ML20055G128 (147) | |
Text
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? ,, c p h U.S. NUCLEAR REGULATORY. COMMISSION f',
REGION I
~ i , Report No.
50-213/90-08: ' License No.
DPR-61 Licensee: Connecticut Yankee Atomic Power' Company P. O. Box 270
Hartford, CT 06141-0270 a p Facility: .Haddam Neck Plant [:; . Location:. Haddam' Neck,' Connecticut ( [H Inspection . i Dates: May 9. 1990, to June 20, 1990 .t Reporting Inspector: John T. Shedlosky, Senior Resident Inspector In spect' ors : Andra A,Asars,. Resident Inspector Herbert J. Kaplan, Senior Reactor Engineer, Materials and
Processes Section-(MPS), Division of-Reactor Safety (DRS) J John R. O'Neal, NDE Technician, MPS, DRS '[ Frederick P Paulitz, Electrical Engineer, Instrumentation and ! Control 5 stems Branch, Division of Systems Technology,
Office of Nuclear Reactor Regulation
John T. Shedlosky, Senior Resident Inspector ' ' James S. Stewart, Project, Engineer, Reactor Projects Section 4A, ' Division.of Reactor Projects (DRP) Approved by: 8 8 / m [ u /p 7/n/4o A Donald R. Haverkamp,. Chi (4A f ' Date Reactor Projects Section ( Division of Reactor Projects Inspection Summary: Inspection on May 9, 1990 - June 20, 1990 I (Inspection Report No. 50-213,90-08) heas' Inspected: Routine safety inspection by the resident inspectors and . . specialist inspecto.rs from Region I and NRR. Areas reviewed included plant
^ operations, reactor fuel reload activities, radiological controls, investiga-
- tion of' an' engineered safety features system activation, loss of power testing, W
implementation of. Revised Technical Specifications, cleaning and inspection of the. reactor coolant system, inspection of pressurizer clad cracks, control room environment and habitability issues, electrical system modificati )s, plant operations review committee meetings, and actions taken in response to
, previous inspection findings.
Results: See Executive Summary 9007200157 900712 PDR ADOCK 05000213 g PDC <
, Executive Summary . Plant Operations: This was the seventh routine resident inspection during the q b 1989/1990 refueling outage.' Reactor core reload was started on June 14 and completed on' June 16.
Reactor reassembly was in progress at the conclusion of the inspection period. 'There were no noteworthy findings in this area.
Radiological Controls: Strong radiological controls were observed in place during the cleaning and inspection of the reactor vessel, the reactor ca/ity and the pressurizer.
. Maintenance and Surveillance: The final cleaning and inspection of the reactor ind the pressufizer vessels for foreign material was completed.
Cracks of the stainless steel clad on the inside surface of the pressurizer were observed during these inspections.
Service water piping for the emergency diesel generators and the control room air' conditioning system was replaced; the piping serving the containment air recirculation (CAR) fan coolers was mechanically cleaned.
Modifications to provide full bypass flow around the C/.R f an cooler supply filters were in pro-- ] gress.
j Emergency Preparedness: An Emergency Preparedness' Exercise was conducted on May 19.
Inspection findings are contained within report 50-213/90-09.
q Security and Safeguards: Modifications were made to the protected area bound-- a ry.
These were provided to include the station office building within the f protected area and to improve the vehicle access point.
Engineering and Technical Suppo~rt: The licensee appears to have failed to meet I a commitment to install radiation and smoke detection devices which isolate the control room ventilation outside air intake,.This item is unresolved (213/ 90-08-01) pending resolution of the status of the commitment.
, Licensee' inspections identified the need for replacement or' cleaning of service > water piping involving the emergency diesel generators, the containment air ! recirculation fan coolers and the control room air conditioning equipment, y That work was completed.
j i There appears to be a weakness in engineering support for the selection of the trip points of high speed fault current protective devices recently installed i in motor' operated valve (MOV) motor control centers (MCC).
This item is un-
resolvedL(213/90-08-02) pending NRC review of the calculation for those trip .; points.. J l 'The possible flaws in the pressurizer shell have been resolved in accordance j-with the Section XI, Article IWB-3000 of the ASME Boiler and Pressure Vessel
Code.
Code authority, NRC,. approval is required for the acceptance by evalua- [ tion of one of those defects.
' f Safety Assessment and Quality Verification: The licensee has developed detailed listings of Butage work activities and project assignments to identify and closely track issues requiring resolution prior to an operational mode change.
Timely review was made of equipment deficiencies; prompt and detailed review of procedure and design change documents was conducted by the plant operations L review committee.
! < : p - E e i TABLE OF CONTENTS Page 1.
Summary of Facility Activities I ...... ........... 2.= Plant Operations'(71707, 71710, and 93702)*............
.. =2.1 Operational Safety Verification 1.
............... 2.2 Follow-up of Events Occurring During the Inspection Period......,....................
2.2.1 Unplanned Engineered: Safety Features System ' Actuation..........
............ . 3.
Radiological Control s (71707)'......... ,..., ....
1l 4.
Maintenance and Surveillance (61726, 62703, and 71707)
.l ...... , ' 4.1 Maintenance Observation.
................... 4. '1.1 - Investigation of ESF Actuation - Containment
Isolation..........
j ............ 4.1.2 Refueling Operations
........... ..... 4.1.3. Service Water Pipe Cleanliness
............ 4,2 Surveillance Observation..
. ................. 4.2.1 Loss of Power. Testing................. 7 --
4.2.2 Implementation of Revised Technical ! Specifications
.................., 4.2.3 Cleaning and Inspection of Reactor , Coolant System
l , .....,,............ S.
. Emergency Preparedness (82301)
.....,............ ' 6.
. Security (71707)
g , .............-,........... 6.1 Physical' Security'0bservation... 10- ............... J
6.2 Fitness for Duty Concern. (RI-90-A-0027).... ,......
, 7.
Engineering and: Technical Support (37700, 37828, 71707, 92701, ! 37701, 73755, 57050, 57080, and 57090).............. 'll - J '7.1 Pressuri zer Clad. Cracking.................. 11.
7.1.1 Summary..
.a ................. .... 7.1.2: Vi sual Exami nations..................
l 7.1.3 Pressure Vesse1~ Design Data..
............ ' 7.1.4 Radiographic Examination
i .......... .... 7.1.5 Ultrasonic Examination
................ 7,1.6 Flaw Evaluation.......
, ............. 7.1.7 Conclusion
...................... T-1 , J I ! b
(=-- i.f ! s I s TABLE OF CONTENTS Page
17.
Engineering and Technical Support (continued)
7.2 Control Room Environment.
' ..........,....... -7.2.1 Control' Room Ventilation System...
......... ' 7.2.2 Control Room Habitability Issues
........... 7.2.3 Service Water Modifications Affecting i . the Control Room 18'
.................. 7.2.4. Operational Experience
. ,............. . 7.2.5 Conclusion - 19 - ...................... - 7.3 Electrical Modifications................
... 7.3.1 Purpose.........
............... 7.3.2 Scope.........................
7.3,3 Findings
1 .......................
7.4 Potential Steam Line Break Near Safety _ 22 ,Related Equipment .............,.....,.
i 7.5 Followup of Previous Inspection Findings
7. 5.1 - (Closed) Unresolved Item 213/87-05-01: i ' Improvements for Post Modification Testing.....
7.5.2 (Closed) Unresolved Item 213/87-08-02: . - Implementation of Preoperational Testing Program.
. 7.5.3 (Closed) Unresolved Item-213/88-06-01: Implementation of a Corporate Level Procedure for Design Change-Turnover and Testing,, .
... 8.
~ Safety Assessment and Quality Verification (40500, 71707, '90712, and 92700),.......................
. -! 8.1 Plant Operations Review Committee
.............. 8.2: Review of Written Reports ............. ,, -.. ~ i 8.3 Tracking of Plant Conditions to Support Operational Mode Changes..............,,..
9.
Exit' Interviews (30703 and 30702)................. 24-9.1 Exit Interviews
....................... 9.2 Outage Activities and Restart Plan Meeting..
4 ........ 9.3 Systematic Assessment of Licensee Performance ' Management Meeting.........
............
- The NRC Inspection Manual inspection procedure or temporary instruction that was used as inspection guidance is listed for each applicable report section.
T-2 . -. -- - -. - a
gL DETAILS 1.
Summary of Facility Activities The' fifteenth 1 refueling outage continued-during this inspection period.
Major work activities included completion of reactor coolant system.and - refueling cavity cleaning, loss of power testing, implementation of
- j Revised Technical Specifications, and reactor reassembly.
Reactor core L reload was initiated and Operational Mode 6'was entered on June 14.
2.
Plant Operations \\
-2.1 Ojerational Safety Verification { 1The 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: security access point ] control room -- -- primary auxiliary building' protected area fence -- -- radiological control point intake structure -- -- electrical switchgear rooms diesel generator rooms -- -- auxiliary feedwater pump room turbine building -- -- Control room instruments and plant computer indications were observed for correlation between channels and for conformance with technical
specification (TS) requirements.
Operability of engineered safety i; features, other safety related systems and on-site and off-site newer j sources'were verified.
The inspectors observed various alarm condi-a .tions and confirmed that operator response was in accordance with j plant operating procedures.
Routine operations surveillance testing
was also observed.
Compliance with TS and implementation of appro-( 'priate action statements for out-of-service equipment.was inspected, i Plant radiation monitoring system indications and plant stack ! effluent monitors 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 storata of flammable material'and other poten-tial safety hazards.
The in., actors also: examined the condition of ~ various fire protection, meteorological, and seismic monitoring systems.
Control room and shif t manning was compared to regulatory , requirements and portions of shift turnovers were observed.
Control room access was properly controlled and a professional atmosphere ! ~
maintained.
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, In addition to normal utility working-hours, the review of plant
. . operations was routinely conducted during portions of backshifts
y-(evening shif ts) 'and deep backshif ts (weekend and midnight shifts)~. . Inspection coverage was provided for 46 hours-during backshifts and ' -28. hours during deep backshifts. Operators were alert and displayed no signs of inattention to duty or fatigue.
2.2 Follow-up of Events Occurring During Inspection ~ Period During-the inspection-period, the inspectors provided on-site cover-s- " age'and follow-up of unplanned events.
Plant _ conditions, alignment y of safety systems, and licensee actions were reviewed.
The inspec-tors confirmed that the required notifications were made to NRC.
During event follow-up, the inspectors reviewed the corresponding plant information report (PIR) package,' including the event details,- .r root cause analysis, and corrective actions taken to prevent recur-rence. The following events were reviewed: , 2.2.1 Unplanned Engineered Safety Features System Actuation On June 6, at 11:50 p.m., an automatic containment isolation was actuated.
No work was in progress and there was no apparent reason for the event.
At the time of occurrence and because of the extended outage, all fuel was removed from the' reactor and most safety-related equipment was removed from service. Opera-tors verified proper operation of safeguards equipment. All operable equipment functioned as required except for the No. 2- .i containment recirculation fan damper position indication; damper i position was verified locally and found to be correct.
Operators performed the appropriate-steps of ES-1.1, Safety Injection (SI) Termination, to reset the SI and containment isolation signals. They were unable to reset the containment , isolation actuation relays.
Plant incident report, PIR 90-111 , was initiated.
' ' The event was discussed.at the morning management meeting; reporting was made in accordance with 10 CFR Part 50.72.
M A defective switch contact was found to be the cause.
) The' inspector observed licensee actions, there were no unac-ceptable conditions identified.
A discussion of the trouble i shooting and testing activities is within report section 4.1.1.
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. j l c 3.. Radiological Controls < During. routine-inspections of the accessible plant areas,-the' inspectors-observed.the implementation of selected portions of the. licensee's radio- . logical controls program.
The utilization and compliance with-radiation
work permits (RWPs) were reviewed to ensure:that-detailed descriptions of L radiological conditions were provided and that mnel adhered to RWP-requirements.
The inspectors observed controls ss to various radi- _ ologically controlled areas and use of personnei rs and frisking methods upon exit from these areas.
Posting and_co.a ol of radiatio _n =l 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.
reactor coolant system and refueling cavity cleaning -- pressurizer cladding inspection j --- ' reactor core reload and reactor reassembly -- ! Health physics -technician control and monitoring of these activities were 'i determined to be adequate.
4.
-Maintenance and Surveillance a 4.1 Maintenance Observation . . .! .The-inspectors observed various corrective and preventive maintenance
-activities for compliance with procedures,. plant technical specifica- 'I tions, end. applicable codes and. standards.
The inspectors-also veri- , fied the. appropriate quality services department (QSD) involvement,
-use of safety. tags, equipment alignment and'use of-jumpers, radio-j logical and. fire prevention controls, personnel qualifications, j post-maintenance testing, and reportability.
Portions of activities y that were reviewed included: .] I "A" service water pump overhaul, i -- "A" auxiliary feedwater pump repair, -- troubleshooting' investigation into spurious actuation of the ! -- containment isolation system, i
containment air recirculation fan cooler supply and return pipe j -- cleaning per procedures CMP 8.7-1 and VP-556, ' replacement of emergency diesel generator service water piping, -- and
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-k ' - , repair to the spper in'ternals core exit thermocouple wire ' ~ -- y conduit-support.
4.1.1 Investigation of ESF Actuation - Containment' Isolation
Licensee personnel investigating the' emergency safeguards , feature actuation, which occurred on June 6, found that a - ! terminal stud bolt head had failed on a single contact of- ! the latching switch used in the "A" channel of the contain- - ment isolation' actuation. logic.
This allowed the contact-to open and pick up'a seal-in circuit and trip the latch i switches in both containment isolation subsystems.
. The licensee is uncertain as to when the terminal stud bolt ! head failed because it appeared that an insulating contact separator may have applied light closing pressure to.the. -broken contact.
This would have masked the problem until-
i pressure relaxed and the contact opened.
Contacts of all Westinghouse type WL. latching switches were inspected by ti,e licensee and found-satisfactory.
, During troubleshooting, a second deficiency was found. A
loose wire terminal was found on-the-main control board-i switch for the air-operated containment isolation valve,- ' WG-A0V-558.
The terminal is in a circuit which insures that control switches for containment isolation valves are ! in the close position before allowing the reset:of the l ' ' containment isolation latch switch.
Successful reset of.
, the isolation would have been prevented by either the ,; failed' latch switch-contact or the loose wire' terminal.
, , The system was successfully tested after these deficiencies I were corrected.
It was the licensee's conclusion that.the-a latch switch was defective for the life of the plant; but, J unused until the installation of modifications during _the -} current outage.
In that case, the root cause would be L improper design' document control during initial plant co r _ struction.
, The investigation into this event caused the. licensee to j reevaluate a feature of the containment isolation actuation ' logic design in which each of.the two sub-channel latch l switches would provide a trip to the. opposite channel ' switch. -In this design configuration, the operator was-required to simultaneously reset both switches.
