IR 05000255/1993024
| ML18059A561 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/06/1993 |
| From: | Kobetz T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18059A560 | List: |
| References | |
| 50-255-93-24, NUDOCS 9312130044 | |
| Download: ML18059A561 (52) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION REGION I I I Report No. 50-255/93024(DRP)
Docket No. 50-255 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name:
Palisades Nuclear Generating Plant Inspection At:
Palisades Site, Covert, Michigan License No. DPR-20 Inspection Conducted:
September 30 through November 15, 1993 Inspectors:
Approved Inspection Summary 2A Inspection from September 30, through November 15. 1993 (Report No. 50-255/93024{DRP))
Areas Inspected:
Routine, unannounced inspection by the resident and regional inspectors of actions on previously identified items, licensee event report followup, followup of events, operational safety verification, radiological controls, maintenance, surveillance, engineering and t~chnical support, and management meeting No Safe~y.lssues Management System (SIMS) items were reviewe *
- Results:
No cited violations or deviations were identified in any of the nine areas inspecte One noncited violation was identified and is described in paragraph The*strengths, weaknesses, and Inspection Followup Items are discussed in paragraph 1, "Management Interview."
-***--- DETAILS Management Interview (71707)
The inspectors met with licensee representatives (denoted in paragraph 11) on November 16, 1993, and informally throughout the inspection period to summarize the scope and findings of the inspection activitie The inspectors also discussed the likely informational content of the inspection report including the attachments, with regard to documents or processes reviewed by the inspector The licensee did not identify any such documents or processes as proprietar Highlights of the exit interview are discussed below: Strengths noted:
(1)
Performance of startup testing without incident. This was noted to be a significant improvement from the last refueling outag (2)
Operator handling of a rapid power reduction that prevented a reactor tri Weaknesses noted:
(1)
Torque values for main turbine hydraulic fluid hose fittings found not within specified value (2)
Procedure performance error during a safety injection system surveillance tes (3)
Housekeeping and material condition in various radiologically controlled areas of the auxiliary buildin.
Actions on Previously Identified Items (92701, 92702) *
(Closed) Inspection Followup Item 50-255/93010-0l(DRS):
The licensee's technical specification required testing of the main steam isolation valves did not accurately reflect their ability to perform their safety function. The main steam isolation valves were tested on power ascensions after they had already been exercised, repaired, or conditione Therefore, they did not reflect the as-found conditio The licensee revised General Operating Procedure 9 (GOP-9), "Plant Cooldown from Hot Standby/Shutdown, 11 Rev.13, to specify that the subject testing be performed when the main steam isolation valves are closed during a plant cooldown, which would reflect as-foun conditions and accurately demonstrate the main steam isolation valves' ability to perform their safety function. This item is close * *
(Closed) Notice Of Deviation 50-255/93010-02(DRS):
For a steam line break inside containment concurrent with a failure of the main steam isolation valve to close on the unaffected steam generator event, the licensee failed to.meet their commitment documented in a response dated April 28, 1986, with regard to the following:
Go to once-through cooling (OTC).
- Ensure maximum feed flow to at least one steam generato *
Maximize containment spray flow and place containment coolers in emergency alignmen *
Maximize service water and component cooling water flo The licensee identified where all of the above items would be accomplished in the existing emergency operating procedures and conducted a review of emergency operating procedure strategy with respect to initiation of OTC. The existing strategy allows automatic initiation of feedwater, ensured by manual actions, with acceptable cooling verified by steam generator level and primary coolant system parameters. If continued use of the steam generators for decay heat removal is not possible, OTC would be initiate The current emergency operating procedure strategy was compared wjth an alternative strategy of immediately initiating OTC upon observfng the symptoms of a steam line break concurrent with a failure of the main steam isolation valve to fully close on the unaffected steam generator. The licensee concluded that either method would result in continued core coolin The licensee maintains that the current emergency operating procedure strategy of using OTC only if cooling using steam generators cannot be verified is preferred to immediate initiation of OTC. The current strategy reduces the risk of losing the ability to cool the core, does not further compound an already complicated event, and conforms to the approved guidance for Combustion Engineering (CE) plant emergency operating procedure Based on the inspectors review of the licensee's response this item is close (Closed) Inspection Followup Item 50-255/93010-03(0RS):
No operator training has been provided on a steam line break inside containment concurrent with failure of the main steam isolation valve to close on the unaffected steam generator event. This event would result in simultaneous blowdown of both steam generators into containment. Additionally, no training has been provided on determining operability of instrumentation which may be adversely affected by this environment.
- The licensee has committeq to complete classroom training prior to the end of 1993 for this event. The training for the subject. even is to include discussions on the following:
How a blowdown of both steam generators could occu *
An explanation of why there are differences between safety analyses and simulator modeling of some event *
Symptoms, expected plant response, emergency operating procedure paths involved, and the potential for significant error or failure of instrumentation located in the containmen *
Verification of instrument reading validity and use of alternate instrumentation for this and other events which degrade the containment environmen Simulator training will also be provided once the necessary simulator modeling changes are completed (Open item 50-255/93010-05(DRS)) and the emergency operating procedures associated with this event are validated. Based on the licensee's commitment to provide the described training, this item is close (Closed) Inspection Followup Item 50-255/93010-04{DRS):. A caution in Emergency Operating Procedure 9.0 (EOP-9), "Functional Recovery Procedure," Rev.3, provided no useful information. The caution statement at step 12 stated the shift supervisor may deviate from the procedure via 10 CFR 50.54X. This caution is unnecessary and inappropriat Emergency Operating Procedure 6.0 (EOP-6), "Excess Steam Demand Event," Rev.4, Attachment 2, contained graphs to account for errors in pressurizer and steam generator narrow range leve However, the attachment was not referenced in the body of the procedur The licensee committed to delete the unnecessary caution from EOP 9.0 during the current emergency operating procedure revision effort. Additionally, the licensee identified that EOP 6.0, Attachment 2, is referred to in several places within the EOPs, such as step 6.a.l of Attachment 1 to EOP 6.0, and considers no change necessary regarding this item. Based on a review of the licensee's response this item is close (Closed) Inspection Followup Item 50-255/93010-05(DRS):_ The simulator modeling of a main steam line break inside containment concurrent with a failure of the main steam isolation valve to close on the unaffected steam generator event was not accurate.
- Simulator modeling of containment temperature and pressure indicated lower values than those expected for the event based on previous analyse The licensee is conducting a comparison between the simulator, CPMAAP (an engineering analysis computer code), and safety analysis calculations for containment response: to the event to determine correct simulator modeling. The corrections necessary for proper simulator modeling are scheduled to be completed prior to the end of 1993. This item is close No violations, deviations, unresolved, or inspection followup items were identified in this are.
