IR 05000255/1991004
| ML18057A771 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 03/04/1991 |
| From: | Jorgensen B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18057A770 | List: |
| References | |
| 50-255-91-04, 50-255-91-4, NUDOCS 9103130012 | |
| Download: ML18057A771 (14) | |
Text
-*
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No; 50-255/91004(DRP)
Docket No. 50-255
- License No. DPR~20 Licensee: Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Fa~ility Name~ Palisades Nuclear Generating Plant Inspection At:
Palisades Site, Covert, MI Inspection Conducted:
January 12 through February 18, 1991 Inspectors: J. K. Heller J. A. Hopkins D. L. Waters Approved 2A Inspection Summary Inspection on January 12 through February 18, 1991 (Report No. 50-255/91004(DRP))
.
.
Date
Areas Inspected:
Rbutine unannounced inspection by the resident inspectors of plant operations,.maintenance, surveillanc*e, refueling activities; and security; and, special unannounced inspection by an NRC contractor of design change No Safety Issues Management System (SIMS) items were closed. *
Results:. Nb violations, deviations, unre~olved or open items were identifie *.The strengths, weaknesses and Open Items ar~ discussed in paragraph 8,
"Management In.terview.
9103130012 910304 PDR ADOCK 05000255 Q
- DETAILS Persons Contacted Consumers Power Company G. B. Slade, Plant General Manager
- R. M. Rice, Plant Operations Manager
- D. J. VandeWalle, Technical Director R. D. Orosz, Engineering and Maintenance Manager K. M. Haas, Radiological Services Manager J. L. Hanson, Operations Superintendent R. B. Kasper, Mechanical Maintenance Superintendent
- K. E. Osborne, System Engineering Superintendent C. S. Kozup, Technical Engineer
- K. A. Toner, Plant Projects Superintendent Nuclear Regulatory Commission (NRC)
- J. K. Heller, Senior Resident Inspector
- J. A. Hopkins, Resident Inspector Parameter, In *D. L. Waters, Consultant to NRC
- Denotes some of those present at the Exit Interview on February 22, 199 Other members of the plant staff, and several members of the contract
. security force, were also contacted during the inspection perio.
Operational Safety Verification (71707, 71710, 42700)
Refueling and plant operations, at cold shutdown, were observed as conducted in the plant and from the control roo The performance of Reactor Operators, 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 General The plant began this reporting period in a refueling shutdown condition with the vessel defueled and all fuel in the spent fuel poo The inspector verified by observation, discussion with the control room operators, and review of checksheets that the spent fuel pool cooling system was operabl This included verification that the fuel pool temperature was maintained, spent fuel ventilation was operable during spent fuel pool activities, cooling water was available to the spent fuel pool heat exchangers, and emergency power was availabl *
b. * *
Major Outage Milestones Completed During This Report Period (1) *Commenced refueling - January 2 (2) Completed main condenser construction and static hydrostatic pressure test - January 2 (3) Completed refueling the reactor - January 27 (see paragraph 5, *
11 Refueling 11 for the inspector's observations).
(4) Completed Steam Generator Hydrostatic pressure test - February 11 (see paragraph 7, 11 D~sign Changes" for the inspector's observations).
(5) Completed containment full pressure structural integrity test February 16 (see inspection report 255/91003(DRS) for the
. inspector's observations).
(6) Completed containment integrated leak rate test - February 17 (see inspection report 255/91003(DRS) for the inspector's observations).
Technical Spe.cification 3.1 On January 25, the licensee determined that Table 3.17.1 of Technical Specification 3.17; "Instrumentation and Control Systems" was unclear and could be misinterprete Technical Specification Amendment 130 (issued March 23, 1990) added a requirement that portions of the reactor protective system be operable when the control rod clutch power supply is energize The licensee needed to energize the power supplies to lower the control rod drive extensions from the head to permit latching of the control rod The reactor protective system would not be operable for a number of day The licensee performed a safety evaluation and determined that the requirement did not apply when the reactor was in cold shutdown with the primary coolant at refueling boron concentratio This appeared consistent with the action.statements of Technical Specification 3.1 The inspector discussed this evaluation with Region III and the NRR project manager-all a*greed with the safety evaluatio The licensee will be revising Table 3.17.1 and will include this item in the revisio CFR 50.72 Notification On January 25, 1991, a left channel containment isolation was initiated while preparing work on refueling isolation radiation monitor RIA-231 All operable components responded as require The planning of the *
- work order that deenergized RIA-2316, did not identify that removal of the power supply for RIA-2316 also deenergized RIA-180 This caused a left channel contai~ment isolation. The inspector interviewed the planner and found that the wiring diagram showed an incorrect configuratio The licensee corrective action for this event
addressed the wiring diagram erro This subject will be revisited when the 10 CFR 50.73 Licensee Event Report is issue * *
No violations, deviations, unresolved or open items were identified.