This operation was considered both unusual and awkward because p as one switch reached the reset position before the other, I the trip coils were-reenergized while the switches were
being held closed.
With this type of switch, there was a i concern that premature failure of the latch trip coil may
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occur if the switch were to be manually held:in'the reset' position for a sustained period while the trip coil is i energized.
The licensee was considering a-change to this recently modi- ' fied design to eliminate the complementary tripping between- - +i the actuation switches since the system logic provided'a. trip' signal:to both switches on high' pressure in any two of the n , four containment pressure channels..The actuation switch a complementary tripping was added with recent modifications ~to !
the reactor protection system.
This~ design change review was in progress at the end of the inspection period.
- L The inspector observed troubleshooting and testing activi-ties.
No deficiencies were identified.
4.1.2 Refueling Operations I On June 14, the licensee initiated reactor core. reload.
The inspectors observed reload activities and verified the U conformance with the Technical Specifications for Opera-i tional Mode 6 which are listed on attachment I to this ' - report.
The licensee's overall administrative controls were contained in administrative control procedure ACP-1.0-10, " Refueling Operations,"-Revision 9, dated June 3, 1990; the surveillance requirements in SUR 5.3-51, " Refuel-ing Operations," Revision 1, dated June 8,1990; the opera-ting procedure' for the containment system in_ N0P 2.13-5, _ t " Establishing Containment Integrity," Revision 10; and, the-m . detailed stepwise refueling and reactor reassembly proce-y
- dure.in vendor procedure'VP-471, " Refueling Procedure Cycle'
XV-XVI, Connecticut Yankee," Original, dated August 31, 1989, which endorsed-the Westinghouse refueling procedure FP-CYW-RIS, dated-August 15, 1989.
The core reload was completed on June 16 and reactor re-assembly started at-that time. At the end of_this inspec-tion period on June 20, the reactor vessel head had been - set in place and the decontamination effort of the reactor refueling cavity commenced.
In addition to the Technical Specifications and procedure , requirements, the twenty procedure changes affecting VP-471 ' were also reviewed.
i , There were no unaccentable conditions identified.
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t , i 4'.1. 3 ' Service. Water Pipe Cleanliness As the result of the engineering analysis of the results of-a special test of service water system performance, the-licensee made. a significant effort to restore portions-of the~ system to a high level.of performance. The, test, ST 11.7-14, " Appendix "R" Service Watcr Performance Test," Revision 1, dated October 20, 1989, revealed that, with the . ' system operating in a ' restrictive configuration, the ser- . vice water cooling flow to the emergency diesel generators; a and to the containment air recirculation (CAR) fan. coolers was less than-the test acceptance criteria, , This test was observed by the inspectors when performed on. ' January 30, 1990; it was discussed in section 4.2.1 of resi- . dent report 50-213/90-02.
The licensee's investigation revealed the buildup of cor- .. rosion products on the inside surface of the pipes.
This.
' - had reduced the effective inside diameter of the piping.
The decision-was made to replace the diesel, generator' pip ing as it was all located within the diesel rooms and to mechanically clean the CAR fan cooler supply and return-piping.
-i l The cleaning evolution was conducted per corrective'mainte-nance procedure CMP-8.7-1, and vendor procedure VP-556.
' following-these evolutions which were completed, the li-censee intends to repeat the applicable portions of.
J ST.11.7-14.
This will occur during the next resident
inspection period.
There were no unacceptable conditions identified.
j l 4.2 Surveillance Observation ' The inspectors witnessed selected surveil-lance tests to determine
whether properly approved procedures were in use; technical specifi-
- cation. frequency and action statement requiraments were satisfied;
[ necessary equipment tagging was performed; ts 'nstrumentation was ! in calibration and properly used; testing was pin formed by. qualified personnel; and test results satisfied acceptance criteria or were
properly dispositioned.
Portions of activities associated with the-following procedures were reviewed-I , ST 11.7-15, " Pre-Fuel Load "B" Switchgear Room Compliance -- Test" ST 11.1-2, " Reactor Coolant System Subsystem Cleanup" -- i u SUR 5.7-148, " Inservice Testing of A, B, C, and D Service Water -- Pumps Surveillance" t , .
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b SUR_5'.2-63, "LTOPS Analog Channel Operational Test"- [ -- _ a ST 11.8-29, "125 Volt D.C. Distribution Panel BX Breaker.
l -- . Replacement (PDCR 995)" - . --- 'SUR 5.7-31, " Boric Acid and RWST Check Valves" ' There were no unacceptable conditions identified, ! if '4.2.1-Loss of Power Testing On May 13, the licensee conducted a special test-to demon-strate compliance with 10 CFR 50, Appendix R, which'was part of the acceptance tests series for the recent elec-i trical distribution modifications. 'The test was-conducted per ST 11.7-15, " Pre-Fuel Load "B" Switchgear Room Compli- - ance Test," Revision 2, dated February 26, 1990, q This test, when evaluated in conjunction with ST 11.7-13, ' "'B' Switchgear Room Compliance Test," was to demonstrate-the ability to achieve and maintain Operational Mode Hot
Shutdown conditions in the event _'of a fire in any plant location. Also, these tests, in conjunction with a ST'11.7-12, "1989 Refueling Outage Plant Integrated-Test," ' provide the final testing of modifications made to the plant electrical distribution system, the reactor protec- , tion system and the nuclear instrumentation system.
Tests ST-11.7-12 and ST 11.7-13 are scheduled to be performed during'the.next resident _ inspection period.. Test ST 11.7-15 demonstrated the' independence.of the a.c. and d.c. elec-trical eg'uipment located in the "B" electrical switchgear s . room from the rest of the plant'-electrical. distribution system in the' event of a fire within the "A" electrical , , switchgear area. That location for a fire is most severe-in that it may disable the control room, both incoming , off-site power lines, safeguards division "A" a.c. and d.c.
distribution. systems, the "A" and "C" batteries, the "A" '. emergency diesel generator and those portions of the_ safe-- guards division "B" a.c. and d.c. distribution system , (designated BX) remaining in the "A" switchgear area.
The test was performed with all fuel stored in the spent ' fuel storage pool.
The reactor was-last at power on September 2, 1989.
Prior to the test, the security system - and the emergency operations facility power was placed on j their respective diesel generators.
l The emergency diesel generator (EDG) automatic start and sequencing capabilities were disabled.
Then both safe-guards division 4160 volt buses were deenergized, and the supply to the safeguards division "BX" bus located in the . i i
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s , iI "A" switchgear room was deenergized.
This was followed by , , - a comp.lete loss of a.c. power initiated by deenergizing the-
station service bus supply from off site. All_ nonessential . circuits supplied from d.c. buses in the "A" switchgear
room were then deener ' division "A" and "BX"gized._ -This included most of safety-d.c. distribution and all of station service bus "C" d.c. distribution.
Essential circuits remaining energized: included those providing-115ky. line-pro +ection.. At this point control circuit disconnects were opened in . the "B" EDG room which isolated it from the control' room.
The diesel engine was sta-ted locally and safeguards 4160 . volt electrical bus No. 9 ("B" EDG room) and 480 volt bus ' No. 11 ("B" switchgear. room) were energized.
Local control: was taken and the "D" service' water pump, the "C" component-cooling water pump and the "B" RHR pump were operated from the "B" switchgear room.
Contairment air recirculation, fans No. 3 and 4 were also operated from their alternate ' feed from the "B"~switchgear room.
q <Because of the test, spent fuel storage pool cooling was
temporarily isolated for the. duration of the test (about 90-l ' minutes) and pool water temperature heated up approximately
one degree F.
! ' The inspectors observed test. conduct from the control-room, EDG room,-and the.new switchgear building.
All equipment' operated as designed.
It was noted that pretest briefings i were effective, personne1 ~ safety was given. consideration,' ] and the test was executed successfully and withoutLinci-dent. There were no unacceptable conditions identified.
1 4.2.2 Implementation of Revised Technical Specifications , -! , 'On June 3, the licensee implemented the standard-format i Revised Technical Specifications (RTS). The quality, j ' clarity and content of the specifications for the plant are-
significantly improved with RTS. This was a two year < effort which required extensive procedure changes as well as numerous submittals to NRC.
To differentiate between ' procedures that follow the old Technical Specifications ' .(TS) and the RTS, the new procedures are printed on pink . paper. Additionally, the notation (Pink) appears on the t master document index for each procedure.
., The inspector reviewed several new surveillance procedures to verify that the requirements of RTS are satisfied.
Additionally, the procedure technical content and reviews and approvals were confirmed to be adequate.
The following i procedures were reviewed:
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- SUR 5.1-13A(B), " Auxiliary Feedwater Pump P-32-A(B)- -- Functional Test"~ SUR 5,1-17A,- " Emergency Diesel Generator EG-2A Manual -- Starting and Loading. Test" SVR 5,1-157A, " Emergency Diesel Generator EG-2A Fast.
--- Start and Load Test" SUR 5.f.-24, " Safeguards Equipment Timer.. Test" -- SUR 5,1-8A, " Train A CAR Fan Test and PORV Accumulator -- Check" SUR 5.5-33, " Manipulator trane Surveillance" -- The inspectors found the transition to RTS to be : smooth and without incident.during.this inspection period.
No inadequacies -were identified by the pro-cedure review.
4.2.3 Cleaning and Inspection of Reactor Coolant System The cleaning flushes of the reactor coolant systems, which were started during the last inspection period, were completed.- This included sections 6.7 through 6.16 of special test procedure ST 11.1-2,-"RCS Subsystem Cleanup,"
Original, dated April 17, 1990.
This procedure was com- -l pleted on May'22.
Following the cleaningLof outlying sup - . port systems,.the bottom entry instrument guide tube con-- ! duit was flushed into to the reacto'r-vessel, Cleaning and-inspection was then performed of. the reactor-vessel, the vessel upper and lower internals packages, the reactor coolant loops and the pressurizer along.with the. fuel'
transfer tube and reactor cavity floor.
. These were accomplished in accordance with the vendor .i ' procedures, - VP-535, " Upper Internals FOSAR," VP-536, I " Pressurizer 'F0SAR," VP-541',. " Fuel Transfer Canal, Cart, i
and Tube F0SAR," VP-544, " Lower Internals FOSAR," VP-546, j " Guide Tube Flush," and VP-553, "RCS F0SAR."
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i ! It was during the visuai inspection of the pressurizer that-cracks were discovered on its inside surface stainless-i steel ciad. The flushing, c' leaning and inspection of these i components was completed on June 11,
There were no unacceptable conditions identified by the in-spectors during' these evolutions.
The investigation into the clad cracking affecting the pressurizer is addressed
within report section 7.1.
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- , 5.
Emergency Preparedness On May 19,Ethe licensee conducted ine Annual Emergency Preparedness- - p Exercise. This-was a partial participation exercise with the State of b Connecticut and local towns.
The inspectors observed portions of the
exercise;as part of an NRC '.egion I team inspection.
The evaluation of the exercise is described in NRC Inspection Report ~ 50-213/90-09.
6.
Security , '6.1 physical Security Observation During routine inspection tours, the inspectors observed implementa-tion of portions of the security plan. Areas observed included ac-cess 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 implementa-tion were determined to be adequate.
The licensee _ completed and implemented changes to the protected area.
, boundary during this iaspection period.
The change was conducted-on a . backshift and was without incident.
There were no unacceptable conditions identified.
6.2 Fitness for Duty Concern (RI-90-A-0027)
- The regional office. staff was informed _ by telephone on February 7, 1990, by an individual who asserted that support personnel previous-ly involved in steam generator inspections had reported to work af ter-
! consuming alcoholic beverages.
That activity had been-completed and all equipment demobilized at the time that the allegation was received.
-, The information was based on the personal experience of the caller.
The inspection and maintenance activities of the steam generators, including the mock-up-training, were performed from August through ," October, 1989.
Subcontract personnel were used to install and ser-vice robotic equipment.
The assertion was directed towards this group; primarily during their mock-up training.
In a follow-up conversation with the inspecter, the individual stated + that subcontractor support personnel were drunk when reporting for work at the steam generator mock-up on the evening shift.
Addition-ally, he asserted that the contractor supervisors were aware of this i but did not want a confrontation with their subcontract work force.
, The mock-up is located outside the plant protected area and therefore may have escaped the direct oversight afforded other activities.
However, the individual added that some of the workers in question did enter the reactor containment for work when in an impaired condi- ! tion.
,
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, The NRC Regional Office convened an allegation panel and concluded that the. licensee should be informed of a possible fitness for duty < issue. The licensee conducted an inquiry after the' Station Director and the Security Supervisor were informed._ The inspector also discussed the fitnest for duty of. contract and subcontract personnel with members of the engineering and radiation protection-staffs who-were involved With the steam generator work and the mock-up training.
' In ' response to its inquiry, the licensee was informed by its contrac-tor of one incident of possible drunkenness at the~ mock-up in which
the person was discharged.
The contractor did not acknowledge a j problem as pervasive as that related to the NRC'by the alleger.
,-This issue is closed for the following reasons: 1) the work force in question has been disbanded; 2) the NRC now requires and the licensee has implemented a fitness for duty program in accordance with '10 CFR Part 26; and 3) the 1989 issue does not appear.to warrant further detailed NRC or licensee followup.
- 7.
Engineering and Technical Support s The inspectors reviewed selected engineering activities.
Particular attention was given to safety evaluations, plant operations. review com-mittee approval of modifications, procedural controls, post-modification - testing' procedure changes resulting from this modification, operator j , -training, and UFSAR and drawing revisions.
7.1 Pressurizer Clad Cracking 7,1.I' Summary
During the inspection period,-visual inspections made of the j inside of.the pressurizer revealed indications-which-appeared ,
to be cracks.
Those inspections were for cleani.iness and were - made following extensive modifications to the reactor internals - and subsequent cleaning of the reactor systems.
Vendor pro-l cedure VP-536 was used for those inspections.
The-cracks were ' found throughout the vessel cylindrical shell region and around.
the surge line in the bottom head.
Intensive visual examina-tion showed that the heaviest concentration of this cracking i was in the lower portion of the shell and between the two ! , heater support plates.
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, > p ! ' )' -, , ~ 12 ' '! g , ! Based on this information 'a series of nondestructive inspec-i tions were made of that region by manual and automated uhra- ' sonic examination, and by radiography..- All but three reflec-l tors were resolved. Two were allowable in size from Table IWB-3510-1 of ASME Section XI. The third, which was assumed to: ,-
be 0.1875 inch long by 0.35 inch deep, was determined to_be ' acceptable by fracture mechanics.
. A meeting was held between the licensee and the NRC on June. 8
' ^ to discuss this issue.