Licensee Event Report Followup (92700, 92720)
The inspectors reviewed the following Licensee Event Report (LER) by means of direct observation, discussions with licensee personnel, and review of record The review addressed compliance to reporting requirements and, as applicable, that immediate corrective action and appropriate action to prevent recurrence had been accomplishe (Closed) LER 255/92009:
Inadvertent Actuation of the Control Room Heating, Ventilation, and Air Conditioning (HVAC) due to Damaged Electrical Cable:
On February 13, 1992, the control room ventilation system inadvertently switched to the emergency mod At the time of occurrence, the plant was in cold shutdown and the primary coolant system was de-pressurize The control room operators immediately verified that a valid containment high pressure (CHP) or a containment high radiation (CHR) signal did not exis The licensee subsequently determined that electrical maintenante personnel were replacing damaged flex conduit on the CHR relay 5R-6 circui When the wire on relay 5R-6, point 16, was disconnected, the 11 A" train of control room HVAC automatically switched to the emergency mod This event was caused by inadequate job planning and personnel erro The job plan was inadequate in that the work instructions did not identify the electrical scheme as a "de-energize to actuate" schem Personnel erred by not referring to the approved drawings prior to lifting wire The drawings clearly identify the "de-energize to actuate" schem The licensee's completed corrective actions were appropriat They include reiterating management's expectations through training that the electrical and l&C staff review electrical schematics against job plans prior to initiating wor Proper job planning requirements were also reviewed with the maintenance planners. This LER is closed.
"
- (Closed) LER 255/92015:
Noncomplvinq Movement of Heavy Loads due to Procedure Error:
On August 24, 1989, the licensee discovered that a commitment for both the reactor engineer and the shift supervisor tfr approve heavy load path deviations was inadve~tently removed from their heavy load procedures FHS-M-23, "Movement of Heavy Loads in the Spent Fuel Pool Area," and FHS-M-24, "Movement of Heavy Loads in the Containment Building Area."
A search commenced to determine if any deviations from safe load paths, approved by only the shift supervisor, had occurre Such a deviation from approved load paths had occurred on September 30, 199 The event was caused by the inappropriate procedure revision and weaknesses in the commitment tracking system used in 198 Since then the reactor engineer approval was re-instituted back into the heavy load procedures and the commitment tracking system was improve This LER is close No violations, deviations, unresolved, or inspection followup items were identified in this are.
Followup of Events (93702)
During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR 50.7 The inspectors pursued the events onsite with licensee and/or other NRC official In each case, the inspectors verified that the notification was correct and timely, if appropriate; that activities were conducted within regulatory requirements, and that corrective actions would prevent future recurrenc The specific events are as fo 11 ows:
October 9, 1993 -
October 12, 1993 -
November 5, 1993 -
November 6, 1993 -
November 12, 1993-Unisolable through wall leak identified on Pressurizer Temperature Element TE-010 Unisolable through wall leak iden~ified on Pressurizer Temperature Element TE-010 Hydraulic fluid leak found on main turbine number 2 stop valve (CV-0571)
Erratic operation of main turbine number 1 governor valve (CV-0570)
Hydraulic fluid leak found on main turbine number 2 intercept valve (CV-0548)
The following are brief summaries of the event The inspectors will evaluate corrective actions for the events when the respective LERs are reviewed.
- On October 9, 1993, during a reactor coolant system walkdown the licensee identified water leaking from a pressurizer head penetration around the base of temperature element TE-010 This
. element is used to determi.n.e the*. vapor.:_. phase. temperature,of theG pressurize At the time the leak was identified, the reactor plant was in cold shutdown with the plant pressurized.to 250 psi The leak rate was estimated to be about one ounce per minut Shortly after this primary coolant system (PCS) leak was identified, actions commenced to depressurize the syste This pressurizer resistance temperature detector (RTD) well was previously scheduled for a visual inspection during plant startu The RTD well had not been worked on during the refueling outage; however, moisture had been observed around this penetration during the power operated relief valve (PORV) line repairs some three weeks earlier (~ee NRC Inspection Report 255/9302l(DRP)).
At the time the PORV leak was found the licensee was uncertain if the moisture seen at TE-0101 was due to extensive wetting of the pressurizer head area and surrounding insulation due to the PORV leak, or due to an actual leak at the RTD nozzl Shortly thereafter, the licensee performed a pressure drop test of the RTD thermowel Although this test did not identify any concerns, the licensee was unable to perform further evaluation until the PCS was pressurize On October 12, 1993, a similar leak to that described above, was found during a followup walkdown of the pressurizer to inspect additional nozzle penetrations.. The. licensee identified moisture and boric acid corrosion around pressurizer liquid phase temperature element TE-010 This walkdown was performed with the PCS depressurized and only a static head of water in the pr~ssurizer. The leak was estimated at several drops per minut Based upon a review of industry experience and NRC Information Notice 90-10, "Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600, 11 the licensee speculated that the leakage came from between the inconel sleeve and the carbon steel shell of the pressurize On October 12, 1993, daily conference calls were established between the licensee and the NR In addition, on October lZ and 21, 1993, technical meetings were held with the licensee to discuss PWSCC on the pressurize The licensee obtained assistance from the Combustion Engineering (CE) owners group to evaluate this proble Region III and NRR specialists followed the results of additional PCS inspections in addition to the PORV line repair The licensee's repair efforts of both TE-0101 and TE-0102 were satisfactor (See NRC inspection report 50-255/93023(DRS) for further details.).
The licensee continued with startup preparations following
successful repairs to the leaking pressurizer PORV welds and pressurizer temperature elements... No additional leaks were observed during those repair efforts and during containment closeout tours.. On November 12, 1993, with the plant at 100 percent power, the plant experienced a transient when a steam supply valve to the "A" low pressure turbine was closed due to a hydraulic fluid lea Operators performed a rapid down power maneuver on the primary side to match the secondary sid The leak occurred on CV-0548, the "A" low pressure turbine reheat intercept valv Upon discovering the leak, the operators isolated the fluid to the valve causing it to go close The resultant pressure spike in moisture separator reheater E-9A caused lifting of its relief valve The operators then isolated the steam supply to moisture separator reheater E-9 This action placed the main turbine steam system back into a configuration covered by plant procedures, namely Standard Operating Procedure 8 (SOP-8), "Main Turbine and Generating Systems," Rev.2 Operators handled the transient well with good coordination between the primary side and secondary side operator The quick reaction by the plant operators and the action of the automatic controllers in the steam generator water level control system likely prevented a reactor trip. The coordination between the operators in the control room is considered a strengt The hydraulic fluid leak was repaired and the unit was returned to full power on November 14, 199 No violations, deviations, unresolved, or inspection followup items were identified in this are.