Maintenance (62703, 42700)
Maintenance activities were routinely inspecte The focus of the i-nspection was to ensure that the maintenance ac.tivities reviewed were conducted in accordance with approved procedures, regulatory guides; industry codes or standards, and were in conformance with Technical Specifications~ The following items were considered during this review:
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 activities were inspected: Primary Coolant Pump P-50A,B,C and D Lube Oil System Modification (Facility Change 860). *
Containment Air Cooler, VHX-4, Modification (Specification Change 89-130).
VHX-4 copper cooling coils and pipe manifolds were replaced due to a hi story. of 1 eakag The new coils have a m6difi ed configuration to r~lieve pipe joiht stresses and use thicker walled pipe.. The licensee will monitor the performance of the new coils and determine if the modifications will be done to the other containment air coolers during a subsequent outag Component Cooling Witer Heat Exchanger Mai~tenance, CCS-M-Compbnent Cool~ng Water Heat Exchanger, E-548, tubes (service water side) were cleaned and inspected using eddy current testing. There was no iridi~~tion of biological fouling or sedime~t build-u Service Water Check Valve Replacement (Work Order (WO) No', 24006585, 24006586, 24006587).
While isolating the service ~ater (SW) header inside containment for this maintenance, SW return header containment isolation valve, CV-0824, failed to tlose properl Investigation discovered parts from the upstream ~ontainment air cooler SW check valves wedged in CV-082 Thre~ of the f6ur check valv~s were ~isassembled. The valve discs were detached from the swing arms and significant corrosion and metal degradation of the valve body interiors were identifie The licensee determined that inadequate flow velocity and upstream flow disturbances induced vibrations and hammering actions which may have caused fatigue failure of the valve internal All four check*
valves were replace (The fourth check valve was not inspected based on the condition of the other three.)
A visual inspection of the interior of the SW p1p1ng from the. check valves, downs~ream to CV-0824, did riot locate any of.th~ missin valve part The licensee believes that the parts were corroded to the point that they were swept out of the SW system and into the discharge tunne Additionally, CV-0824 was refurbished due to normal system wea Post maintenance hydr6static testing of the SW system was completed satisfactoril During the.maintenance activity to replace the check valves, contract personnel "tack welded" clamps tq the SW piping to assist in the alignm~nt of two valves. This is a commonly a~cepted practice for. alignment of valves but was not authorized by the work orde The clamps were removed and non-destructive examinati~n was performed on the.SW pi pi.n No defects were i dent ifi e The licensee reviewed the incident with the appropriate work group~.
The inspector had no additional concern Escape Airlock Inner Door Viewing Po~t Gasket L~ak (WO No. 24100098).
Durin~ the performance of Technical Speciffcation test S0-48, 11 Escape Airlock Penetration Leak Test 11, the inner door viewing port gasket exhibited air leakage gre~ter than the maximum acceptable.
'va 1 ue.. Mai nt.enance personne 1 found di rt on the. sea 1 i ng surface of the viewing port and.a number of loose bolt The gasket was
- replaced, sealing surfaces cleaned and all bolts ~ightene *
S0~4B was performed satisfactorily..
f:
Modifications of Clark Relay for the Containment Isolation System (WO 24100711, 24100456, 24004652)..
During post maintenance.testing of Clark Relay 5R-5, the licensee discovered that some contact spring relays were installed upside*
down, which could affect operability of the relay. It appears that the manufacturer's instructfons lacked sufficient information to
~6nvert individual c~ntacts f~om normally open to normally ~losed, resulting in improper installation of the c6ntact springs;* The licens~e's internal corrective action document (D-PAL-91-029) implied that the problem wa~ isolated to this outage, which eliminated the question of operability because the associated components were not yet required to be operabl~. The licensee inspected a sample of the Clark Relays which were modified during this outag Based on th results of the inspection, the licensee determined that the relay were properly assembled and will operate correctly. Additionally, surveillance testing ccinfirmed proper operation of the relay During a previous outage, the closing logic for the feedwater regulating valves, which uses Clark Relays, was modified:
The inspector asked if a similar problem exists with these relay The licensee determined that. the design change package had sufficient information to convert the Clark Relay In addition, the system engineer confirmed that the contact springs were correctly installe The inspector had no additional questions.