The licensee committed to providing ( the NRC with the structural analysis for the pressurizer ves-sel, which was contained-in the licensee's' letter to the NRC dated June 29, 1990. The June 8 meeting was summarized by the ' NRC in.a meeting report dated June 20.
l 7.'1/2 Visual Examinations . Because of the. nature of this problem,_ additional inspection.
support was dispatched from the NRC Region I Office.
The _ following-inspection documentation was provided by those in- . ' [ spectors who specialize-in materials and nondestructive exami- 'l nations.
On May 21-23, the inspectors reviewed the licensee's findings , regarding the cracks initially observed during a camera in-spection of the inside of the pressurizer. The= cracks were observed at the bottom of the pressurizer, in an area surround-ing the strainer which.is positioned over the surge nozzle.
, The nozzle is' fillet welded to the stainless clad surface of ' the' bottom head.
The head of the pressurizer was made from-3-3/16" thick A216-WCB carbon steel castings with shell sec-i tions from 5-7/16" thick A302 low alloy steel plate.
The ' strainer cracks were intermittent in nature and, for the most '! part, were located approximately 1" outboard from the edge.of , the fillet weld.
In some areas the cracks were observed in ' the heat affected zone and the toe of the fillet weld. Addi- .tionally, circumferential cracks were observed in the vertical shell; sections adjoining the head. The most extensive crack-ing was observed in the vertical shell-sections between the l first and second heater support plates.
The cracks diminished rapidly in number toward the top of the vessel.
No cracks
were observed in-the top head except for small cracks in the i area of the spray nozzle.
-i i The inspector's-review of video tapes of the visual examina-
tion indicated that some cracks were irregular and some were j straight.
Some staining was observed in the cracks, but there was no discernible evidence of rust buildup.
, !
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ed that the vessel was manufactured by both Westinghouse in-c their,Lester plant and Chicago Bridge and Iron (CB&I).
CB&I l fabricated the shell sections including the cladding opera- [ tion;. Westinghouse fabricated the remaining sections, includ ' ' ing cladding, as well as final assembly.
The licensee's re -
. view of Westinghouse drawings and.available fabrication _ records i indicated ~that the-inside surfaces of the pressurizer had been-- clad with'one layer of type 308L stainless weld metal-except in the area of the. bottom head where two, layers of type 308L_ ' weld metal.were deposited to accommodate the heater sleeves.
_; .Thus, it appeared that the cracking was confined to areas-with ? one layer of cladding which includes the shell sections and- - the area in which the strainer was welded tc the head. ~The licensee reported that both Westinghouse and CB&I employed a ' series submerged arc process, generally characterized by low: dilution effects. -Ferrite enntent of the stainless cladding ' was reported to be an' acceptable level of.6-9 percent..After visual inspection, the licensee proceeded to perform radio-graphy and ultrasonic testing in an area between the heater support. plates, 360 degrees around the circumference of the .
vessel. The res'ults of radiography and ultrasonic testing as witnessed by the inspectors are. reported below.
7.1.4 Radiographic Examination Radiography of the shell sections between the support-plates was performed usino a cobalt 60 source positioned on the.
. inside of the vessel with:the film on the outside surface in accordance with'ASME Section V Procedure NV-RI-1. A more' optimum' technique with the source on the outside and the film-
- .
on the inside was not employed because of-radiation'considera-tions. The inspectors' review of the radiographs confirmed ' the licensee' evaluation that no linear or crack-like indica-tions were'. observed. The radiographs were judged to be an acceptable 2-2T quality level. This_ was not surprising - i because of tightness of the cracks and the thickness of. metal (5-7/16") involved.
=i 7.1.5 Ultrasonic Examination The licensee utilized a Southwest Research Institute (SRI) Enhanced Data Acquisition System (EDAS)~to collect and analyze ultrasonic data. The inspectors reviewed the associated procedures, personnel certification, equipment and material E certifications pertinent to this activity and found them 'o be ' I{ .. . .
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% ' t F y satisfactory.=.The ultrasonic examination was conducted in accordance with. SRI procedure HNP-AUTT-18 which utilized ~an-angle beam technique. The inspectors reviewed and witnessed the associated calibration techniques and found'them to be satisfactory.
t t A review of-SRI's ultrasonic data revealed twenty-six'(26): indications in the cladding in the vertical _ sections of the ( ' shell between the first and'second support plate.
All of the
indications except for three were judged to extend.tolthe.
interface between the cladding and base metal and stop, Three ~' indications extended into the base metal to an approximate t depth of 1/4".
The true nature of the indications in'the base .. E metal could not be established with certaint'y because of the masking effects of ultrasonics indications found in the areas ! ' which were believed to be due to lack of bond at the cladding < interface and/or segregated streaks of nonmetallic inclusion.
A limited ultrasonic inspection of the head in the. area of the
strainer did not reveal any indications.
Inspection-in the
remaining portions of the head was not performed because of l the interference of-the heaters, , o< The following procedures were used for manual and au'tomated' 't analysis: VP-545, " Manual Ultrasonic Indication Sizing Proce-dure," dated May 16, 1990; VP-547, " Setup, Checkout and Opera-H tion of the Enhanced Data Acquisition System," dated May 22,- i 1990; VP-548, " Automated Outside Surface Ultrasonic Examina- , tion Indication Resolution and Sizing," dated May 22,.1990; VP-549, " Automated Outside Surface Ultrasonic' Examination of Ferritic Vessels Greater than Two Inches in Thickness," dated May'22, 1990; VP-550, " Assembly and Checkout of Sonic MKII j Ultrasonic Data Acquisition System, dated May 22, 1990; and, j VP-551, " Manual Ultrasonic Examination of Nozzle Inside Radius [ Section from Vessel Material',".dsted May 29,1990.
7.1.6 Flaw Evaluation In a meeting. held-in NRR on June 7, 1990, the licensee re-viewed the radiographic and ultrasonic results as described above.
Because of the licensee's inability to-establish the true nature of indications in the base metal areas where three cladding cracks appeared to extend into the base metal, the three ultrasonic indications were treated as flaws. Two of the flaws were judged to extend to a depth of.2" and were j L deemed to be acceptable in accordance with Table LWB-3580-1 of ' ASME Section XI.
The third indication (.35" deep x.25" long) was judged acceptable on the basis of a fracture mechanics u R analysis.
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' w - _ 5'[ .15 h , - - ' Several possible causes for the cracking were postulated by the licensee such as thermal stress, stress cc-rosion cracking or thermal-fatigue, with the origin of cracking very likely'- . stemming from-a metallurgical deficiency or; abnormality in the ! ' areas containing one layer of cladding (no cracks were found . 'in the areas containing two layers of cladding).
Cladding
deficiencies or abnormalities could be related to (a) insuffi-- i cient ferrite, (b) excessive ferrite that transformed to a f _ brittle phase (sigma) during post-weld heat treatment,~and (c).. ' ' , base metal dilution effects causing-the formation'ot 6 unique
. cladding composition particularly-susceptible to corrosion Or . weld cracking.
r . With. regard to thermal stress being related' to the cracking, , the licensee reported it may have stemmed from'an incident in-
- which the water level had inadvertently reached-.a low level (lower. support plate).
The possibility of obtaining-a boat sample'or a. core bore sample containing a cladding defect was: . considered by the licensee, but was rejected because of.the ' difficulty involved in' obtaining it and subsequent-rapair.
It should be noted that corrosion of the base. metal. cladding interface where most cracks terminated'was judged to be'of minimal consequence because of pH and oxygen: control.
7.1.7. Conclusion
On the. bases of the evidence presented, it=was concluded that ' the -pressurizer (with the vast majority of the -cracks' con-tained within the cladding,-and a few cracks which may have [ -penetrated into the-base metal to a depth of 3/8") could be - ' operated safely"for-another cycle. As. required ~by Section XI, , the pressurizer will be reinspected during the?next outage
using the same ultrasonic techniques used to characterizeithe defects presently detected in the pressurizer.
- 7.2 Control Room Environment . ! During this inspection period, the inspector. reviewed several aspects of the control room envelope focusing primarily on control room " , habitability. issues.
, i , 7.2.1 Control Room Ventilation System , The primary function of the control room ventilation system is to maintain a temperature within the control room for person-nel comfort and equipment operability. Additionally, the system maintains a positive pressure gradient to limit in leakage into the control room, provides control room smoke removal capability in the event of a control room fire,
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p b and isolates the control room atmosphere under adverse exter - nal' conditions.
The. ventilation system is comprised of one air handling unit,. with a heating coil and cooling _ coil', a fan', bypass and face dampers, and associated duct work _.The cooling coil'is cooled h by service water.
Power to the fan and air _ handling unit is supplied by motor control center 6 which can be manually' tied to the "B" emergency diesel generator.
The air handling unitL and fan are categorized as safety related;.the-service water supply is not.
- A 7.2.2 Control Room Habitability Issues The issue of control room habitability first cppeared in the Haddam Neck Plant' Safety Evaluation Program (SEP).- SEP Topic VI8, Control Room Habitability, concerned plants which did not fully comply with 10 CFR 50,-Appendix A, General Design
. Criterion 19.
Specifically, analyses were required'of the control' room air filtration rate, ventilation system isolationt l and' filter efficiency, shielding, emergency' breathing appara- , tus, short distance atmospheric dispersion, operator radiation exposure; and on-site toxic gas proximity.
This SEP topic was . deleted based on similarity to a TMI Action Plan Item Require- . ment.
.TMI Action Plan, NUREG 0737, Item III.D.3.4, Control Room c Habitability, requires that-control room operators be ade-quately protected against the-accidental' release of toxic and radioactive gases to ensure that the plant can be safely operated and shutdown under-such design basis conditions.
.. Licensee; evaluation of this item concluded that-major modifi- ' cations would be necessary!to comply with the new require-- , ments. To alternately satisfy:the specifics of this item, the - . , licensee committed, by letters dated July 1,1981,'and July' 8, j 1982, to install'the following equipment.
t -- additional breathing apparatus in the control room, j -- a radiation detector in-the outside air intake'to provide for control room isolation from high radiation, and-
i -- a smoke detector in the outside air intake to also provide I
for control room isolation.
. Pending installation by December 31, 1984, and a
post-implementation review by NRC inspectors, this item was resolved by NRC on July 20, 1982.
Breathing air bottles were . placed in the control room and subsequently relocated to the
p 1-
m-~ - p p' y x l 17, , ,. second floor of the turbine building under plant design change-evaluation-(PDCE).CY89-19,.completedJune,1989. Air is sup- , plied to operators through a combination of_hard pipe and hoses to connections near the operator desks'in the control ' room.
Hoses and masks ~are:provided in the-control room'as well as self-contained air packs with additional breathing air.
There are no radiation and smoke detectors ~ in -the control room outside air intake; this item remains unresolved pending further licensee investigation (50-213/90-08-01), ' Preliminary design work for a ventilation system reconfigura-tion was completed and the licensee concluded that Haddam Neck i ' would need new, redundant ventilation and filtration systems i for the control-room.
However, the impact of control room ' space limitations and ongoing changes altered the design ! conditions and required substantial additional resources.. The- [ request was made to defer implementation of this modification.
- NRC reviewed the licensee's evaluation and determined that the , inclusion of this item into the expanded integrated safety assessment was justified and acceptable.
This was reflected . ' - ' ' Lin Haddam. Neck Techni al Specifications Amendment No. 63,
- l-dated July 1, 1985.
The licensee's Integrated Safety Assessment Program (ISAP), f ' dated July, 1987, defines this issue as Topic 1.12, Control 'j Room Habitability.
Because of the high cost of controlcroom
Lheating ventilation and air conditioning (HVAC) design modi-f fications, the licensee proposed alternatives.for resolution j-(mentioned above) at significantly' lower cost.
The NRC ac-'- J ~cepted these methods.
However, the NRC concluded-that-the -[ effect of smoke, high humidity, and high temperatures on oper-ator behavior were not adequately addressed.
The review of this' topic has been completed and is described ~ l in the April 30, 1990 Haddam Neck ISAP Update Report.
The { proposed HVAC modification includes installation of.a_ fully
redundant, low leakage, filtered control room ventilation
y system..The licensee. requested that_ Topic 1.12 be closed with
the transfer of the concern for control room environmental j ' affects on operators to Topic 1.19.
y
ISAP Topic 1.19, Human Engineering Deficiencies, encompasses j the Control Room Design Review (CRDR) performed in accordance.
with NUREG 0737 and Generic Letter 82-33.
The CRDR identified f OER9, Control Room Environmental Control, which states that: { The air conditioning in the control room is inadequate at ' - i temperatures over 85 degrees F.
The temperature is uneven in
the control room. Air is too dry in the winter and too hot in ] the summer.
Ventilation is inadequate." CRDR recommended y that the ventilation system be modified to increase operator s y .
- . , . .. i $} , ~' ' f T, 18 -- -comfort and extend equipment life.
The licensee is proposing [ modifications to install new equipment'or modify existing l' equipment during the Cycle 16 Refueling Outage.(late 1991).
, 7.2.3 Service Water Modifications Affecting the Control Room In parallel with these evaluations, for about two years, the C 111censee has been. reviewing service water system adequacy in L' ' meeting the design basis.
In March, 1989, the licensee im- ~ plemented a design change to the service water system which affects the control room air conditioning capabilities.
Plant, ' ' design change record (PDCR) 955, " Installation of= Emergency Automatic Closure Circuitry for the Turbine Building Service Water Header Isolation Valves, SW-MOV-1 and SW-MOV-2," pro- ,vides for automatic isolation of service water ~ to the turbine' building upon. receipt of a safety injection signal.
Thi s=,i so-lation.previously only occurred in response to a loss of nor-mal. power.
Service water to the control room air handling unit cooling coil is one of the loads which is isolated.
The.PDCR integrated safety evaluation states that, with ser- ,vice water to the cooling coil-isolated, the control room temperature will rise to 102 degrees F, and, therefore, no - equipment,in the control room should bt adversely affected and there is no~ increase in failure probability of equipment in ... the control room. The evaluation references a control room ' i
.. ventilation study performed for the-Appendix R Evaluation.
' The licensee is currently reevaluating this study-and its assumptions to assure' accuracy.
During this outage, ' service water system flow. testing was
- conducted.
Flow blockage in service water to many plant components was identified. The licensee embarked on'an ex-
tensive service water piping cleaning and replacement endeavor ' which included replacement of-the piping supplying the control j room air conditioner.
Removed piping contained significant'
' blockage which reduced system flow.
The restoration of normal i flow is expected to provide more eff.icient control room cool-- ing under normal operating conditions.
! ! 7.E.4 Operational Experience Historically, the control room has been hot and humid during-f the summer months, as reflected by CRDR OER-9 (see detail
6.2.2).. Abnormal-operating procedure (AOP) 3.2-46, " Control
Room High Temperature," provides instructions to mitigate the
consequences of high temperature in the control room or a loss of the control room ventilation fan.