Operational Safety Verification (71707, 71710, 42700)
Routine facility operating activities were observed as conducted in the plant and from the main control roo Plant startup, steady power operation, plant shutdown, and system lineup and operation were observed as applicabl The performance of reactor operators and senior reactor operators, shift engineers, and auxiliary equipment operators was observed and evaluate Included in the review were procedure use and adherence, records and logs, communications, shift/duty turnover, and the degree of professionalism of control room activitie Evaluation, corrective action, and response for off normal conditions were examine This included compliance to any reporting requirement Observations of the control room monitors, indicators, and recorders were made to verify the operability of emergency systems, radiation
- monitoring systems, and nuclear reactor protection system Reviews of surveillance, equipment condition, and tagout logs were conducte Proper return to service of selected components was verifie Periodic verification of Engineered Safety Features status was conducted by the inspector Equipment alignment was verified against plant procedures and drawings and detailed walkdowns selectively verified:
equipment labeling, the absence of leaks, housekeeping, calibration dates, operability of support systems, breaker and switch alignment, as appropriat General Plant operators brought the plant back on line on November 8; 1993, ending a 156 day refueling outag The plant began the inspection period in cold shutdown with the primary coolant system partially drained with preparations for leaving cold shutdown in progres Some relevant dates are:
October 27, 1993 -
Reactor left cold shutdow October 28, 1993 -
Reactor in hot shutdow November 3, 1993 -
Reactor critical at 10-4 percent powe November 4, 1993 -
Reactor entered power operations at greater than 2 percent powe November 8, 1993 -
Turbine Generator on line following successful overspeed testing The inspectors followed various primary coolant system parameters during and after plant startup. Primary coolant system leakage was very low, with no observable trend in containment sump level Primary coolant activity values were normal, at or below measured values seen following previous refueling outage Dose equivalent iodine averaged less than two percent of the technical specification limit. Secondary side equipment operated well, except for some hydraulic fluid leaks associated with the turbine generator steam supply valves, as previously mentione The inspectors provided expanded site coverage beginning immediately prior to criticality and extending through the start of power escalation. Assistance from the D. C. Cook resident inspector office was obtained. Coverage of all three shifts were provided daily. Major activities observed were:
Approach to criticality
Reactor criticality
- *
Turbine Generator Overspeed trip testing
Synchronization of the turbine generator to the grid for power ascension Criticality The unit went critical at 11:09 a.m.(EST) on November 3, 199 This started the low power physics testing portion of the startup progra The estimated critical rod height and boron concentration were within the predicted target ban Plant Tours Tours of the control room were routinely mad Staffing requirements were met, operators were cognizant of changing plant conditions, the equipment status and the Limiting Condition for Operation status boards were maintained up to dat Portions of the following startup activities were observed:
(1)
GOP-3, "Hot Shutdown to Critical in Hot Standby," Rev.12 (2)
T-191, "Startup Physics Test Program," Re (3)
SOP-8, "Main Turbine and Generating Sys.terns," Rev.27 No violations, deviations, unresolved, or inspection followu~ items wer identified in this are Radiological Controls (71707)
During routine tours of radiologically controlled plant facilities or areas, the inspector observed occupational radiation safety practices by the radiation protection staff and other worker Effluent releases were routinely checked, including examination of on-1 ine recorder traces and proper operation of automatic monitoring equipmen Independent surveys were performed in various radiologically controlled area On one tour the inspector observed the entrance to the East Engineering Safeguards Room was a posted high radiation area; this area is typically only a radiation are Investigation found the reason was a 180 R/hr hot spot lodged in place near the suction of the low pressure safety injection/shutdown cooling pump P-67 The hot spot was appropriately shielde A gamma scan indicated the hot spot to be a fuel particle, most likely due to the failed fuel found earlier this outage and carried through the shutdown cooling system while that system was operating.
The licensee's attempt to capture and further characterize the particle proved unsuccessfu The original* intention was to jo P-67A while tracking the hot particle from the suction piping into the pump casin Using a hot spot flush-rig they.would flush the particle through a drain on the casing and capture it in the rig for further analysi Instead, the licensee attempted to flush the hot spot with only the head of water from the hot spot flush rig, without jogging P-67A, due to other ongoing testin The hot spot did not move during this attemp Later during hot shutdown surveillance testing of this system, the particle became dislodged and was flushed out of the pipin The licensee checked all accessible piping within the system and concluded the particle is probably residing in the safety injection and refueling water tan There has been no indication of abnormal or excessive radiation doses received by any individuals nor have any anomalous trends been note The licensee has an acceptable monitoring program in place to locate and shield the particle should it reappear in the syste Housekeeping and material condition was noted to have deteriorated in various auxiliary building areas as observed during an NRC management tour just prior to the end of the current refueling outag Of note was the excessive amount of contaminated area in the east and west safeguards pump room Additionally, throughout the plant an increased use of catchments to direct or contain leaks was observe Some areas had work request tags to repair valve and pump leaks and improve material condition that were several months to several years ol This item is of concern since there is an increased potential for personnel contamination events and unnecessary exposur These observations were discussed with the licensee, who has commenced aggressive action to reclaim and clean up contaminated area The inspectors will continue to monitor the licensee's progress in this are No violations, deviations, unresolved, or inspection followup items were identified in this are.
Maintenance (62703, 42700)
Maintenance activities in the plant were routinely inspected, including both corrective maintenance (repairs) and preventive maintenanc Mechanical, electrical, and instrument and control group maintenanc activities were included, as availabl The focus of the inspection was to assure the maintenance activities reviewed were conducted in accordance with approved procedures,
regulatory guides and industry codes or standards, and in conformance with Technical Specifications. The following items were considered during this review: the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; and post maintenance testing was performed as applicabl The following maintenance activities were observed:
(1)
Repairs to the pressurizer PORV weld and temperature element nozzles (2)
Troubleshooting of an inoperable subcooled monitoring channel (3)
Containment Air Cooler VHX-2 service water leak repair (4)
Troubleshooting of the hydraulic fluid leak found on main turbine number 2 stop valve (CV-0571)
(5)
Troubleshooting of the erratic operation of main turbine number 1 governor valve (CV-0570)
(6)
Troubleshooting of the hydraulic fluid leak found on main turbine number 2 intercept valve (CV-0548)
No violations, deviations, unresolved, or inspection followup items were identified in this are.