- *
No violations, deviations, unresolved or open items were identified.
Surveillance (61726, 42700)
The i~spector 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 fo~ 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 activities were inspected: T-213 T-261 Q0-02 S0-4A R0-65 T-226 Component Cooling Water Flow Test of the Component Cooling Water, Low Pressure Safety Injection, High Pressure Safety Injection and Containment Spray Pumps, and the Component Cooling Water Heat Exchange *Low. Pressure Safety Injection Pump, P-67A and P-678, Performance Tes Recirculation Actuation System and Containment Sump Check Valve. (Operability Verification).
Personnel _Air Lock Penetration Leak Tes High Pressure Safety Injection Train 1 and 2 and Hot Leg Injection Check Valve Tes Component Cooling Water (CCW) Accumulator Te~t for Containment Isolation Valves CV-0911 and CV-094 During the performance of T-226, CCW containment isolation valve CV-0911 failed to remain shu~ for the required time interval due to a leak in the air accumulator syste The accumulator (a backup system) was designed to maintain CV-0911 shut if the normal instrument air supply was interrupte (CV-0911 and CV-0940 are "fail open" valves.)
The leak was repaired under Work Order 24005748 and applicable portions.of T-226 were completed satisfactoril CV-0940 successfully completed T-226 without any problem FWS-I-18 Auxiliary Feedwater (AFW) Pump, P-8C, Trip on Low Suction Water Pressure (Logic Test).
While performing FWS-I-18, an alarm circuit card would not rese The card was repaired under Work Order 2410024 The test was completed satisfactorily.
-* RE-83A and B Service Test - Battery ED-01 and ED-02.
i.
Surveillances.RE-83A and B were performed to verify that the capacity of station batteries ED-01 and ED-02 were adequate to supply and maintain actual emergency loads for two hour RE-83A
. was completed on ED-01 satisfactoril Ho~ever, one minute into RE-83B, an equipment malfunction interrupted the tes Trouble-shooting allowed ED...;02 to "rest" for approximately ninety minute The test was re-initiated from the point it was interrupted and no additional concerns were identified by the license The licensee performed an engineering analysis to determine the overall acceptability of RE-83B with the ninety minute "rest".
The licensee determined that based on the results of the last test (1986), ED-02 would have responded satisfactorily without the ninety minute interruption. Additionally, since the lead-calcium C and D Model LC-25 batteries ~ere designed for infrequent discharges, performing.a second test discharge in a short time frame would unnecessarily accelerate battery agin The inspector reviewed the licensee's engineering analysis, with the assistance of regirin based specialists, and determined that RE-83B was accepta~l T-SC-90-022 Hot Leg Injection/Cold Leg Injection Flow Balanc *
On January 18, the inspector observed the pre-test briefing and observed that an extra licensed control operator (CO) was assigned to perform the test, a test engineer was assigned to coordinate the test, and the system engineer and a Quality Assurance Inspector were present to observe the tes The test was completed satisfactorily; however, some deficiencies we~e observe ( 1)
(2)
The procedure did not specify either the wide or narrow range on reactor vessel level instrument LlA-010 The test en~ineer stated that wide range (which was selected) was correc The method of communication between the field operators and control room was poo The field operators had to relay test data to a*telephone communicator to contact the control roo High noise levels at the telephone stations required control room personnel to shout their instruction This gave the *
control room a chaotic atmosphere and hindered other control room activities but apparently did not affect the outcome of the tes The licensee stated that poor reception at various plant locations prevented the use of radio The licensee has been evaluating the need for additional radio 11 repeaters 11 *
(3) Initially two steps in the procedure were missed, which were identified by the Quality Assurance inspecto He immediately iriformed the CO and the test enginee The steps were then performed satisfactorily. It appears that the steps were missed because the CO was not following the procedure in parallel with
- RE-39 the test engineer. *The inspector discussed this apparent loss of activity control with the Operations Superintendent who stated that licensed operators, not the test engineer, were responsible for the performance of test procedure The individuals involved, as well as the rest of the licensed operators, were briefed on their resp~ctive responsibilities during testin Prior to this event, the inspector had not observed similar deficiencies during the performance of other test procedure Hydrogen Recombiner M-69 A and B (Operability Verification).