The A0P covers contin-gency actions for plant conditions with and without off-site
power available.
Instructions are given to utilize the I d
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, F ' m computer room air conditioning system and a series of flexible ' ducts and fans to disperse cool air throughout the control
-room if the. air temperature cannot be maintained below 90
i degrees F.
If the control room and main control board temper-I atures'cannot be' maintained at or below-104 degrees F, the A0P ' requires that the plant be placed in hot standby within the next six hours.
! C , , - 7.2.5 Conclusion
1 Considering the additional service water flow available with- 'I the replacement of the supply piping, the effectiveness of the j ' control room air conditioning system will be different-in.ihe ' coming months than that of previous summers.
The inspectors
will continue to monitor conditions in'the control room and " follow the verification effort of the, control room heatup .i . study performed for Appendix R.
7.3 Electrical Modifications 7.3.1 F_urpose The licensee has made extensive modifications-to the electri-I cal systems for compliance with the fire protection.-require- ! ments of 10 CFR Part 50 Appendix'R to' assure a. safe shutdown - ' of the-unit in the event there is a fire at the facility and also to correct several previously. identified single failure
issues.
.This electrical inspection was conducted to ascertain whether .these modifications were des.igned,-and whether. components were
purchased, installed and tested, in accordance with NRC regu-~
lations and the-licensee commitments.
Q
7.3.2 -Scope
The inspector r3 viewed documeats, inspected installed electri-I cal equipment, and witnessed special electrical tests to .. evaluate the licensee activities in. the following areas: -- Design change process ,- -- Drawing control -- Engineering support 1; -- Design changes ! -- Problem reporting and corrective action ci1 -- Security-and tornado protection -- Mechanical and electrical system interactions -- Electrical separation , L ' 1,' ,
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-- Electric magnetic and radio interference with instrumentation systems -- Station batteries surveillance and testing
-- Special electrical-testing ' , 7.3.3 Findings As part of the recent electrical modifications, new 480 volt motor control center (MCC) equipment for motor operated valves-(MOVs) were installed under plant design change record'PDCR' 931.
Circuit fault protection was.provided by Westinghouse , type HMCP motor circuit protectors.
During testing of the -; MOV's of MCC-13 w May 11, occasional trips of these devices
occurred.
This was documented on plant information report PIR 90-95.- On installation, these devices were adjusted to ten (10) times- ' motor running current.
Following their unexpected trips,'they were reset to a value equivalent to twelve (12) times motor-current.
i . The HMCP protective devices provide high speed fault _ current ., protection to the circuit. -Longer time response motor over l - load protection is provided by other components of the motor i controls.
The manufacturer indicates that the devices are intended to clear a fault in less thanzone cycle and prevent ! single phase operation during that period.
'l The inspector found that there was little engineering basis, , for these settings.
They were based a on generic = product: instal- ' 'lation instruction and motor name plate data. This method , apparently failed; to account for the wo_rst case motor full ' , load current of the_ valve' operator and its interaction with these high speed devices.- The manufacturer's literature,
' , which was the technical basis for these settings, did not ' address the specific characteristics of valve operator motors.
As demonstrated by the unexpected trips, valve motor operator reliability is questionable.
, A trip of the device requires thht an operator be dispatched a to the MCC, in this case the "A" electrical switchgear room,- ~ ' open the cubical door and' reset the device in the same manner as a molded case circuit breaker, j The licensee has also experienced a problem in testing the devices at high current but below the set trip point.
The-device has a relatively low continuous current capacity when
compared to its trip setting range.
They have failed when ' fault current was applied for seven (7) cycles which is the ' minimum time for the licensee's test set.
A'ove the trip ' u i
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. 1 L[ setting, the device actuates in-one to two (l'to 2) cycles.and
' is not damaged.
' - Without the ability of testing both above and below the _ protective device trip point, valve motor operator reliability is questionable, i r The engineering basis for the device original trip setting was-t not available at the end of the inspection period.
Also,-the: licensee _ lacked the-ability.to fully test-the devices.. This' ' " issue is considered to be unresolved (50-213/90-08-02)'and ' will be addressed during a future inspection.
,
Other problems involving the electrical distribution system which were encountered by the-licensee include: a. The discoverytwas made that the 125 volt d.c. safeguards < bus "B" to bus "BX" tie breaker was of an incorrect size.
i An 800 ampere breaker specified in the original design was
installed instead of a 200 ampere called for in a design
change. The change in breaker size was to accompany a .: change to a smaller sized bus tie cable.
The installation ' . error was not noted until after turnover for operations.
This was reported internally as PIR 90-66, ' b. A ground accompanied by high current on_ safeguards d.c. bus "A" occurred while preparing to remove d.c._ bus "BX" from service on June 10. The condition'was noted when 480 volt safeguards buses No. 6 and 7 were deenergized and;the emar- , L gency lighting shifted to its_ backup, the "A" d.c. bus.
The ground cleared when the primary auxiliary building-(PAB) d.c. lighting panel wasideenergized. These' lights _ are original plant equipme_nt'and not the' lighting required i e by 10CFR Part 50, Appendix'R. This issue was reported internally'as PIR 90-122.
c. A minor problem,' relating to 125 volt d.c. breakers on I safeguards bus "B," was found in that they were selected and installed without consideration =of derating factors as-signed by-the supplier as the result 'of qualification test- ! ing. Deratings.should have been made at 45.1 percent-for ! 100 ampere, 70.0 percent for 200 ampere and 57.2 percent.
j > for 800' ampere breakers. The-licensee indicated that they- ' i had not.been informed of the results of this testing and > that it was conducted in a more severe environment than , > 's required for the "B" electrical switchgear room.
, The NRC review of these problems is continuing past this inspection period. The inspection findings and conclusions j i will be documented in NRC Inspection Report 50-213/90-12.
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_ < Pr e s-22 7.4 Potential Steam Line Break Near Safety Related Equipmeg
_ The licensee discovered a preexisting condition concerning house heating or auxiliary steam lines within the "A" electrical switchgear > room and the mechanical equipment room which contains control room HVAC equipment. A recent evaluation indicated that an unacceptable environment may be created by a rupture of one of these lines. This- ) issue was documented internally as PIR 90-96.
Based on the current projection for plant startup the licensee l intends to remove this tazard by cutting and capping.the lines outside the sensitive aieas, _ There were no unacceptable conditions identified.
, t 7.5 Followup of Previous Inspection Findings Licensee actions taken in response to open items and findings from j previous inspections were reviewed.
The inspectors determined if corrective actions were appropriate and thorough and whether previous
concerns were resolved.
Items were closed where the inspector deter-mined that corrective actions would prevent recurrence. The follow-ing items were reviewed.
7.5.1 (Closed) Unresolved Item-213/87-05-01: Improvements for Post
. Modification Testing >r-a This item involved-licensee' identified deficiencies in plant U design change record (PDCR) implementation; specifically, post modification testing to verify equipment operability. -Speci-fic PDCR anomalies have been corrected in a timely manner.
To l-prevent future errors in PDCR review, implementation, and. test- ' ing, a program was developed prior to the 1987/1988: Refueling Outage. The. controlling procedure is ACP 1,2-3,2, "Adminis - tration of PDCR. Turnovers Preoperational Testing, and Release for Operation " In 1989, a corporate level pengram wa's imple-mented under NEO 7.04, " Administration cf Plant Design Change Turnover and Preoperational Testing." The inspectors have reviewed designs and observed preoperational testing for_many 'PDCR's since' ACP 1.2-3.2 implementation (during plant opera-tion and two refueling outages).
Specific attention was paid to the adequacy of test scope, performance, and results.
The administration of the PDCR process was also reviewed.
Deficien - cies similar to those identified frior to this program were l not observed.
The inspectors have concluded that this program
.l has'been effectively implemented to ensure adequate modifica-tion testing and release.
! 7.5.2 (Closed) Unresolved Item 213/87-08-02: Implementation of Preoperational Testing Program . kw } ,
fm , M9 ' ' - , ' - ,. '
, s , (' e n .This item concerned investigation and resolution of deft- ! ' ' ' ciencies in a design change to the reactor coolant system l ,. , temperature detectors.
Specifically, design change open items ' ' ~ . were not brought to resolution in a timely manner. This issue-
" was previously reviewed in NRC Inspection-Report 50-213/87-12.
The licensee has developed and implemented guidance for design
change completion to adequately' address this concern.
This
program is discussed in Section 7.5.1 of this report.
7.5.3 (Ciosed) Unresolved Item 213/88-06-01: Implernentation of ! a_ Corporate Level Procedure for Design Change Turnover and ' Testing , This item concerned the licensee's resolution'of difficulties i >
in administration-of ACP 1.2-3._2.
Action regarding-this issue t was pending the issuance and implementation of NEO 7.04 _NEO
' 7.04 has been implemented and ACP 1.2-3.2 revised accordingly, i This item is also discussed in Section 7.5.1 of this report.
8, Safety Assessment and Quality Verification t
P.1 Plant Operations ieview Committee ' i The inspectors attenjed several plant operations review committee (PORC) meetings. Technical specification 6.5 requirements for.re-L quired member attendance were verified.
The meeting agendas included i procedural changes, proposed changes to the Technical Specifications, ' plant design change records, and minutes from previous meetings. The , 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 com-mittee members time for further review and comment' Distenting opin-t . ions were encouraged and resolved to the satisfaction of the commit-
tee 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.
^! A joint PURC/ Nuclear Review Board meeting was held on Jane 5.
The
f meeting was convened for a presentation and discussion of the Primary
' System Cleanup Program.and the status of program act1vities and-results.- The inspector attended the meeting and found the presenta-tion to be a well-detailed synopsis of the reactor coolant and associated system flushes and inspections.
t 8.2 Review of Written Reports Periodic and special reports, 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 inspetors determined whether further information was required.
The inspectors also verified that the reporting requirements of <
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9,e i y: 1 gg, , , , p E p O 10 CFR 73.71,. station administrative and operating procedures, I tecurity procedures, and Technical Specification 6.9 had been met.
The following reports were reviewed: SER 90-!05, Safeguards Event Report j -- , ' ' Haddam Neck Plant Monthly Operating Report 90-04, covering the -- period April 1, 1990 to April 30, 1990
. . .t ~ I Haddam Neck Plant New Switchgear Building Bimonthly Progress ,i -- L Report No. 22, letter to the NRC dated May 31, 1990 . p Haddam Neck. Plant Findings and Potential Concern Regarding ASEA ) -- i.
Brown Boveri Inc. Breakers, letter to the NRC dated June 5, 1990 ' No~ unacceptable conditions were identified.
-- ' ! - > - Tracking'of Plant Conditions to Support Oyerational Mode Changes l 8.3 , ,, A system to manage and. track the details of system and component ! status has been implemented to support the recovery of the plant from ,. the current extended outage.
Identification and tracking has been
' provided for work orders and various project assignments in the ! system being managed by a senior staff engineer.
The-inspectors have noted frequent meetings are being held with l station management including representatives of the various working i and operating departments. The status of individual items is dis-cussed with that group along with its assignment for completion as a l prerequisite for a specified Operational Mode change, a This; system appears to be a valuable management conuol based on the .: wide scope of the present outage and to the issuance of the Revised ! Technical Specifications. These were written in the standard format
of NUREG-0452, Revision 4 and are a significant change from the pre-vious " custom" specifications.
There were no unacceptable conditions identified.
, , ,
Management Meetings [ 9.1 - Exit Interviews . During this inspection, periodic meetings were held with staticn t b management to discuss inspection observations and findings. At the ,j h
- o . . , . . - .
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i .
i ' ? close of the inspection period,-an exit meeting was held to summa- [ rize the conclusions of the inspection. _No written material was ' m given to the licensee and no proprietary information related to this ' inspection was identified.
,
- "
In addition to the exit meeting for the routine resident inspection, l the following meetings were held for inspections conducted by-Region l I based inspectors.
' ' e . t ! > Inspection Reporting Areas
Report No; ' Dates Inspector Inspected '213/90-09 May 18 - C. Amato Annual Emergency May 20 Preparedness Exercise
.i 213/90-10 June 4 - P. Bissett Operator Requalification [ <: June 8 Examinations - rg 213/90-11-June 4 - T. Dragoun Radiological Controls June 8 .; .; 9.2 Outage Activities ~and Restart plan Meeting . , On May 8,"the inspectors' attended a meeting between licensee and NRC' , representatives in the NRC Region I Office.
This meeting was held ^! to. discuss the plant outage activities, restart plans,'and the Northeast' Utilities Fitness For Duty Program Attachments 2, 3, and 4 to this report were presented by the licensee during the meeting, j s i 9.3 Systematic Assessment of Licentee Performance Management' Meeting ^ ' On May 10, the_ Systematic' Assessment of Licensee Performance (SALP)_ { Management Meeting was held at the Northeast Utilities Corporate v Office in Berlin, Connecticut.
Representatives of the NRC and the licensee attended the meeting as well as the resident inspectors.
' The meeting agendas included discussion of_the SALP process.and the ) , Haddam Neck 1989/1990 SALP report results.
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- Attachment 1
, , ~ l: ' ~ ) Technical $pecifications verified for Operational M' ode 6:
. f-3/4.4.1.6 Reactor Coolant System (RCS) Isolated Loop-
' n m
h ' 13/4.4.1.9 RCS. Idled Loop .i ' e L ' 3/4.4.1.11.
.RCS Idled Loop Startup~ ' ' - < L !
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' Primary Auxiliary Building Air Cleanupj System (PAB). .f iu > ' . .3/4.8.1.2-Electrical? Power Systems A.C. Sources
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' 13/4.8.2.2- . <; . . . . Electrical Power Systems D.C, Sources ! n, y .
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3/4.8.3.2 Electrical: Power Systems-On-site Distribution-l r- . 3/4.9.1 Refueling Operations - Boron Concentration j n . , t 3/4.9.2! Refueling. Operations - Instrumentation = '
- 3/4'. 9. 4
. Refueling Operations - Containment Penetrations: ! ~ ,g , m 3/4'. 9 i 5 : Refueling' Operations - Communications ' ' ' , 3/4.9.6_ Refueling Operations - Manipulator Crane.
~ a i3/4.9.8.. Refueling Operations - Residual Heat Removal 'and Coolant Circula-tion..High and Low Water Level; !! . I 3/4.9.9-Refueling Operations - Containment Purge and. Exhaust.lsolation i - m System ' , P.
L3/4.9.10'- Refueling Operations - Reactor Vessel Water Level
3/4.9;11.