Surveillance (61726, 42700)
The inspector reviewed technical specifications required surveillance testing as described below, and verified that testing was performed in accordance with adequate procedures. Additionally, test instrumentation~
was calibrated, limiting conditions for operation were met, removal and restoration of the affected components were properly accomplished, and test results conformed with technical specifications and procedure requirement The results were reviewed by personnel other than the individual directing the test, and deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The following surveillance tests were observed: Rl-47, "Rod Withdrawal Prohibit Interlock Matrix Check, 11 Re MI-SA, "Containment High Pressure Test, 11 Re Q0-1, "Safety Injection System," Rev.34 The licensee issued a deficiency report after testing the left channel of safety injection when the operators performing the test
- missed a procedure step to place the control switch for certain valves in their required positions. Step 5.2.5.b required the operator to position the control switch for each of eight valves to the "as-left" position.. However, the operators only visually verified the valve positions without manipulating the control switche Step 5.2.4 of the procedure requires the operator to push and hold the safety injection actuation signal (SIAS) push button until step 5. Releasing the SIAS button in step 5.2.7 removes the SIAS test signa When step 5.2.5.a was performed with the SIAS button pressed, the operators properly verified that the eight valves, two of which were CV-0913 and CV-0950, changed to their required test position CV-0913 and CV-0950 supply seal cooling water to the high and low pressure safety injection pump Both valves properly went to the "open" test positio The following step 5.2.5.b required the operator to position the control switches for CV-0913 and CV-0950 to the "as-left" (in this case the "open") positio However, the operators only visually verified the valves were open without turning the control switch to the "open" positio The intent of using the control switch was to leave a standing open signal to CV-0913 and CV-0950, so that after the SIAS button was released the valves would not reclose with the high and low pressure safety injection pumps still runnin As a result, the left train of high and low pressure safety injection pumps (P-668 and P-678, respectively) ran for about fifteen minutes without cooling flow to the seals and bearings, until an operator discovered the conditio Technical Specification 6.8.1.a requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2 (February 1978), Quality Assurance Program Requirements, as endorsed by CPC-2A, Quality Program Descriptio The Quality Program Description in CPC-2A endorses Emergency Core Cooling System Tests in Appendix A of Regulatory Guide 1.33, section 8.b. (j).
The inspector considers the failure to implement procedure step 5.2.5.b to be a violation of the above requiremen The cause of the failure was personnel error, although the instructions at this section of the procedure could be cleare The safety significance of running P-668 without seal and bearing cooling flow for this short time was minima Component cooling water flows in parallel to cool the pump bearings, stuffing boxes, and the seal flushing water coole Final Safety Analysis Report
(FSAR) section 6.1.2.2.3 states that the seals are designed for 300°F and are provided with cooling to.extend seal life. Since the seal cooling was 60°F water from the Safety Injection and Refueling Water (SIRW) Tank, little if any seal degradation occurred. Additionally, correspondence from the pump* manufacturer stated that bearing cooling is not needed for the high pressure safety injection pumps below 250° A similar argument can be made for P-67 FSAR section 6.1.2. states that the seals are designed for 325°F and since the seal cooling was 60°F water from the SIRW tank, little if any seal degradation occurred. Also, no significant temperature trends developed during a 1989 test to track performance of the seals, bearings, and stuffing boxes without cooling water flow during a 45 minute perio Although the operators made a procedure error, there were several positive observation The prejob briefing went extremely wel The operator in charge of the test read through the entire procedure with the rest of the crew and expectations were laid ou The system engineer was involved at the onset of the prejob brief until the test was completed and provided some good comment There was good discussion on contingency actions should problems arise during the test, such as an unforeseen loss of noncritical service wate Additionally, there was good involvement by the Nuclear Plant Assessment Department (NPAD) observer present for the tes Upon discovering the loss of seal cooling to the pumps the NPA observer suggested that someone be sent to the safeguards rooms to check on the condition of the pump Therefore, the violation will not be cited since the licens~e discussed the problem and quickly performed appropriate recovery actions and since the criteria specified in Section VII.B.2 of the
"General Statement of Policy and Procedures for NRC Enforcement Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C), were satisfie No violations, deviations, unresolved, or inspection followup items were identified in this are.
Engineering and Technical Support (37700,92705)
The inspector monitored engineering and technical support activities at the site and, on occasion, as provided to the site from the corporate offic The purpose was to assess the adequacy of these functions i contributing properly to other functions such as operations, maintenance, testing, training, fire protection, and configuration managemen *
-- Various portions of the startup testing program were observe Hot shutdown testing, low power physics testing,. and power escalation proceeded relatively smoothly as a result of being conducted in a well coordinated fashion.- Unlike similar testing performed at the conclusion of the previous refueling outage~ the startup program was performed without any unplanned engineered safety feature actuations. This is a significant improvement from the last refueling outage and is considered a strengt The licensee's investigation into the hydraulic fluid leaks on the main turbine steam supply valves was followe The causes of the two hydraulic fluid leaks were totally separat The cause of the November 5, 1993, leak on main turbine stop va 1 ve CV-0571 was that one of the hydraulic fluid hose fittings had been insufficiently torque The cause of the November 15, 1993, hydraulic fluid leak on main turbine intercept valve CV-0548 was due to a failed a-ring in a test solenoid valve that supplies the hydraulic flui The licensee has inspected the other main turbine steam valves and stated that no other fittings were found to be under-torqued, but some were found with a higher than specified torque valu All the fittings found outside the specified torque range were correcte Further analysis has begun to show that the as-found values were acceptable from an engineering standpoint, although they were outside the licensee's internally specified values.. The licensee's evaluation of this issue is still in progres The licensee has inspected other solenoid valves for signs of o-ring failure with no immediate concerns identified. Their investigation into this problem is also still in progress and will continue to be followe No violations, deviations, unresolved, or inspection followup items were identified in this are.
Management Meeting (30702)
A management meeting was held on September 30, 1993, between Hoffman, Vice President Consumers Power Company, and H.J. Miller, Deputy Regional Administrator, Riii, and their respective staffs. The purpose of the meeting was to discuss significant events which have occurred during the current refueling outage and the short and long-term initiatives Palisades plans to implement for corrective actio Attachment 1 is a copy of the material present~d by the license A second meeting was held on October 21, 1993, to discuss events and corrective actions associated with primary.water stress corrosio cracking (PWSCC) *at Palisades. Attachment 2 to this report is a copy of the licensee's handout from this meeting.
1 Persons Contacted Consumers Power Company
- D. P. Hoffman, Vice President, Nuclear Operations
- G. B. Slade, Plant General Manager
- R. D. Orosz, Nuclear Engineering & Construction Manager
- R. M. Rice, Director, NPAD T. J. Palmisano, Plant Operations Manager
- D. W. Rogers, Safety & Licensing Director
- K. M. Haas, Radiological Services Manager J. L. Hanson, Operations Superintendent R. B. Kasper, Maintenance Manager
- K. E. Osborne, System Engineering Manager
- C. R. Ritt, Administrative Manager
- J. C. Griggs, Human Resource Director
- H. A. Heavin, Controller
- D. J. Fitzgibbon, Shift Supervisor
- G. J. Daggett, Material Management Superintendent
- G. B. Szczotka, Staff Engineer, Nuclear Training Department Nuclear Regulatory Commission (NRC)
- J. B. Martin, Regional Administrator
- H. J. Miller, Deputy Regional Administrator G. E. Grant, Director Designate, Division of Reactor Safety W. M. Dean, Acting Director, Project Directorate, 111-1, NRR
- A. H. Hsia, Project Manager, NRR
- B. L. Jorgensen, Acting Chief, Reactor Projects Branch 2
- T. J. Kobetz, Acting Chief Reactor Projects Section 2A
- C. N. Orsini, Reactor Engineer, Reactor Projects Section 2A
- *M. E. Parker, Senior Resident Inspector
- D. G. Passehl, Resident Inspector
- Denotes those present at the management meeting on September 30, 199 *Denotes those present at the management meeting on October 21, 199 *Denotes those present at the exit meeting on November 16, 1993
Other members of the plant staff, and several members of the contract security force, were also contacted during the inspection perio..-**
....