During the performance of RE-39, M-69 A and B did not achieve rated power and temperatur Trouble shooting determined that potentiometers in the power controller circuit cards were out of adjustmen The circuit cards were adjusted and RE-39 was completed satisfactoril A review of M-69 A and B maintenance history identifi.ed that the power controller cfrc~it cards were sent to the manufacturer (Westinghouse.) for routine refurbishment at the beg-inning cif the current refueling outag The manufacturer tested the circuit ~ards using technical data for a later model in the hydrogen recombiner series, which resulted in ~he misadjustment of the potentiometer The licensee apparently did not supply sufficient information to ensure that the appropriate post refurbish~ent testing was performe The licensee was evaluating the instructions provided to vendors wh~n equipment was sent off site for refurbishment and testin The inspector reviewed the everit and determined that M-69 A and B were not req~ired to be operable for the existing plant conditions and that RE-39 was.scheduled prior to the plant achieving the applicable condition The inspector had no additional concern *No violations, deviations, unresolved or open.items we~e identifie.
Refueling 60710 On January 20, the licensee commenced fuel reload activitie The inspector reviewed procedures and checklists to verify that the refueling equipment and support systems were operabl During refueling operations, the irispectors observed fuel moves from the control room, spent fuel pool and the containmen On January 20, the licensee performed a video inspection of the core*
support plate to verify the location of the bolts for the core support barrel. A washer, approximately 2 inches in di~meter, was discovered on the core support plat Retrieval attempts were unsuccessful and the washer fell through a core support plate flow hol Subsequent attempts to retrieve the washer were unsuccessful.
- *
The licensee performed an engineering evaluation to det~rmine the potential effects the washer would have on the* fuel if the washer was not retrieve The licensee determined that although the source of th~ washer was unknown, it was similar to washers removed during the 1988 refueling outage ahd may not have _been removed at that tim The following is a summary of the licensee's engineering evaluation:
(1) Primary coolant system (PCS) flow could lift the washer through the flow holes in the core support plat *
(2) If the washer was pinned against the fuel assembly lower tie plate, it would not affect the Departure from N~cleate Boiling (DNB) condition during normal operatio (3)
The chances of the washer getting into the control rod channel are very remot *
(4) If the washer got through. the fuel assembly lower tie plate openings, it would wedge itself between the fuel rod Failure of 4 to 5 fuel rods due to fretting may occu The licensee concluded th~t additional attempts to ~etrieve the washer wer~ not appropriate because of the fuel performance during the last operating cycle when the washer was probably in the reactor, and the low probability of retrieving the washer without removing the ~ore support barrel.
The inspector reviewed the engineering evaluation and discussed the event with the system engineer and licensee.management and had no additional concern On January 21, during the reactor refue.l ing, *one fuel assembly was discovered 11 stuck 11 in its spent fuel pool (SFP) storage locatio The licensee continued with the refueling and raised SFP temperature to the upper end of the normal operating band to 11 loosen 11 the
assembl A second attempt to remove the assembly was u~successfu On January 24, the inspector attended a plant review committee (PRC)
meeting that reviewed the ~rocedure for removal of a stuck fuel bundle from the SF The inspector questioned if the safety review addressed the FSAR statement th_at mechanical interlocks are in place on all fuel handling equipment to assure 10 feet of water was maintained over the bundl The procedure required use of the overhead crane and specified that a person was stationed at the circuit breaker to prevent uncontrolled vertical movement of the overhead cran The inspector questioned if a person-was adequate compensation for a mechanical interloc Subsequent to the PRC
meeting, the inspector discussed the procedure with.the PRC chairma The safety evaluation was changed to reflect the inspector's question In addition, the inspector noted that the* _
procedure did not compensate for the weight of the rigging below the
load cel This meant that the lifting force* specified in the safety evaluation, was reduced by approximately 400 pound The licensee conservatively chose to implement the procedure with the reduced lifting forc *
The attempt to remove the stuck fuel assembly was unsuccessfu Reactor engineering identified an acceptable.alternate assembly and continued with the refuelin Misplaced Fuel Assembly On January 23, with 118 out of 204 fuel assemblies in the reactor vessel, an incorrect fuel bundle, serial number C-137, was removed from the spent fuel pool (SFP) and placed in the reacto The error was identified after the refueling machine (RFM) operator identified extensive bowing of the assembl The control room operator asked the RFM operator to verify the fuel assembly serial number and orientatio Neither were vi~ible. Further investigation determined that the correct fuel assembly, serial number J-42, was still in its correct SFP location, G-Refueling activities were immediately suspended and reactor subcriticality was re-verified using sourc~ range neutron