Refueling Operations - Irradiated Fuel Storage Pool Water Level .j L 3/4.9.12~ ~ Refueling Operations - Fuel Storage Building Air Cleanup System - ! '[t . i ' - . , -
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CONNECTICUT YANKEE MEETING WITH REGION I ' AGENDA ' L
l l O Introductory Remarks (Eric A. DeBarba) ! . O Maintenance and Refueling Outage (Gary H. Bouchard) ! O Startup Issues Appendix R and Integrated Testing (Michael II. Brothers)
' i
Reactor Protection System and Nuclear Instrumentation System Testing (Jere J. LaPlatney)
Revised Technical Specifications (Eric V. Fries) _j
Fuel Reconstitution (Michael P. Hills)
Fuel Integrity Monitoring Program (Gunti Goncarovs) f
Primary System Cleanup Program (Michael F. Marino)
, ! . ' t i
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CONNECTICUT YANKEE. MEETING :WITH REGION I . , AGENDA ' . t k . i l.
O Licensing Issues (Gerard P. van Noordennen)- l " ! . O Fitness for Duty Program Overview (George Malchiodi)
I , L O Summary (Wayne D. Romberg) [ l ! l + i ' i , ' ! ! .: t I ! ! I ' i l '. _. - _. _. _,. -. ,..,., .-.._-_........ _,._..... _...._.. _ ...~....... .. . . _ ... _._.,
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INTRODUCTION
I t' , !
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t t ! Eric A. DeBarba , . i i Station Director , i
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-
- - - - . . i INTRODUCTORY REMARKS , O Connecticut Yankee Station Director effective April 28,1990 l Station Services Director at Connecticut Yankee since
l January 1,1987 i System Manager NUSCO Generation Mechanical Engineering
1981-1987 !
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- L INTRODUCTORY REMARKS , .!
i ! ) O Connecticut Yankee's mission for the future is to continue with and
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strengthen our commitment to: . ! The positive culture changes that have occurred since 1987
I Rigorous self evaluation and self improvement
! Maintaining open working relationships with regulatory
l-groups i Nuclear and personnel safety
i Building and maintaining strong ties to neighboring
! towns and communities t
[ i I , -_.- _ _ _ .... _ _ _.... -. _.. ., . _. - _., - - -. -. -., , _.. _ .... .,, _,. _ _ ., _. _ _. _ _ _. _ _ _ _. _ _ _ _ _ _ _.... -
- - - - - - - - - - - - - - - - - - - - _ _ _ _ _ - - _ _ - - _ _ . .
INTRODUCTORY REMARKS '
! !
! ! ' ! Connecticut Yankee's challenge for the
1990's is to continue to make excellence l our standard and take those actions , l necessary to meet this chaIIenge r i l ! ' ! ' ! ! I . .-- . - . -. - .-. . . -. .
. . < . . j ! '
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! MATVTENANCE AND
REFUELING OUTAGE t i !
... i ,
i ! ' . l
I ! l ' i !' Gary H. Bouchard Unit Director . l
... _,. _ ,,, ... _ ~.. _, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _... _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _.. _ _ _.' ... _ .
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MAINTENANCE AND REFUELING OUTAGE ! , t .
! O Outage Schedule Overview ! ! O Thermal Shield Status i
O containment Status -
l O Service Water Status ! j O OperatorTrainingInitiatives to Support Startup .
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4 j- , ! OUTAGE SCHEDULE , OVERVIEW . l l , y . -rs-+s -- m ,y 4- ,.m.-, m '- w-wn- ..v-- - n -as4a w,,.. m>-+ - n-w, .,s_nw_ .w,-s,-_. - _ _ _-_~_aea_m .
_ _ _ - - _- _ - _ - - - _ - - - _ - - _ _ _ - _ - - - - _ - - ._- - _- _____-. __ . _ _ _ . . a v . OUTAGE SCHEDULE OVERVIEW . . . m
O Projected outage duration was 58 days from 100 percent power to j . i 100 percent power i O Outage began September 2,1989 i
O Current projections I . Phase to Grid June 27,1990 l l 100 percent power July 5,1990 (306 days) ! l
I
! ! !
. ' , I i
. . . _.
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- - - - - -
-- - - - _ - - - _ - -_ __ . _ . . $ e - < f . : [ REFUELING OUTAGE SUMMARY l REMAINING CRITICAL ACTIVITIES l . Owaage Cycle 15 i May l Jun lJul
6 l 13 l 20 l 27 l 3 l 10 l 17 l 24 l 1 l5 ( i , Perform * t, -- Fheshes E - ?: - ^_^_ Core Barret ^^ f ~-- -.:_ ---:; : :^'- E ^' ^ t i insert Core Barrel into Rx Wesset/Cleese Re0ueling Cavey
k !
Perform Pre-Fuel Load Appendiu R C _. f_____e Test { j (Requires Total Station sumrenne) l t ' ! Drain Cawlty/Fhssh Incore Flum Thirnt:4es, l ,
Pressuruer, Pzr Surge unes and For Spray unes E Post-9000 ECCS Flour Test l l Per or Reactor veesei. RCS % up,.r to er ! Interness hispecmonsionsten Equipawne Hatch f ! Reactor Core Resomeorain Cavey E New Seechgear Room Compsonce Toegng l Reassemese Reactor 3- ! Post Coreload Plant integrated Tessing ' i j CIoss A Integrated Leak Rate Test
j Perform System Line-ups/ Plant Heelup i j Phone to Gn@Asceneson to 100% Power l
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THERMAL SHIELD STATUS . i
i
O Based on damage observed during inspections performed this outage, the decision
was made to remove thermal shield and associated support hardware '
- Excessive clearance on upper limiter keys
, !
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- Broken, missing and displaced hardware
. . O All removal work performed in the refueling cavity under flooded conditions ! t
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- . .. . m k THERMAL SHIELD STATUS , i i i l O Removal sequence
'
Shield divided into three approximately equal (120*) sections i l Each section moved underwater to cutting stand in cast end of
' refueling cavity
Section then cut lengthwise into sub-sections based on predetermined
! plan , , Sub-sections loaded into liners and then casks for shipping to !
! Barnwell, S.C.
! ! ! ' !
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. . . - l . THERMAL SHIELD STATUS i
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! ! ] O Remaining work
i l Remove lower support blocks and associated hardware (on-going)
' , , l Remove limiter blocks (on-going)
Install flow (cover) plates
! - t
Partial removal of damaged specimen basket ' , ! NDE of core barrel
! ! 'i !
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- THERMAL SHIELD STATUS
i
1 i i O Personnel exposure-
! Projected 91.5 man-rem j
Actual to date 75.4 man-rem i q i i ! I ' ! O Radwaste generation ! Actual to date approximately 2,450 cubic feet
, ! ! Total station goal for 1990 is 8,500 cubic feet
.
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_ _ . _ _ . _- ' ~ ~ ' _, - - ' THERMAL SHIELD. STATUS O Actions taken during thermal shield removal and cut-up.
to avoid introduction of debris into RCS - I Major cutting processes do not generate -
machining chips . MDM generates a fine dust which was . l vacuumed to filter system during cutting I Plasma cutting generates a slag which was collected in the cutting tank-
Debris barriers installed over critical areas , Upper internals Transfer canal Under core barrel
Reactor vessel .
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- . . .
. L THF,RMAL SHIELD STATUS . ! .
Secondary cutting of thermal shield segments ' performed in a " tank" i
Minimal amount of machining performed under j strictly controlled conditions l
Debris pans installed under machining
operations
Vacuuming performed during machining
Comprehensive RCS inspection to be performed to verify cleanness a -- . . -... . . -. -.. - . . - . - - - -. a. - -. -. -.- .. - - -
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CONTAINMENT STATUS
i a [] Programmatic enhancements
, ' Upgrade ILRT and LLRT procedures.
'
'
Upgrade containment boundary identification
drawings - ! Improve PMMS equipment database identification
of containment boundary components !
, Upgrade the Operations procedure for
establishing containment integrity - ~ i
' Improve containment isolation and boundary j . valve identification and locked valve checklist
Clarify containment boundary as described l
! in the FSAR.
! l l Upgrade the Retest /Furactional Verification. l
' procedure to include requirements for CIV l ' and associated containment boundary components , i ! .. - . . . _ _ _,. _ -...... _ _ _., .,._. _.... _ .. _. _.,,... _. _. _.... _.. _. _.. _, -.,. _ _.. _ _ _...... _
- . . . __ CONTAINMENT STATUS -
, ,
O Past performance enchancements i ' The following penetrations mcdified during the 1987 outage
, ! to reduce leakage and increase compliance with design / testing i J requirements
' i ' o P7 (RCP seal water return) o . ' o P10 (Letdown) o P23B, C, D (Containment leakage monitoring) < ' o P30 (Containment heating steam) , o P38 (CCW to RCP thermal barrier)
o P41 (Loop drain header) j o P60 (CCW to neutron shield tanks) . i
l- - - - - - - - - - - - - - - - - , __ . . - - .. - __ . . . . _ _ _..,_ .. _ _ _ _. _. _.. _ _ _.. _ _, -
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, CONTAINMENT STATUS- ,
< O Past performance enchancements (Continued) ~
Enhanced corrective maintenance on " problem" penetrations
i Mid-cycle testing of " problem" penetrations
Introduction of Volumetric LLRT testing methodology in 1987
l l Full pressure ILRT in 1987
Penetrations P71 (primary vent header) and P80 (auxiliary
containment spray from fire water) receive increased corrective maintenance (cleaning) and testing following operation i ! J i ... -- ,.... -, , _.... .. ,,.,.. ... - .. =.. - - - - - - . . - .- .
__ - _ _ _ - , } CONTAINMENT STATUS Q Present performance enchancements Penetrations 74,75,76, and 77 (RCP seal water supply lines)
have been modified this outage through installation of new-
isolation valves, check valves and test connections o Increase compliance wian Appendix A design requirements , Reduce historical leakage i o '
Penetration 63 (neutron shield tank fill line) has been i modified this outage by removing CIV and cappmg pipe
i
increase compliance with Appendix A design requirements o
-l ,
Penetrations 34 (CCW return from RCP thermal ba rier) ! and 68 (primary water supply to containment) converted from liquid collection to volumetric test method Increase comphance with Appendix J test requirements o j i Full pressure ILRT this outage i
. I . Appendix J Exemption Request clarifies proposed test
methodology for all penetrations
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^a A '. A b CONTAINMENT'- STATUS.
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Q Future performance. enhancements
j .-
- Resolve all Appendix J Exemption Requests
1 Continue comprehensive review of containment - * boundary components other than valves ! i Further physical penetration modifications
-as appropriate .
Update penetration P&ID drawings
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-1 _ _ ' g ~ CONTAINMENT STATUS O Based on LLRT-results this outage, we expect the ~ . 1989 as framd " minimum path" ILRT test results will-be wit'ce.c 3 - v'ical E.3cification limits - a
O Conta%mr etegrat 1 Leak Rate Testing enhancements [
Use of Maes Point test method - -
Increased training for test personnel i
l Improved test boundary identification and line-ups d
! Increased test instrumentation
, Additional Dew cells
.
Additional RTD's Computer' data acquisition system o
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. ~ CONTAINMENT : STATUS - , n - , n
, d Q Connecticut Yankee leak rate program goals ' Increase confidence in containment integrity .
- -
o Resolve Appendix A and Appendix-J issues -
Return ILRT to 'a normal schedule
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SERVICEJWATER STATUS O During.the summer of 1988, increasing river water temperature prompted- , concerns over the service water system's cooling ability ' Original plant. design' basis for maximum service water- -
system inlet temperature was 85"F Extended warm weather caused unusually elevated river
, temperature Justification for Continued Operation prepared to support operation
i with river temperature.up to 90*F Detailed engineering evaluation begun
i O The detailed engineering evaluation led to other concerns regarding the system's ability to meet original design requirements The majority of concerns were addressed on c.n as-found basis to.
- support continued plant operation l
Remaining concerns are being addressed prior to upcoming plant startup - I
j
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- = _ . _..
, .Q lSERVICEDWATER STATUS ' ~ _ li Q Tasks completed prior to shutdown Detailed system moddling and comprehensive
flow analysis , - i . Physical plant modification and administrative
l controls to remove or limit non sacfty-related . l loads during LOCA without LNP
. l Clean CAR: coolers of flow blockage
l ! Replacement of CAR fan motor coolers -
^ * Increase of system design basis to 90*F .
Special flow test' conducted to verify heat removal capability of SFP cooling system - prior to core offload . . E v-- , .s - s--t+ - w.-.r+#-.--e- -~a' y .. * - -wr -- (-- - g-. -
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_ ..E-d K SERVICE uWATER STATUS- " - 1.. . . ' O Outage activities Total system flow l tests and balancing conducted i
i s ' Extensive local instrumentation added to permit data collection l o I
- a l
Results of testing to be compared to system analysis as a ! means of validating analytic model projected flows and j l pressure drops Initial test results under evaluation.. Mechanical cleaning ' o of 6~ inch piping to CAR units and. partial replacement of _ . ' 4 inch piping to emergency. diesel generators required along . with possible adjustments of flow model - Test will be reperformed following resolution of initial o discrepancies i , . o i ! ' _ m- , ._.
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. Appendix R flow test Demonstrates the ability of the system to supply critical-o post-accident loads using only the D service water pump ' Provide additional validation of analytic flow model; ~ o o Initial test results also under critical evaluation. Corrective . actions underway and planned O Test will be reperformed following resolution of . initial test discrepancies
Performance flow;te'sts on Service Water Pumps. Test'results on A, B and C pumps unacceptable and , under evaluation . ' o-Test results on D pump fully acceptable - ,
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. -
. . - . - . . - - . .. _ . . _ . - s
- ... . - _ l SERVICE WATER STATUS i
, Installation of auxiliary service water pump to assist with
secondary side cooling loads ~during summer months
a
Upgrades of various system instrumentation to assist in
performance monitoring
i l Installation of MOV bypasses on each service water filter to .
. permit filter bypass dunng accident conditions with high containment pressure.
'
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! SERVICE WATER STATUS- - l
O Resolution of total system flow test and Appendix R flow test discrepancies; i ,
Mechanical cleaning of supply and return lines to CAR fans
! ! , Replacement of accessible portions of supply and return lines to l
. Emergency Diesel Generator coolers
! Initial results to be used to evaluate analytic system flow model !
!
Both flow tests will be reperformed and results reconciled with system flow model !
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. SF,RVICE cWATER STATUS
O Service Water System Safety System Functional Inspection (SSFI).
i
Internal SSFI of system conducted from January 8,1990 o to March 12,1990 - - .
The purpose of the inspection was to determine if the - systemfwill perform the safety funetion required by the ,
system's design basis t j = Inspection areas- . ' o Testing (preoperational, post modification and surveillance)
o Maintenance Operator and functional training.
j o o Man-machine interface.