,,,.----.
STRATEGIES TO PRODUCE C SISTENT, IIlGII PERFORMANCE PERFORMANCE ISSUFS ACKNOWLEDGE CHALLENGE - BY EVERYONE (including DPHoffman and staf Driving Force: Our human peifonnance is not allowing us to achieve the standards and goals set in our Business Pla AN EXCESSIVE COOLDOWN RATE OF THE PRIMARY COOLANT SYSTEM OCCURRED. WE MUST LEARN ALL WE CAN FROM THIS AND PREVENT THIS KlND OF PERFORMANCE ERROR FROM HAPPENING IN THE FUTUR NOD MANAGEMENT HAS A RESPONSIBILITY TO ASSURE THAT THE APPLICATION OF DESIGN AUTHORITY IS ROBUST AND EFFECTIV Pnli~arlcs Nuclear Plant s.~ph!mbt~r 29, I 99]
STRATEGY Senior NOD management will communicate to all NOD employees the significance of inconsistent performance. Gain acceptance by all NOD employees that perfonnance needs to be improve Root cause assessment of this event will be performed. The implications for senior NOD management action will be assessed prior to startup. The NRC will be briefed regarding the actions *related to the cooldown even A review of recent design engineering related plant events will be conducted, including the inoperable safeguards room _cooler fan Necessary corrective actions will be determine Immediate corrective actions will he completed prior to startu Page I I RFSPONSIBILITY I TPHagan -
Prior to Critical GBSladc - Prior to llcatup RDOrosz - Prior to Power Operations
.. _ ~* *~ *.::::
STRATEGIES TO PRODUCE NSISTENT, lDGll PEIU"'ORMANCE PERFORMANCE ISSU~
RECENTLY, PLANT EQUIPMENT WAS CHANGED WITHOUT INVOKING THE
'MODIFICATION PROCES A PRlMARY OBJECTIVE OF NOD IS TO MAINTAIN THE MATERIAL CONDITION OF THE PLANT WITHIN ITS DESIGN BASIS. RECENTLY, A PRESSURIZE NOZZLE TO PIPING WELD CRACK DEVELOPED TO BECOME A THROUGH WALL CRACK. THE IMPLICATIONS OF THIS EVENT MUST BE ADDRESSE Pafo;adcs Nuclear Plant September 29, 1993 STRATEGY RESPONSIBILITY Plant administrative processes will be reviewed to RDOrosz - Prior to determine if they allow 'changes' to the plant Power Operations outside the approved Design Basis without adequate review. Procedure and organizational responsibilities will be changed as necessary to prevent thi A plant specific engineering evaluation of the RDOrosz - Prior to primary water stress corrosion cracking concern II cat Up will be performed and will be completed prior to plant star1up. A strategy will he developed to deal with industry concerns related to primary water stress corrosion cracking. Palisades issues will be assessed and dealt with as a generic concern relevant to the industr All actions required to assure safe op~ration in the next operating cycle will be completed prior to startu Page 2
STRATEGIFS TO PRODUCE CONSISTENT, IIlGII PERFORMANCE PERFORMANCE ISSumi NOD MANAGEMENT RECOGNIZES ITS NEED TO BECOME MORE EFFECTIVE IN CREATING THE CONDITIONS AND CAPABILITIES WHICH WILL PRODUCE MORE EFFECTIVE PERFORMANCE. IT WILL SEEK CANDID AND OBJECTIVE FEEDBACK ABOUT ITS PERFORMANCE AND WILL CONTINUOUSLY WORK TO IMPROV MANY VERBAL COMMITMENTS HA VE HEEN MADE IN PUHLIC AND NRC MEETlNGS. NOD TAKES THESE COMMITMENTS SERIOUSL THE SAFETY SIGNIFICANCE OF*
DISCREPANCIES IDENTIFIED BY CCP REVIEW NEEDS TO BE QUANTIFIE Palisades Nuclear Plant September 29, 1993 STRATEGY An assessment will be performed to review management's current approach. to the challenges we face and will provide information to corporate management about the effectiveness of our management process and proposed action An independent consultant (Tenera) will be ut.ilized to support this assess.men Corporate management will be briefed on INPO field notes and the information obtained from the site debrie Public meetings and enforcement conference meeting minutes will he reviewed to ensure all commitments have been me The CCP discrepancy list will be reviewed to ensure the Plant Review Committee has been made aware of any significant issues. Any new issues will be dispositioned prior to plant startu Page 3 RESPONSIBILITY RMRice - Prior to Critical DPlloffman - Prior to Power Operations G BS lade - Prior to Critical RDOrosz - Prior to Heat up r
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S'rRATEGIES TO PRODUCE CONSISTENT IIIGH PERFORMANCE
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PERFORMANCE ISSUES PROCEDURAL PROBLEMS RELATED TO TFST AND SURVEILLANCE ACTIVITIES HAVE BEEN PREVIOUSLY IDENTIFIED AND MUST BE RESOLVE THE INDETERMINATE STATUS OF NODS HAS BEEN A SIGNIFICANT CONTRIBUTOR TO AMBIGUITY. SOME NODS HA VE NOT BEEN CANCELLED AS PLANNED; AND SHOULD B HUMAN PERFORMANCE ROOT CAUSE EVALUATIONS PROVIDE lMPORTANT INFORMATION FOR NOD MANAGEMEN NOD MANAGEMENT RECOGNIZES ITS RESPONSIBILITY AS A NUCLEAR LICENSEE TO KEEP THE NRC INFORME Palisades Nuclear Plant September 29, 1993 STRATEGY A nmlti-disciplirrnry h'am review will he conducted of all new or previously problematic special tests and surveillance procedures that will be implemented prior to startup. The review and any procedural revisions will be performed prior to their use during plant startu An SMSC Meeting will be conducted to cancel NODS which are ready for cancellatio Palisades HPES Coordinator will review 1993 Human Performance Event/Deviation Reports to determine if there are any common causal factors which have not been addressed. The *results for Palisades will be reviewed prior to plant startup at a Plant Review Committee meeting which will include the NOD Senior Management tea A conlrnunication strategy to keep NRC informed of significant events will he developed. Key contacts, frequency of meetings, and communication tools will be develope Page 4 RFSPONSIBILITY C.l\\Slaclc - All milestones (as procedure is required)
JJ Fremeau - Prior to Critical CBSlade - Prior to Critical PMl>onnelly - Prior to lleatup
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PROCESS FOR LONG-TERM HUMAN PERFORMANCE IMPROVEMENT BUSINESS PLANNING MEETING -
OCTOBER 14
Accelerated Human Performance Action Plan
Long-Te~m Human Performance Action Plans
Status of Short-Term Human Performance Action Plans NRC UPDATE MEETING -
LATE OCTOBER TENERA DIAGNOSTIC ASSESSMENT -
OCTOBER 29
Root Cause of Human Performance Problems
Cultural/Institutional Issues
Recommendations for Improvement BUSINESS PLANNING MEETING
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NOVEMBER *
TENERA Diagnostic Recommendations
"Performing on the Job" Strategy
Review of Performance Against Business Plan Targets
Status and Update of Business Plan Action Plans The outcome of this meeting will be an updat~ to the Business Plan for 199 NRC UPDATE MEETING -
LATE NOVEMBER BUSINESS PLAN MEETING -
EARLY JANUARY
1994 Goals and Objectives (Reflecting the Updated Business Plan)
NRC UPDATE MEETING -
LATE JANUARY 1994 BUSINESS PLAN MEETINGS TO BE SCHEDULED