, o Management controls y .. l . - .-=:2_,, . -. -..- a.
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. i
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Major'SSFI observations
a r , ' o . Service Water system pump modifications made throughout plant i life raise design control concerns = . .. ! ~ i Service Water system piping downstream of various critical-o components was not identified as quality-related or seismic ' Potential flow restricting deposits present in various areas of.- o
system piping and components "q l SSFI observations under review and resolution in progress
Overall assessment of system's ability to meet safety function design '
basis still under review '
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- - -- ' _. - 1[- -- ' ' ' - _ .. + - - , , ' -- LOPERATOR:HTRAININGL ~ = != 'INITIATIVESJTOJSUPPORT STARTUP O Appendix-R Modifications
Classroom training o-New AC/DC distribution system New inverters and' distribution system O CAR fan alternate power supplies and transfer. switches - New operation outside Control Room and fire mitigation procedures 1
' Remote mstrumentation and hot shutdown panels , New steam generator pressure instrumentation - , s
1 -,
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INITIATIVES TO1 SUPPORT STARTUP? ~ f - g.
_ , j , .; _. _ -. - -
In plant training _ '
, -
' Perform rhek-in/ rack-out of new ~ breakers on Bus'11 -! - , Startup and transfer inverters-i o
-Operation of metering pump at remote - ! instrumentation ' panel' l , l l Cooldown to hot shutdown outside Control Room' l
Transfer CAR fans"to alternate power ! - supplies ! a
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INITIATIVESLTO? SUPPORT STARTUP
.
Simulator training
' LLoss'of Vital Bus B / D -
j Loss of DC Bus A / Bus B j Loss of' semi-vital
' ~
- Loss of MCC-5
. . .
New'switchgear modifications, both
- AC and DC changes' have been. -! installed on simulator.
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- - l . - . .. OPERATOR l TRAINING INITIATIVES TO ; SUPPORT STARTUP.-
NIS Modifications
Classroom training New NIS Control Room instrumentation Arrangement, operation and' applicable Technical Specifications Dilution to criticality ._ ---
, .. - . . . . ,, . - - . . . , . S
OPER'ATOR-TRAINING 1 -
INITIATIVES. TO. SUPPORT 1STARTUP.
,
In plant training -
i ' Scaler timer and shutdown monitor - operation i
4 Power range calibration
-
QFTR calculations
" Axial-offset calculations ~
. i
Loss.of NIS ,
NIS modifications scheduled to be installed ~on simulator during j Summer,1990
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, - LOPERATOR LTRAINING ~ = - - - INITIATIVES TO1S~UPPORTLSTARTUP: - O - RPS and ECCS Modifications. .
Classroom training RPS' logic racks
' Trip signal inserdon New flow indication and logic -'New containment' pressure transmitters and logic - New wide and narrow range' RCS pressure transmitters and functions
NewiECCS valves and power supplies LRevisions'to Emergency Operating ' Procedures . . . LS 1:3 (Transfer to Sump Recirculation).- ES'l.4 (Transfer to Two' Path Recirculation) - 44+--4.. e meds- .---A.
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INITIATIVES TOESUPPORTL STARTUP.
" . i -
In plant training .j o Simulate use of ES 1.3 and ES 1.4 l
Simulator twining RPS logics and applicable o
Technical-Specifications
o Use of ESL 1.3 and ECA 1.1 (Loss of l ' Emergency. Coolant Recirculation) . ' .
New ECCS valves, RPS logics (software . only) and new indication have been ! installed on simulator ~ - :i
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' " ~ ' - ,, _ c _ , OPERATOR TRAINING INITIATIVES :TO.: SUPPORT STARTUP . O RevisedTechnical Specifications
. Classroom training conducted on.
Revised Technical Specifications ' ' ' Revised Tecimical Specifications have been used in all' Simulator training since September,1989 l l '
Requalification examination-given using Revised Technical' Specifications L l ! = i.,_ _,. _.. _ - .._, __. _, ... , - __-;m.-__._-._,._ ..,._,___y,.. - __ _..,. ~,3. _... _ _ _ _ _ _ _ _.___ _ ___1_______ _ _ _ _ _ _ _ _ _ _ _.. _. _, _.__, __,.,____.____1___ . _____._..fy _ ,_ _
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- - - ! APPENDIX R TESTING l O The purpose of.the-Appendix' R modification'is to provide a separate protected electrical train to assure the~ capability to achieve'a safe shutdown under analyzed fire scenarios f O Equipment and components added under the Appendix R Modification ! B Train station battery . Non-Cat IE battery bus
480 VAC Bus (Bus l1)
Two MCCs'(MCC 12 and~13)- . Remote monitoring and shutdown instrumentation
B Train 120 VAC-inverters . Relocated the primary source of 120 VAC semi-vital power to MCC 12
, New seismically qualified, environmentally l designed building to house . remote instrumentation, Bus 11, MCC 12, and the new battery and DC bus
Building provides ready access to new and existing electrical components -. - - - . - . .~ - .. -. . . - . - -
-- - - - _---- - - - - - - - - _ _ _. _ _ _.
. -.. , ~ 90A APENDIX"R" TESTING O Component level (Phase I) testing conducted on all new.cquipment O System level (Phase II) testing included the initial energization and. operation of project equipment and provided an overlap of , component level testin'g O Integrated (Phase III) testing Appendix R service-water system performance test
- Demonstrates the. ability of the D service water pump to adequately supply a composite lineup of safety-related system loads , x n-..- , -. .. _., n -.~ a, e.,w,,s sgr __,,,,,,,,. s l,.,,,.,_ ,..,. ,,,, _,, , _ _ _, _ ,_ _, _ _, ._',,_,__ _ ' _ _ ' _
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- _ . A.Q t- + .- _ . - - . APENDIXORETESTING- - + b . O Pre-fuel load Appendix R compliance test-
Demonstrates the ability to mitigate the consequences of a fire in ' the 'A (old) Switchgear. Room l > O-Post-fuel load Appendix R compliance test
Individually demonstrates the ability to mitigate the consequences
of a fire in the Control Room and the B (new) Switchgear Room ' i - + i '> . pu mes ,w.J y ,ph =c = 4M"4ege%t-. nevwf g
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- . - _ _ . r , . ,, __ ~' ~ - , - - % PLANT :INTnGRATEDL ; TESTING -
.i O Plant Integrated Testing encompasses the following operational.
, scenarios ] Loss of Normal Power (LNP) and Safety Injection Actuation
, Signal (SIAS) on A Train ", LNP and SIAS on B Train.
High Containment Pressure ~(HCP) on A Train
HCP on B Train , . Simultaneous LNP on A~and B Trains !
Simultaneous SIAS and LNP on A and B Trains
a ! ! ~f
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PLANT! INTEGRATED -TESTING.
- .
.
i D Reactor Protection System Electromagnetic Interference (EMI) testing..will- .; ' . verify a~ lack of power cable ~induc~ed electromagnetic interference;on -
instrumentation channels.; Testing will encompass '
Shutdown Testing
- .
- o' Short-term test (4 minute" duration)
o' Long-term test'(4 day duration) ,
Startup. Testing
. . .
- q o-Short-term test (4 minute duration)
- iy Plant Operating Testing '
Short-term test (4 minute duration)- . Long-term test (4 day ~ duration) > ! , Radiated Magnetic Field.EMI measurements
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! . ... a PLANT EINTEGRATED -TESTINGL -
i EMI Testing Data Reduction j
Data from clamp-on meters will be plotted in scatter plots to l illustrate the relationship frenuency and peak to peak amps
_ l ' ! l ~ l Photographs of oscilloscope displays will be obtained for various-representative sampling' periods l l ! ' i si
Overall, the. observed magnetic / electrical noise. envircnment will be j verified to be enveloped by Foxboro Qualification Test Report No. 88-1033a - 't
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Switchgear project is complete '
. The Appendix R Service Water System performance test has been
' l performed and will be reperformed following resolution of l dehciencies from the initial performance l ' ! ! ! r i The Pre-Fuel Load Appendix R Compliance rest is scheduled to
he performed on May 11,1990
i ! . The Appendix R Compliance Test is scheduled to be performed j ' . j on June 6,1990
i ! The Plant Integrated Test sequence is scheduled to begin on i
i June 8,1990 i i
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. - .-
. . REACTOR PROTECTION SYSTEM TESTING
. .
, t !
- SUCLEAR
i i L IS'STRUMKNTATION SYSTEM TESTING ' i ' l ! l Jere J. LaPlatney l
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REACTOR PROTECTION SYSTEM BLOCK DIAGRAM '
J i
!
I f i ! ! Primary Foxboro Process Foxboro Logic Reactor , , , , Sensor Cabinet Cabinet Trip Breakers ' ' ' ' I !
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W l Process Engineered Controls + Safeguards - l Systems , L k i k l t i . ~-
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) i l FOXBORO PROCESS CABINET SCHEMATIC
l l l ! i i ' ? i Primary Sensor .!
I i ! . . ! it i i
i Test 1/V Converter Spec 200 Micro-i . . i Switch Isolator Logic / Control g Isolator , ' '
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a . i t 1 r , t Process Foxboro Logic ' Process Cabinet Boundary Controls Cabinet ' , i .
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FOXBORO LOGIC CABINET SCHEMATIC L 1, i l ! - t -. i i Process
Cabinet ,
P i ' ' - - . , t i V i ! !
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Switch Tdp bgk _ ! ! I i i
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t Jk -. V V , t Engineered Reactor l Logic Cabinet Boundary Safeguards Trip Breakers l Systems ! , ! t i I , --- .~,--m-- .- --~-.,~.#- i-w >w ~e.-- -e-~ ,^- + r= -- - - .
. - ~ . - M - . . _ _ _ - d , L REACTOR PROTECTION SYSTEM TESTING . O Testing Summary . ! .
Factory acceptance j i l Construction Verification
' i
Scheme Verification
l Component Level Testing (Phase I)
! ' .
System Level Testing (Phase II) ' , ! , i i ! !
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__ __ - - - , ._ s > - - - ~ i . . . REACTOR PROTECTION SYSTEM TESTING U ' - , ! l O' Factory acceptance te; ting consisted of cabinet ! point-to-point wiring verification along with , j scheme verifications and functional testing using . ! ! artificial input signals.
. i ! O Construction Verifications t '
i l
- Tube leak checks
- Logic cabinet and protection rack l
l point-to-point wiring checks
Root valve inspections l
Power supply testing , ! t I E k l ! . , I . -... _ _.
. ... -... -.. -. . -..... .. . - .. - - - - - - --..---~---- - ---- - - - - - - - - - -
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - _- _.
. , i
. . ! . ! REACTOR. PROTECTION SYSTEM TESTING i
- a j j i O Scheme Verifications ! '
Logic cabinet rack hand switch i ! checks !
! l Logic cabinet and protection rack . j loss of power checks Nest Level f I I Cabinet Level i l
Logic cabmet trip checks . .. . i i l
Component Level (Phase I) Testing f ' ! ! Individual loop calibrations using Technical Specification .
surveillance procedures - l
' c l-Overlap testing conducted using modified surveillance .
} procedures ! ! '
_ _. , - -. ..,. - - . . .. . ... ..... .. ... - . - - - - . - - -
, ._ < - , ~
- -
m- _ . L REACTOR PROTECTION SYSTEM TESTING-e .
O-System Level (Phase II) Functional Test i
- T cold interlock
LTOPS interlock
i Logic cabinet coincidence logic
j-
Reactor trip logic )
Permissive verification , l - SI actuation and block logic
.
RCS low flow trip i
Protection rack analog operational channel checks i
- Time response testing i
- Testing carried out using Technical Specification
[ ( surveillance procedures modified as required to l j provide overlap-
l
O - Reactor protection system integrated testing has demonstrated channel integrity from sensor through l protection and control functions , . , t u ,.y -
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ ______ _ _ -___ _ _ _ _ _ _ _ _ ._.
__ - . '(.. NUCLEAR INSTRUMENTATION SYSTEM BLOCK DIAGRAM
! Inside Outside Containment Contabment _ .
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aTrain A s! WRI-- - I fjeg;. e x=::A --:g 4 .- - Located in Main oA r - m- - 4-Contn>I Hoard - . - W' +~w- ~w~ Jer:p % . (Located in Train B j Switchgcar Building) . . . - , .. . - - -... -.. - . - . -. .. - -. -.....-. -
__--_----_--_-_ _ _ -__-. - . - .
' NUCLEAR INSTRUMENTATION SYSTEM TESTING - ! - i
' i >
. i i O Testing Summary
j Product acceptance ! ,
Construction verifications
Precore officad channel check ! . . ! Component level (Phase I) testing l
, i i System level (Phase II) testng
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- .- .. . . .. . -. . . .. . _ - i
NUCLEAR INSTRUMENTATION SYSTEM TESTING
- l
. , !; G Product acceptance testing
r . Calibration checks and functional testing using pulse generator input
o Audio countrate t
o Shutdown monitor , o Internals vibration monitoroutput ~ c
- o Axial offset ' i . ' . Operational test of installed test equipmcat .!
! o Overpower trips ! ! o Permissives I ' o Containment evacuation alarm i i . Overpower setpoint checks through input to A and B coincidentors . ~ '
- Coincidentor logic testing (A and B Trains)
t . Contact logic checks i.
.- . .. -. - -.. -. - - . .- -. -.. -. - - - -. - - - -. -.
- ,r . _ _ ..
l - NUCLEAR INSTRUMENTATION SYSTEM TESTING
! l' i O Construction verification ' i Point-to-point wiring checks l e
Cable testing i i o Megger checks
o Pressure testing i o Ground isolation verifications l , Coincidence logic testing l !
O Training and procedure validation performed in shop using actual equipment prior to equipment installation i ! t . ! i < .- ! , :
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l.
, O Prior to core off load, output from existing Source Range channels 11 and 14 were compared to output from new Wide Range channel 4
- All three channels read approximately 100 counts per second
, l . This comparison provided added surety of proper response from f ' new equipment ' ! t O Component level (Phase I) testing f I
- Channel calibrations and coincidenier logic checks using
! Technical Specification surveillance procedures
! !
- Peripheral calibrations usmg surveillance procedures
- Overlap testing conducted using modified surveillance procedures
. . l ! l ! I i , ' . . . . .- ~.1 - m. - -4 - ~.- _ ~,.... -..~. --...,....__. .,._.. -.... - ..... ... - .
_ 4- . .
' i f
NUCLEAR INSTRUMENTATI_ON SYSTEM TESTING t i ! O System level (phase 11) functional testing I
- Reactor trip logic checkout will entail complete functional f
check of RPS (including NIS) up to and including reactor j scram breakers l 0 Testing will cover all coincidences
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. . . .
i NUCLEAR INSTRUMENTATION SYSTEM. TESTING i i . O System startup l '
,
- Based on historical data from the startup of Cycle 15,
' the old NIS read 10-20 counts per second (CPS) in the
vicinity of the secondary sources l l .