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News of immediate interest to employees of NOD September 27, 1993 PALISADES MONITORING PROGRAM UPDATE This year a number of issues and events arose involving human performance at Palisades. These issues and events have impacted our continued success as an organization and as such, demands all of our attention to improv Over the past several weeks, the Palisades Management Team has been collectively meeting to create a plan of action to help us understand and address these human performance issues. The first step of.the plan involves systematic monitoring of individual and group performance through field observation as well as Corrective Action and Human Performance Enhancement System (HPES) trending. The field nitoring of performance will be conducted as an addition to the periodic reviews organizational performance, equipment performance, and material condition that have been on-going to determine the need for action in these area The principal concept of the program is to observe the general environment employees work if): the procedures, the processes, and the barriers that employees cope with to accomplish their work. Then we will work to break down those barrier Field notes will be used by the Management observers, similar to an INPO Field Team, to permit recall and follow-up. Each manager will conduct two observations monthl * We will then meet to compare notes and define actions and additional areas to monitor the next mont In addition, within the coming week, Vice President, Nuclear Operations, David Hoffman, will be issuing an all-employee communication detailing an Action Plan for the NOD Management Team to address Human Performance improvement for both the short-and long-term futur (
PALISADES MONITORING PROGRAM Revision 0 PURPOSE To systematically monitor organizational performance issues through field observation, corrective action and HPES trending, NPAD monitoring, and NOD management monitorin II. CONCEPT Palisades Plant Management will establish a focused monitoring plan which incorporates direct field observations, reviews of results of existing management systems, and outside observation activities by NPAO and other A periodic review of the current level of organizational performance, equipment performance, and material condition will be held to determine the need for additional actio II PLAN Palisades Plant Management staff shall meet and identify specific topics or areas to be monitored for next quarte The purpose of this discussion is to focus monitoring efforts so that consistent data is obtained for further revie A monitoring schedule shall be established monthly identifying topics or activities to be monitored and specific individuals assigned to monito Monitoring shall be performed as assigne Written summary of observations should be submitted within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of monitorin Monthly review meeting shall be held with Palisades Management Staff to review results of following, as appropriate: Field monitorin.
HPES review.
Corrective Action trendin ;
Departmental monitoring systems (RDR's, rework monitoring, etc.). NOD management visit progra.
NPAD monitoring result.
Internal and External audit report result (
Page Two Monthly review meeting should have a set forma Specific topics which should be covered include: Human Performanc.
Work Process performanc.
Equipment performanc.
Effectiveness of previous action.
Topics for future field monitorin Action assignments shall be made based on the results of the review meetin Quarterly, the Plant General Manager should report the results of this monitoring program with the VP Nuclear Operation I RESPONSIBILITIES The Plant General Manager is responsible for the implementation of this progra He is also responsible for quarterly reporting of results to the ~p Nuclear Operation The Operations Manager is responsible for the scheduling of monthly review meetings, including written documentation of the results of the meetin Various Management personnel are responsible for performing monitoring activities as assigne RECORD REQUIREMENTS The only records required to be retained are the minutes of the monthly Management Review Meetin These records shall be fiied in DC V PROGRAM CHANGES The program may be changed with the approval of the Plant General Manage Changes to the program shall be docwmented as revisions to the progra Plant General Manager
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FIELD OBSERVATION GUIDANCE
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- Observe* as* many aspects of the activity as is practical including the pre-'job briefing and the follow-up documentatio.
Observe - don't supervise. You are there to observe not consul (Unless, of course, you need to intervene to prevent personal injury or equipment damage.) It is okay to ask questions but do not provide directio.
Log TIME on the Field Notes periodicall (If something unusual occurs you may want to increase the frequency of your timing notes. For example: if they are interrupted by the page and it would help your evaluation to know how long it took them to get back on track.) *Include every factual observation and interaction that occur.
Try to withhold judgement while taking Field Notes. Record facts not opinions.. Evaluate the Field Note facts using the Field Observation* Checklis.
Conduct a feedback session with the workers and supervisor. If a problem or *
deviation occurred, determine if a corrective action document is warrante..
Forward all documentation to Debbie. Beac *..
Workers Involved Supervisor Observer Activity I TIME I
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OBSERVATION FIELD NOTES OBSERVATION FACTS bate/Time Location I EVALUATION I
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I TIME I OBSERVATION FIELD NOTES (Continuation Sheet)
OBSERVATION FACTS Page_ of_
I EVALUATION I
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FIELD OBSERVATION JOB AID Workers Involved Supervisor Observer DatefTime Location
. Activity PREPARATION Individuals are trained and qualified to perform the task Prejob briefing conducted IAW prejob briefing job aid Prerequisites were completed and verified Procedures, precautions, and limitations were reviewed Tools/materials were available and ready Job was properly planned and scheduled PERFORMANCE EVALUATION Crew successfully completed the task Written instructions were used and adhered to and were adequate Crew(s) worked together as a team to.complete the task Industrial safety standards were met Radiation safety standards were met Initiative and innovation were demonstrated in completing the job Decisions made were conservative Supervision kept appraised of job status and problems Communications IAW departmental communications policies REMEDIAL ACTIVITIES Effective proper feedback provided to workers Appropriate corrective action documents generated Procedures changed as required CLEANUP/DOCUMENTATION Lessons learned captured in historical file Tools and equipment returned to the original/proper condition Documents/records were completed correctly Work area returned to the original/proper condition SAT NII
NOTES/COMMENTS ~nclude comments for all needs improvement (N/I) items and for any exceptional performance)
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To From Date Subject cc MAEngle, Palisades BAFarnworth, Palisades DLBeach, Palisades September 8, 1993 PALISADES PLANT:
MANAGERS MEETINGS CSmith, Palisades CONSUMERS POWER COMPANY Internal Correspondence Below is a list of upcoming managers' meetings, to be held in the Managers Conference Room from 1100-1300, including dates and topic(s) for discussio DATE TOPIC(S)
September 14 STANDARDS AHO EXPECTATIONS
Development
Communicate
Monitor
Accountability October 6 HOHITORING
Review Program
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MAHAGIHG CHAHGE Identification Assessment Planning Implementation Priorities October 19 COHMUHICATIOH
Content
Tools Feedback November
LEADERSHIP ANO MANAGEMENT SKILLS
Coaching
Counseling
Feedback
Development Moves Department Head Alignment December. 9 PERFORMANCE RECOGNITION
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- (AS DETERMINED BY ER'S AND DR'S)
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1989 1990 1991 1992 1993 YEAR TO DATE Both D/Gs Out Of Servic Misaligned Iso-Phase Bus Cooling.
CV-0521 Mispositionin Tagging Error On C-903 Improper Lineup For Q0- CRDM 20 & 31 Excercized In Erro SFHM Underload Bypassed.
SID Cooling Temperature <70° Failure To Uncouple Control Rod Failure To Close MV-PCI094A During R0-65 19.93 (YTD)
500 450 400 350 300 250 200 150 100
935090 PCS HEA~P I COOLDOWN
PORV LINE CRACKED WELD LOCATED IN HOT SHUTDOWN SDCRETURN~
TEMP 50-+-r-r-r---r---,---~~r-T--r--.---,--,----.--.--r---T-.---r--r--.--.--.-----.--.---.--.--,----.--.-.--.--.---,---.----i
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12:00 AM 12:00 PM 12:00 AM 12:00 PM 12:00 AM 12:00 PM 12:00 AM Sep 15,'93 Sep 16,'93 Sep 17,'93 Sep 18, '93
935090 PCS-COOLDOWN
PCS COOLOOWN AFTER PORV LINE CRACKED WELD FOUND IN HOT SHUTDOWN op 200 SDCRETURN TEMP
)Iii 150 100 ----------------111111111m---...--J 50-t--.--r---.-r-r--r--.-r-,--r-r-r~-~---r-~.----..-~-.-.,__._~~--r----r-~----r-~.----..-~_,_.----..-~-.-.~---r-.,.-,.~--.----i 6:00 PM 9:00 PM 12:00 AM 3:00 AM 6:00AM 9:00AM 12:00 PM 3:00PM 6:00PM Sep 16, '93 Sep 17, '93 2 MINUTECFMS DATA
935090 PCS COOLDOWN PCS COOLDOWN AFfER PORV LINE CRACKED WELD.FOUND IN HOT SHUTDOWN 150 OF 130....................................................................................................................................................................................................
120......................................................................................................................................................................................,...................................................
100-1-r-~~~-.-.-~-.-.--~~~-r-T-.-.--.-r-~..-.-~..-..-.---..-r--,--,.--.-.---.-..-..-.-..-.-,~---..-r-.-.--.-.---.-..-..-.-..-.-,-.--.--.-.--.-.--.-..-~.-.-.-.-r-1 12:00 PM 12:30 PM 1:00 PM 1:30 PM 2:00 PM 2:30 PM 3:00 PM 3:30 PM 4:00 PM 2 MJNUTECFMS DATA Sep 17, '93
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Attachment 2 CONSUMERS PO\\VER COl\\1PAiW l\\'IBETING WITH NRC INCONEL 600 ISSUES INDUSTRY AND PALISADES EXPERIENCE OCTOBER 21, 1993 AGENDA Introduction RDOrosz Industry Inconel 600 Experience PD Fitton
- Failures and Corrective Actions
- Safety Assessments Palisades Experience DJV aride Walle
- Failures and Corrective Actions
- Safety Assessments 4. Cause of the Relief Valve Nozzle RBJenkins Safe-End Crack Palisades Inconel 600 Program DABemis Long Term Corrective Actions RDOrosz
INDUSTRY INCONEL 600 EXPERIENCE CEOG formed an Inconel 600 Working Group in late 1989. This group was formed to investigate the Primary Water Stress Corrosion Cracking (PWSCC) leaks at:
Calvert Cliffs San Onofre Unit 3 (1986)
St Lucie Unit 2 (1987)
EdF (7 Units) (1989) (France)
AN0-1 (B&W) (1990)
GOALS:
1. Determine root cause and contributors 2. Evaluate susceptibility to PWSCC at other CE plants 3. Determine safety consequences if a leak were to occur 4. Establish methods and frequency of inspection 5. Develop repair methodologies
INDUSTRYINCONEL600EXPERIENCE IN SUMMARY WE LEARNED: *
1. For Primary Water Stress Corrosion Cracking (PWSCC) to occur we must have: Susceptible material High temperature High tensile stre.sses* *
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INDUSTRY INCONEL 600 EXPERIENCE OUR PRELIMINARY EVALUATION CONCLUDED:
1. The pressurizer is by for the most susceptible location due to normal operating temperature. High strength material was most susceptible because it could retain higher levels of residual stress without yielding, and hence limiting the stres. PWSCC in inconel 600 penetrations is an economic issue, not a safety issue. This conclusion was based on work demonstrating:
A. We expect axial cracks with J weld configurations 1. Field data 2. Analytical modeling 3. Mockup measurement of residual stresses B. Leakage could occur for several fuel cycles without compromising the integrity of the PCS pressure boundar. Leakage is most likely to occur in the pressurizer vapor spac (
INDUSTRYINCONEL600EXPERIENCE PRESSURIZER LEVEL TAPS Unique Butt weld safe end 2. Contracted CE to measure residual stresses in mockup 3. Conclusion: No significant residual tensile stresses are not produced by this configuration. (Reference 91-ESP-77 A)
CONTROL ROD DRIVE MECHANISM REACTOR HEAD PENETRATIONS 1. Goals same as pressurizer 2. Palisades is participating in these activities 3. Work to date has supported conclusion that PWSCC does not jeopardize PCS integrity for these penetrations
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PALISADES INCONEL 600 EXPERIENCE
Pressurizer relief valve nozzle safe-end crac *
circumferentially oriented crack in heat affected zone of safe-end-to-pipe weld repaired weld with like design Pressurizer temperature instrumentation nozzle leaks.