For this outage, core reload will begin in the vicinity of I .
new Wide Range channels 1 and 3
n I ! Initial assemblies will include a secondary source ' O 10-20 CPS expected on Wide Range channels I and 3 i Up to four fuel assemblies may be loaded - then expected
o indication must be present on two Wide Range channels l
I ! I
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_... , _.,, . . %._ , ,.,,, #, ,, - __.,._ _,... _ _ ___ . ,_ _ _ _ _ _ _, _ _ ___, ,, _,, __ _ _ _ __ _, _,,_., _ _ _ _, _ _ _ _, - , ,
_ -_--
._ _ - _ _ _ _ _ _ _ _ ._ _ - . - . NUCLEAR INSTRUMENTATION SYSTEM TESTING
O Reactor Startup -
- Fixed channel gain are initially set conservatively high
based on calculations and are adjusted after initial criticality i
- Adjustable channel gains are set to the maximum by procedure initially, then reduced as required to hold indicated reactor
' power below the first overpower reactor trip (17 percent)
- Expect to observe response on Power Range channels approximately four decades below the Point of Adding Heat on initial criticality
- Lower power physics testing amplifier has 100:1 gain for initial testing of one NI channel
. - .- . - . -. -. .- . . - - -..... . -.
_ - - . . _ _ _ -
. i.. l NUCLEAR INSTRUMENTATION SYSTEM TESTING
O Reactor Power Determination ! . !
- Fixed channel gain are initially set conservatively high
. based on calculations and are adjusted after initial criticality
i ! l o NI Power Range channci gains are adjusted to match calorimetric results Calorimetric considers primary and secondary inputs
, Primary side "zero point" established just prior
'
to reactor startup-O During NIS replacement project, detector locations relative ,
to core were not modified - O Effects of thermal shield removal were conservatively considered l
relative to the initial values of the Power Range channel fixed gains i O Wide Range channels have startup rate protection only and i do not contain overpower trips . ! ' , t _ , .. _. _ .. ...., . _. -... _ _... .. -. _ _ _ _ _.., .. .. . - ... . . ., _ _ _ _ _ _ _. _.
- - - --- ~ a " .; .
. . REVISED , . I .
l l TECHNICAT , i
-}
SPECIFICATIONS , I ! ! i J L ! ! ! ! i
1 i' Eric V. Fries . i > i , w t t ? o.
- , _. , ... _. - - -., -. -. -..,.,. _,, ,, _ -. _ _ _, _. - . _., _.... _ _. - .. -. - -.. -. . _... ~ _ - _ -,I
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. . _.
. . , . t REVISED TECHNICAL SPECIFICATIONS ! . O Connecticut-Yankee custom Technical Specifications have I ' been converted to Westinghouse standard format . i O Development, review and intemal approval of revised , i Specifications has been underway since 1986 .
l O Six submittals made between October 1988 and
August 1989 ' i ! ! t ! l O Supplemental submittals made in August and November 1989 i , ! . b ~ q . . _ _ _ - _ _, _ .- ~._-c_,_ . _; _. _,_ _- . . -...... . . ..
-- - _ -- _ - _. _ =- . . m.
!. REVISED TECHNICAL SPECIFICATIONS ! I , ! O Development of Revised Technical Specification (RTS) surveillance procedures i l'
- - Station procedums to implement RTS surveillance
., requirements have been developed i . i
RTS surveillance procedures have been developed i m parallel to existing station surveillance procedures i i ! i RTS procedures printed on pink paper and have
' special designations on cover sheet to assure
, distinction from existing procedures - l Separate files maintained for RTS surveillance
' procedures i ' .
! l l
' _- .. _. . ... -..... .. ..... .. _ -. ~.-.. . .. - - .
- - g . _ .. .
- . - , REVISED TECHNICAL SPECIFICATIONS
! O Adequacy review of RTS surveillance procedures RTS implementing surveillance procedures have been reviewed
- against RTS surveillance requirements to verify correct and complete content O Implementation of RTS surveillance procedures
A final adequacy review will be performed as part of the implementation process
i
Following incorporation of any required changes, RTS surveillance ! procedures will be issued, on pink paper, to implementing ' departments
From that point on, only RTS surveillance procedures will be utilized
RTS surveillance procedure impicmentation to occur within i 60 days of receipt of RTD . . .. - _ - - . - - - -- . . . - -
-....-..~...~....--_.......---~...-.--.....n... - - -., - _ - ~. - - - -. -.. - -.. ~... _ ~. - t ! I . Go l M
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~ - -,- - - - -,, <. - --y ...._._.. _ _. -........, -..... _. , _. _.... . - . . - -. - -
. . - .
FUEL RECONSTITUTION- , , O Preoutage Background . i ,
Approximately five fuci rods failed during previous
- l (Cycle 14) operation i-
, , l Cycle 15 primary coolant radiochemistry values
, ! were very similar to Cycle 14 , j . a Use of I-131/134 based fuel integrity monitoring program I
i i Monitoring methods predicted no greater than
! 12 failed fuel rods ! (
Preoutage preparation had been made to reconstitute I
i a maximum of 40 fuel rods in 20 assemblies using j stamless steel " dummy" rods as replacements ' .
. j All reconstitution to be manual - based on the
i j projected limited scope
i ! , , . -, , -.~~.a.,,.,,--# y $ 4.-- ..e -, -,-, - -, - y .c.
r -% .- w--,,y-ge ., - - - -r.
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. . FUEL RECONSTITUTION , O Post Shutdown Activities .
' j Increase in coolant activity during cooldown and
depressurization indicated a substantial number of failed fuel rods . '
All fuel assemblics were ultrasonically tested
. k Visual inspections revealed presence of debris between
the lower assembly nozzles and the first spacer grids .
Inspection, cleaning and reconstitution efforts were '
, upgraded to handle larger scope l ' . -! i i , __ .. .,. ~..,,.... , ., . _., .,,. . _ _,. .,,.., _. ,....., _, ... _ , _ , _ _ .-. -
_ - - - _ - - _ - _ _ - _ _ _ - _ _ _ - - _ _ _ _ _ - _ _ _ _ ._ _- __ . . L FUEL RECONSTITUION
O Fuel assembly debris removed using a specialized inspection , and cleaning station located in the spent fuel pool i
O Reconstitution approach Visual and ECT inspections were utilized
I All fuel rods located at debris sites were i
pulled and inspected
Fuel rods directly adjacent to debris sites also
- pulled and inspected , Additional random rod pulls undertaken
.a
Failed fuel rods and those rods with greater than 20 percent
through-wall defects wcre replaced with " donor" rods l ! I
i . -. - - - .. . . . .- . ... -.. . -,- _ - - - -
- . . .
FUEL RECONSTITUION !
O Reconstitution summary
3,333 fuel rods were pulled and inspected , i Greater than 500 fuel rods were discharged based on
ultrasonic, ECT and visual inspections 92 of 109 reload fuel assemblics were reconstituted
5 reload fuel assemblies had no damage o l o 12 reload fuel assemblies were used for " donor" rods o 6 of 12 " donor" assemblies were reconstituted for future reload (using " donor" pins from the other 6 " donor" assemblies) Inspection, cleaning and reconstitution took 5 months
-. -- .-. - - - -. - - - -. ..-...-...-- -.-. -..-.- -. - --
. 7 77 - .- - n- .- - m _- -
-- w ygg, - - .- . .. . *." ^' s - _M _ _ ,, ] _ . y _ FUELDRECONSTITFJION
I y 0-Core. Reload Strategy l . Revised reload design frozen in' November 1989. -
. ,. . L . 92 reconstituted assemblics to be reloaded e ' ' 48 new 4.0 w/o assemblies to be loaded (original
' contingent of new assemblics)
l ! Five reload assemblies did not require reconstitution l
a l Four 4.0 w/o and four 3.0 w/o new fuel assemblies were -! . l procured to replace " donor" assemblies. New assemblics
l . received April 25,.1990 l ! l Four thrice-burned assemblics.will be reloaded on the-L! e ' " flats" of the core
' Defects remaining in reload fuel assemblies evaluated to assure
operational acceptability l Xc-based fuel integrity monitoring program to be utilized to
i . detect and chcractize fuel rod failures- ' , . p . _.
_ _ _,... _.
._ _.. a _ _ _ ~. _. _ .. _ .., ... _. _ _ _ _, _ _ _ _ __ )
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. . . - . ,. FUEL INTEGRITY MONITORING PROGRA'M
i
O Traditional integrity monitoring program Based on analysis of coolant Iodine, r iums and Insolubles . es
Program improved over the years as data was collected and analyzed .
Good correlation established between observed activity levels and I e fuel pin failures
l
Cycle 15 operating data appeared consistent with expected pin failures l l Prediction based on Iodine and Cesium data .- - Abnormal items were present but explainable ! . Cause-effect of abnormal data not completely understood
o o Cycle 8 data was reviewed l o Industry experts censulted I t .
_- - _ _ _ - . ~,.. _. _ _ _ _, ._ .
_ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ - _ _ _ _ _ _ _ _ _ _ ___ - _ _ _ _ - . RADIONUCLIDE? LEVELS IN THE;RCS DURING CYCLE 15 OPERATION , imme Xc-133 - I-131 mum Cs-137 rem Expected - Reactor Xe-133 Thermal Power 1.(MIE + 03 j
- 1.(NIE + 02 - , 2 @o l N.
1.(NIE + 01 --* -- - -
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- L I-131 LEVELSTIN1RCS DURING l: SHUTDOWN AND PRIMARY-a , , i DEPRESSURIZATION ' 1.0E + 02 l A I-131 Level , ' em RCS Pressure .
^ , W l.0E + 01
E %; Db
$ $ }~*% ~ s <> i N n$$ Oh (N) '.\\ Jb 1.UE'O l '/ ^ . u t; ~i e % (*g\\ - '$ E< \\ .
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> - % ~~ =_ _=,f _ 3,3 j I I I I' I I I I I I ! ' ' ' "- 1.0E-02 8/278/288/298/308/31 9/1 9/2'9/39/4 9/5 9/69/79/8.9/99/109/119/12_ j DATE ,
, i ! & . -. - .- . .; . ... - -
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- . - - - "' _ " ' ',, ^ ,J-mw .. + - _ -- - - - - . . - -. L FUtLVINTEGRITYHMONITORINGi PROGRAM ! L . . O-Shutdown activity levels significantly exceeded those expected - , l-Iodine:
= Cesium ' ' , , Noble gas .
y O Existing plant equipment was utilized-with special precautions to degas ! primary plant and get' ready for refueling ' O Decision made to re-evaluate fuel integrity monitoring program j !
) i . i
y.
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, FUELDINTEGRITYLMONITORING PROGRAM . I O Action plan to enhance fuel monitoring'at Connecticut Yankee Critically evaluate the incident.
- o.
Inspect fuel to characterize quantity of pin failures ' ~ Understand the mechanism of fission product release o - . Scrutinize all available coolant radio chemistry data.
!
o What abnormalities existed j Quantify / explain abnormalities o Develop a new model for evaluation )
i Test model with historical data ~
. ; l Report to management and implement new monitoring program
'
l Further refm' e monitoring program.through on-going data evaluation - !
-. -... . ~.. _ _. _, _ . . _.. ,,, _ _.., _ , _. ~.,._,.,, ..,r, __ .c .- r., ... ~ ._m._,,,,, ,,,.~m., .,.m .. ~.. ~.,,,,. m.,._.~ ,,., -,.... ,
= . . - , y .- , - - 'F l.. ~ ~~
- 59 :
,_ , ._ .- FUEL INTEGRITY MONITORING.. PROGRAM.
- . O Development of the enhanced monitoring program Consult industry experts to assist in data interpertation
. Intensive self-evaluation of data to characterize those abnormalities
that did exist - 3y . o Xenon-133 levels l \\ o Correlate abnormal to normal data - ,
l Consider operational differences between Cycle 15 to past cycles o , l ' - seal leakage reduced .! - changes in lithium. treatment program - several downpowers but no trips or depressurizations [ !
t ,__-..a-.._ _..., _ _ ~ 2~,..._ -. . __., _ _. _ _. _ - . .. _ _.
.. _ _ -. _ -._.., _
. . _ . .. l ~ ' l.
. . rv '" _ . e w-CHARACTERIZATION OF FUEL DEFECT TYPES
i . . Open Defect Tight Defect --+-- A_,1 w . m w -
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wy Characteristics Characteristics i e lodine' Ratio Down e lodine Ratio Up H e Xenon Ratio Stable.
- .Xcnon Ratio Stable e Gas, Solubles and e Gas and Solubles in
? Insolubles in Primary Primary Coolant , , ~ Coolant e Insolubics in Fuel
. > + .- + -.. u-w~.-sm-gy l
- '*-,'
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'~ " _ -- - < . . ' ~~ ^ - - - _ i..: . t n . . FUELL INTEGRITYMMONITORING PROGRAML d ,
- -,
i
- ,
Postulate release mechanism for Cycle 15 . ,. . - .
l _.
.; . o 1Belljareffect . o --I-131/I-133 ratio.similarityi !
, Xe-133/short lived gas ratio j , o , ' . o-New type of fuel defect ". .
i D! "} . .. ; ~! .
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- :
n s a a w i e y GG P d - s Cl n a p O. d l d c l a e d C_ e e i t u n v v C s F a n i i l i e d a L L re n e C
u n t F E c n r i t g a s s o n i r a d a . h o le a GIoGS L - u h F C e e e e e - _ . - .. - - - - - .
. _ . _. _. . _ - _ - _ - . -. -_ _. - - _ - -. _ - - - _ - -. _ - - - - - - - _ _ ,
- - . - _ _
' ~ ~ '
- _
. FUELEINTEGRITY MONITORING PROGRAM
- .
'! i Evaluate release mechanism
i o No secondary hydriding of fuel cladding ' M! o Stainless steel relatively impervious to secondary hydriding j - 'F o Zircalloy may have secondary hydriding !
o Misinterpretation of radiochemistry data
.
, ! i
, - , .> ' i
_ _____,_,_I ,L;;.__,_..,_,.. ...;... ._ _, _. _,. _ ~. _. - _., ._....i, _ _. _. _.. _ _.. _., _. _, _ _ _ _ _ ,. _. _ _... _ ... .. ~ _. _ . _ _,
.. _ - ~
- .,
. T... + _ .. . . . . _,__ . . - - - % . . .~ . , "%c, ' .m .s '+ FUEL 1 INTEGRITY MONITORING PROGRAM , -
- Develop actual model o
Evaluate Xenon level and mechanism o Determine Xenon-133-level with 1-failed pin-site specific value o Determine the osmotic coefficient for Xenon-133 Evaluate accuracy of new model utilizing old ' data . 1.