axially oriented cracks in nozzle near structural "J" weld to pressurizer vessel temporary modification involving installation of exterior weld
"pad" addition and severing of Inconel sleeve between "pad" and "J" welds
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3" Relief Valve
~z:zle 3" s atety Valve Nozzle l"
l" Level Nozzle Toi) View Level Noz:zle-~~~-4 Surqe Nozzle Bottom View Manway 3"
Safety Valve
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110 !&' O IA II nside Cladding Elevation rlC:zUP.£ f l
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PRESSURIZER SHED GENERAL EQUIPMENT LAYOUT
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<TEMPERATURE ELEMENT PROPOSED EXTERIOR*
WElO "PAO" AOOiTION C.S. PRESSURIZEF:
S.S. CLADDING/
T-72 PRESSURIZER
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. PALISADES INCONEL 600 EXPERIENCE Safety Assessment
Pressurizer relief valve nozzle safe-end crack Evaluated other pressurizer and PCS nozzles with Inconel 600 for susceptibility Examined susceptible nozzles on the pressurizer Evaluated lifetime of repaired pressurizer relief valve nozzl Repair is considered temporary with a lifetime of one refueling cycle. A permanent repair will be implemented* during the 1995 refueling shutdow Evaluated margin-to-failure of cracked pressurizer relief valve nozzle and concluded that nozzle had sufficient strength to withstand normal and faulted conditions Evaluated primary.coolant system leakage detection system and
. concluded that similar cracks would be detectable. This leak detection capability was demonstrated during the pressurizer relief valve nozzle lea fi
PALISADES INCONEL 600 EXPERIENCE Safety Assessment
Pressurizer temperature instrumentation nozzle leaks Confirmed that cracking is similar (i.e. axial orientation) to that experienced in other industry instrument nozzles Evaluated other pressurizer and PCS nozzles for susceptibility Examined susceptible nozzles on pressurizer and PCS loop Evaluated lifetime of temporary modification. Modification is
. considered temporary with a lifetime of one refueling cycl * Combustion Engineering Owners Group generic safety evaluation applies
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PALISADES PRESSURIZER SAFE-END CRACK Cause of the Safe-End Crack
For primary water stress corrosion cracking (PWSCC), there are three required factors. Th.:,ese are:
A metallurgical condition An aggressive steam/water environment A threshold stress value
The safe-end was made of a material with a high yield stress which made it vulnerable to PWSCC. These vulnerabilities are reflected by:
A high hardness HRC-22 A high yield stress of 77KSI An assumed low (1600-1700°F) post-forging heat treatment
The safe-end was exposed to stagnant steam at approximately 640°F. The material would be highly sensitive to cracking at this temperatur *
Although externally applied piping system loads induced by pressure, weight and thermal expansion meet design requirements ~d are relatively low, significant local stresses appear to exist at the Inconel/stainless steel safe-end wel These stresses are due to:
The field welding process The mismatch in pipe/safe-end sizes The materials mismatch thermal effects
To add insight into the load assessment, analysis of dissimilar metal effects and temperature stratification were conducte These evaluations led to the conclusion that these loadings along with the external piping loads, as induced by the piping
- support system, were not the driving forces for the crack. The local pipe/safe-end configuration and fabrication process are judged to be as the primary contributor.,;,
PALISADES PRESSURIZER SAFE-END CRACK Repair of the Pressurizer Safe-End
The heat affected zone of the inconel material was removed and the weld prep machined to assure that the new heat affected *
zone does not have intergranular penetration or discontinuities that could act as stress riser *
The fit-up between the safe-end and stainless steel pipe was improve * * The inner surface of the weld was ground smoot *
There were no repair weld deposits made to the inner surface of the wel *
All of these improvements and changes will extend potential crack initiation time significantly~ resulting in increased lifetime for the repaired weld as compared with the original wel *
Very conservative crack propagation calculations indicate more than 20 months at temperature and pressure (640°F, 2060 psia)
would be required for a crack to grow through wall. For the
- 15 month fuel cycle, an initial 0.039in semi-elliptical crack of a 6: 1 aspeet ratio would just grow through-wal *
Time to crack initiation in the heat affected zone of the new safe-end should be comparable with that of the original safe-end heat affected zone. No credit is taken for this in the calculation of time to through wall crackin *
Therefore, the lifetime of the new weld will exceed one operating cycl *
PALISADES INSPECTIONS/LEAKAGE MONITORING Palisades outage/startup inspections 1. ASME Section XI ISI inspections 2. PCS boric acid walkdowns 3. System Engineering inspection of pressurizer heater sleeves 4. PCS walkdown at 2150 psia after each refueling outag On line leakage monitoring 1. Daily PCS leakage calculation 2. Monitoring containment parameters Sump level Containment temperature Containment humidity Radiological conditions Biweekly containment tour of lower elevation
- PALISADES INCONEL 600 PROGRAM BEFORE 1995 REFOUT Plan for replacement or justification of continued operation of temporary Inconel 600 penetration repair Implement comprehensive PCS penetration inspection program, based upon industry and Palisades experience: Maintain list of PCS penetrations prioritized for inspection and maintenance, including the following considerations: Material type Postulated mode(s) of failur Potential safety and economic impacts from failur Expected reliable life (Reliable life* ends when continued use or repairs are not justifiable for safety, reliability, exposure, or economic reasons). Conditions required for inspection, repair, and replacemen.
Establish enhanced inspection program for PCS penetrations approaching end of expected reliable lif.
Develop plan to* improve or replace penetrations prior to expected end of reliable lif.
Ensure that inspection program effectively bounds conditions under which cracking is foun.
Evaluate and qualify non-destructive examination techniques for detection of PWSC C.. Develop remediation plan for cracked and leaking Inconel 600 penetration PALISADES INCONEL 600 PROGRAM I REF_OUT Replace: Temporarily repaired penetrations that are not justified for continued operation past one cycl.
Other penetrations identified for replacement with improved materials or design Inspect old, removed penetration components for: Adequacy of previously performed repair.
Evidence of aging as compared to expected rat.
Verification of conformance to expected failure mode Inspect penetrations in accordance with PCS penetration inspection progra Based upon inspection results, expand inspection scope as needed to bound cracked penetration Remedy found cracked and leaking penetrations in accordance with remediation pla II SUBSEQUENT TO 1995 REFOUT Adapt penetration inspection program and remediation plan after each Refout, based upon updated industry and Palisades experienc Execute penetration inspection program and remediation plan during each Refou PALISADES INCONEL 600 EXPERIENCE Long Term Corrective Actions
Evaluate design of pressurizer relief valve nozzle and PORV line and perform modifications necessary to assure a suitable nozzle lifetim *
Perform permanent modification to pressurizer temperature instrumentation nozzle *
Develop comprehensive program to deal with Inconel 600 issues at Palisades. Program to include:
Evaluation and qualification of non-destructive examination techniques for detection of PWSC Development of an augmented inspection program for Inconel 600, including temperature nozzles, safe-ends and control rod drive nozzle Planning for replacement of Inconel 60 Contingency planning for inspections/repairs of ~y future leaks.