, ! l l l . ....._..... . . ..... . _
. - - - - - - - - - - _ - - . y_ n - '"% e XENONLRATIOS.-
. r Cycle 15/ Cycle 8 . Cycle 13 Cycle 14 . Cycle 15-Avg (8,1.3,14) , Xe133/Xc135 3.3-3.1-2.5 5.0 1.7 Xe133/Xe138 6.3 7.1 6.0
10.1-Xe133/Xe133m
46
45 1.0 si . Xe133/Kr88 9.7 9.2-9.4 -
1.5 ! . Isotope Halflives: Xe133 .5.3 Days Xe133m 2.3 Days o Xe135.
9 Hours-
Xe138 17 Minutes
Kr88 - 2.8 Hours ,
j
. _ _ _. .- _ , _ _ , .. _ _ _ _ _. _ . _ . _ _ , _. _ _ _. . . -
. _ . - . . / . . s i MODEL PREDICTIONS VS.
' 1 ACTUAL FAILURES ,
-1-Based Xe-Based Actual Pin Failure Pin Failure Failed - , Defect Type Prediction Xe Ratio Prediction Pins . : l Cycle 8-Open Defects N 100 Pins
150 Pins . N 100 Pins *
,
Cycle 13 Tight Cracks 2 Pins-
3 Pins No inspections
Cycle 14 Tight Cracks 6 Pins.
9 Pins 6 Pins-l Cycle 15 Debris Induced 12 Pins
450 Pins 448 Pins-i Failures at ' Extreme Lower End of Fuel Pins-l ! '
- Cycle 8 - Fuci Sipping Performed Total Failed Pins Estimated l-
= -- - - - - .-- - . -- -- -
- .. = - . - , .~ : -
, - - + . . .
_ u, .n ' -. .,,_ FUEL ;INTEGRITYJ: MONITORING PRO. GRAM.
- ^ . I O Implementation of new fuel integrity monitoring program Develop procedures and establish appropriate action levels l
Secondary party expert review of' program
. Secure solid management commitment
On-going refinement of program based on operational experience ' ]
, I ' f
i f f ~:,- ..., -.,. .,, .,,, -.,,-J ,. - ,-l~l -..,. ,,-,-.l;s.,,,,- ~.,. - ,.. .~c.. ---,.O,, , ..,,,, , .s.-v.. ce ~n . ..w.
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.'4 ~. t FUEL INTEGRITYEMONITO_ RING PROGRAM L Q Conclusion I Connecticut Yankee has aggressively reviewed physical and
radiochemical evidence in an effort to accurately determine. j mechanism and establish an analytical model to explain observed ! events ' . .. t This model has been critically tested and determined accurate
,
A comprehensive revised monitoring program has been implemented t that will improve' predictive capabilities i Management strongly supports program and is prepared to implement. [
appropriate remedial actions i !
Further refinements to program' will occur based on in-service ! experience i l . - . - - .~ - .. . -
- - - - - - - - - - - - - - - - - - . -- -- = - - -. -. ~
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.- - . . - .. - . . ..
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.- . .--: .- - ~ _ _ , _ - - L l.
-PRIMARY SYSTEM CLEANUPL PROGRAM
l O The objective'of the program is to provide a high degree.- . of confidence that the potential for any future chip induced ' damage to the fuel has been reduced to the absolute minimum value , i
-. =.- ,, -. - - - _. - .,. -. - - -, ,i.-4., . .- - - - - -- -. -. - - -, - -t-~+1-1 r- - - - - - - - m
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PRIMARY 1SYSTEMLCLEANUP1 PROGRAM
> $ _ D Background i .. First indication of problem was high iodine . levels following shutdown
Material found in fuel during inspection
t Material source was' identified as 304 stainless ~
steel resulting from milling machine. operations on the thermal shield during the 1987 refueling - ! a , I . b I f i '
\\.- .. -: . a.
- . > - . . . . - - _ __;
_ . e +. . . - - - - . i:
PRIMARY l SYSTEM CLEANUP PROGRAM? O Background
The core, in-effect, filtered the primary system e' for 461 days of operation' Steam generator channel heads were inspected
and no debris was observed ' Reactor coolant system piping (section from
steam generator to coolant pump suction) in all loops a .was inspected - o Particles 2 to 3 mm in size were identified. I o Piping was essentially clean y Health Physics surveys of primary systems have not
~ identified any increases in Hot Spots ! o Surveys indicate fewer primary system Hot Spots .
. -.. - ... ., _m.c.,- ,
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-. - _ _.
- _ - , . . - - -- ' - _ , _, - -;. - - -- - . , . - ,. . ,
- .
. - . PRIMARY SYSTEM:'C.LEANUP? PROGRAM l
- l
. + -
. . . i No debris identified in system valves during . work on.the following systems during the
current outage.
! o Charging ! o Residual heat removal-i o Drain' header:
o - Fill header j o Pressurizer ! t
[
! i
l ' - - . . . ... _.
... - . = . . .
_ _ _ _ _ _ _ _ - - _ _--_-----__ _-__ __ __ ._ . _ _
- - - - . .. _ . . PRIMARYESYSTEM? CLEANUP PROGRAM . , ly i O Program consists of an outlying system flush (Phase 1) combined .i with an RCS inspection and debris retrieval program (Phase-2) ] .
This entails, following outlying system flushes:'
' RCS piping inspections utilizing a remote controlled
submersible vehicle carrying a high resolution video.
ll camera and a manipulator arm with grabber and suction hose j ' Reactor vessel, upper internals, lower internals and I pressurizer inspections utilizing underwater video camera
. ' system and high suction vacuum system ! Final video inspeption ' Safety assessment based on final conditions i
i ! _.
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. . r o . - - - - . PRIMARY! SYSTEM CLEANU.P PROGRAM
- SCHEMATIC
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l . .. c , g W'. - $p#ggpQy3-{f ? T1 lia} l . . . . l Outlying System Flush (Phase 1).
RCS Cleanup (Phase 2) ! Includes: RHR,' CVCS, HPSI -LPSI
. , i i > I' , ? h
1 l ... - ~ + -w,-.-.~y , .-,,,., - ... _,., ...- _, _., ,.,.,, ., y..... r .. y _,., ,,_.,. 1,,,.,
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- ~
PRIMARYJSYSTEML CLEANUPLPROGRAM.
_ g Advantages of the primary system cleanup program -- Ability to ' establish cleaness verification
- - Lower exposure
- - Lower cost.
- Least schedule; impact Eliminates operational concerns
i - ~=L --= gy _ _ _ _ _ _. .- _. -. .
- - - -. -
- - .
. . .. - . . . -- .. . .m , "*'""'c-p^ - - c -. _ _ , - - - . .. _ .,e - m _ - P RY1 SYSTEM 1CLEANUPLPROGRAM ~
u
. O Status. and Results (as of May17,1990)L .
ApproximatelyL 90 percent of the RCS.
outlying system flush program has been completed ' .
- .No machining debris has been found o
Flush strainers have been essentially l o clean of all debris >
. .. . . _
Pressurizer inspections found " dirt type" debris j
- No machining debris has.been verified
' o Further mspections of filter debris will-provide more detailed information , i
_.. _. _ _ -.. ~__ .,,.-.._,_;.1_.. .;,....., 1.__.__.u__L_...,._ __. -..... _., -......,...., . _ -......,.. _. .....
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- .-
~ ' ~ PRIMARY SYSTEMLCLEANUP PROGRAM- ~ _ ,-.:
Conclusions
Based on the results of the cleanup program to date, ~ it appears the hypothesis that the vast majority;of ' . the machining debris was collected and retained by the fuel is correct and that machining debris does
not remain disposed throughout the RCS and l . associated systems -! ,
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"' . STARTUP LICENSING ISSUES O Revised Technical Specifications Reload, ECCS, RPS/NIS, Fire Protection, misc.
- Recognize the possible need for emergency changes
O Technical Specificaron changes Stainless steel r ds in fuel . ECCS check valve surveillance testing e New fire protection features
Service water header clarification e ' Material surveillance specimens - 10CFR50 Appemlix H . i _ - _ --- ___
- - - - - - - - - - - - - -. -- - - - _ . r i -. I ' STARTUP LICENSING ISSUES O License Condition Fuel monitoring program
j O Other submittals Revised LOCA analysis / reload report
EMI testing plan
$ l Spent fuel pool cooling analysis
, ISAP/IIS update "
i I ! l I '
$ . s .- , . . . . -. , . . . .. .
_.. - - - - - - - - - - - - - - - -_ _ - - _ - _ -. .- . .- - ._ , + .
-APPENDIX R LICENSING ISSUES i l O New Switchgear Building j Electrical design approval
Civil design approval
- Fire protection approval
O Post outage submittals Revised compliance report
Revised Fire Hazards-Analysis (FHA) i
! O Post outage approvals Revised Appendix R exemptions
Revised Appendix R SER
i l l l i L...... __ .. _ - _ ~..
._ _ _ . . SUBMITTALS REQUIRED AFTER STARTUP , O Final report on thermal overload protection , O Bulletin 88-08 (thermal stresses in piping connected to RCS) ! O Generic Letter 88-14 (instrument air supply system problems) f O ILRT report j
Startup physics testing ! ) I-i L , . D . ! _.
--_ _- __ - .. .. -. - - - - - - l
(. }l ' A Attechnent 4
. ?. , i: L OVERVIEW
NORTHEAST UTILITIES FITNESS FOR DUTY PROGRAM j INTRODUCTION: 1.
Our research shows a need for drug and alcohol testing, i.e. A major Connecticut based defense contractor had a 17% positive drug testing rate for " pre hires" over a two year period (mid 1987 to mid ! 1989).
l' 2.
. State of Connecticut politics and court system are liberal in regards to ! drug and alcohol testing.
, A.
Connecticut state law (1987)d for cause basis, allows only for drug testing (NI panel of 5) on a pre hire an it does not allow for random or post accident drug or alcohol testing.
It allows for
alcohol testing only by state or local police for DWI.
B.
NU's legal department was heavily involved in the FFD program development.
PROGRAM COVERAGE: 1.- Those with unescorted access to Millstone or CY - A.
Millstone - Unit
Unit - 2 l Unit - 3 ! B.
Haddam Neck - Connecticut Yankee , I C.
Corporate - Berlin Rocky Hill InterMaintenanceForce(IMF) D.
Random Pool (5): Millstone
1.
Employees (715) 2.
Contractor (833) Connecticut Yankee 1.
Employee (195) 2.
Contractor (504) r % - .,. .- . .-. -
i
o2- . . . Corporate Pool 1.
Includes: (1137) a.
Employees badged at both CY and MP b.
Inter Maintenance Force (IMF) c.
Employees with badges who report to Berlin E.
Testing facilities: Millstone 1.
Medical facility 2.
Badge Processing Center CY 1.
Medical Facility Corporate Building 1.
Medical Facility PROCEDURES: ' 1.
Fitness for Duty Manual A.
28 procedures needed to administer the program.
B.
Manual approval process signed of by: Senior Vice President Nuclear Engineering and Operations and Senior Vice President Administrative Services.
DRUG TESTING: 1.
10 Drug Panels Initial Substance: Test Limit Test limit Alcohol 0.04% BAC 0.04% BAC Marijuana 50 ng/ml 15 ng/ml Cocaine 300 ng/mi 150 ng/ml Opiate 300 ng/ml 300 ng/ml Phencyclidine 25 ng/ml 25 ng/ml Amphetamines 1000 ng/mi 500 ng/ml ^ Barbituatas 300 ng/ml 300 ng/mi Benzodiazepines 300 ng/ml 300 ng/ml Methadone 300 ng/ml 300 ng/ml Methaqualone 300 ng/ml 300 ng/mi Propoxyphine 300 ng/ml 300 ng/ml 2.
THC cutoff leve'is: 50 mg/ml . - . . . .
h
, w 3.
Split sample: To insure program integrity in the eyes of the employees.
' ALCOHOL TESTING: 1.
Intoximeter 3000: The only breathalyzer certified for use by the State of Connecticut and upheld in the Connecticut state court system.
SANCTIONS: (DrugandAlcohol) 1.
Random, first time positive a.
NU employees: minimumof(t) weeks b.
Contract employees: minimum of 365 days 2. - Random, second time positive a.
NU employees termination b.
Contract employees: termination 3.
Preaccess positive, for NU and contract employees: not allowed to apply for a minimum of 365 days.
4.
Possession, use or sale while in the PA, for NU and contract employees: termination.
TRAINING 1.
Initial: A.
All badged personnel (5006) l Initial FFD Awareness / Escort Training (2 hrs) B.
Supervisors (425) l Observation Awareness Training (OAT) (2 days) l 2.
Annual Requal:
l A.
All badged personnel FFD Awareness / Escort Refresher Training l 1.
Employee - included in GET (30 min.)
l 2.
Contractors - included in SS&E (30 min.)
' B.
Supervisors: i Observation Awareness Refresher Training (4 hrs',
[ ! -4 '- o .
3.
New Employees A.
All badged personnel Initial FFD Awareness / Escort Training , 1.
Employees included in NET (30 min.)
2.
Contractors included in SS&E (30 min.)
8.
Supervisors Observation Awareness Training (OAT)2 days).
Offered every quarter.
DRUG AND ALCOHOL TEST RESULTS: 1.
Started testing December 1,1989 at 50% rate / January 3,1990 at 100% rate.
2.
Drug and Alcohol Test Results Dec. 1, 1989 - April 30, 1990 Type Test Tot. # Tested
- Positive Percent Random:
NV-799
0.38% Contractor 517
0.39% Preaccess: NV
1 1.18% Contractor 553 15* 2.71% for cause: NU
0
Contractor
4 .)
- 1 Alcohol positive.
PROBLEM AREAS: L l 1.
Refusal to sign forms (1) 2.
Refusal to empty pockets (1) l 3.
Test results turn around the , 4.
Convincing those who have been randomly selected (2) or (3) times that I the random generator is in fact random.
! Number testing (1) time 894 Number testing (2) times 164 Number testing (3) times
' Number testing (4) times
i i
_.
.. _ . _ -.._ . . 5- . .. , i...: i
5.
Poppy seed issue (5 poppy seed positives) 6.
Testing during outages . a.
We badge process up to 100 contractors a day /we can only test 40 people a day.
b.
Options we are considering: l
1.
Hiring'a contractor to being in a mobile unit to support our
testing.
2.
To purchase a mobile testing unit we could drive from site to site.
! FAVORABLE IMPLICATIONS: 1.
Testing results show that the FFD program does act as a deterent.
2.
Word is out at the union halls, "If you have a drug or alcohol problem, don't bother going to CY or Millstone...go work on a building or build a
bridge."
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