instrumentatio Fuel assembly C-137 was removed from ~he reactor vessel and placed in the SF The inspector observed control room activities from the time the bundle was transferred to the reactor and recovery operations to transfer the bundle back to the spent fuel poo At all times, communications were clear and documentation of additional fuel moves were appropriately recorde Recovery operations were directed by the shift supervisor with input provided by the senior reactor operator stationed in the containment and the operations superintendent who was in the control roo The licensee's immediate corrective action was to brief each of the refue 1 i ng.crews on the event and stress the importance of notifying the control room as soon as questions or problems were identifie Reactor engineers also explained that all of the reload fuel assemblies had a serial number and orientation mark on top of the assembl Additionally, the Operations Superintendent made a new interpretation of Step 7.6.2.c (removing fuel from SFP locations) of System Operating Procedure (SOP) 28, "Fuel Handling Systems, 11 as requiring the spent fuel handling machine (SFHM) operator to check the manual index position of the SFHM prior to removing the assembl The procedure already required this check when inserting assemblies into the reactor vesse After the on-shift refueling crew was briefed, assembly J-42 was inserted into the reacto No other misplaced fuel bundles were identified during the remainder of the refueling proces The inspector attended the corrective action review board at which time the SFHM operator was interviewe The root cause of the event was the SFHM operator removing assembly C-137 from SFP location G-2 vice
..
- *
assembly J-42 from location G-The operator used a touch sensitive computer screen to identify the assembly's coordinates* and the SFHM automatically positioned itsel In this case, the operator selected G-2 vice G-It could not be determined whether the operator's error resulted from misreading the procedure or the touch sensitive scree Equipment failure was ruled otit, because subsequent comparison of the SFHM computer position ~ncoder against the manual index position display and the SFHM rail scribe marks did not identify any discrepancie The licensee performed an engineering evaluation to determine what affect the misplaced fuel assembly had on the shutdown margin (SOM)
of the cor The analysis assumed that the core was fully-loaded with all control rods removed and determined that the SOM was reduced approximately 0.15 percen This was approximately equal to reducing primary coolant system boron concentration by 13 pp Actual boron concentration was 1842 ppm and minimum required was 1720 pp The misplaced fuel bundle did not challenge reactor safet The licens~e 1 s immediate corrective action appeared to be adequat However, the licensee has not documerited the new interpretation of Step 7.6.2, SOP-2 The inspector discussed this observatibn at the exit interview Cl:nd the licensee agreed to revise the procedure prior to the next usag Loose Debris on Core Support Plate On January 25, the RFM operatoridentified a fuel assembly which was interfering with the insertion of a control rod into the cor Investigati~n determined that a piece of debris was in one of the core support plate fuel assembly alignment holes and was preventing level seating of the assembl The object could not be identifie The object appeared to drop through the alignment hole during retrieval effort Video inspection of tha area below the core support plate could not locate the objec The licensee concluded that it had fallen into the reactor vessel botto The licensee discussed potential retrieval plans and determined that based on the configuration of the core support plate with a full fuel load, continued retrieval efforts did not have a reasonable probability of succes The inspector asked if an engineering evaluation was performed to determine the potential effect of the loose object if it came in contact with a fuel elemen The licensee stated that due to the uncertain size of the object, there was not enough information to perform a reasonable evaluatio During the January 20, core. support plate video scan to verify support barrel bolt lo~ations (see paragraph 5.a), several pieces of debris, such as a plastic *tie wrap, a washer and a piece of wire, had been identified and all were removed except the washe Other vague discontinuities were observed but could not be identified due to poor water clarity and poor video imag The exact location of the
discontinuities was not known due to the 11 random 11 nature of the sca **
The licensee later determined that one of the discontinuities appeared to be in the same general area as the object s~en i~ the alignment hol Due to a communication breakdown no attempt had been made to remove this vague discontinuit *
The licensee was evaluating the need forcore barrel vacuuming after each core off-loa The inspectors considered controls to ensure removal of foreign materials. from the reactor vessel to be wea This was discussed at the management intervie No violations, deviations, unresolved or open items were identifie.
Security (71707)
Routine facility security measures - including control of access for vehicles, p~ckages, and personnel - were observe Performance of dedicated physical security equipment was verified during inspections in various plant area The activities of the professional security force in maintaining facility s*ecurity protection were occasionally examined or reviewed, and interviews were occasionally conducted with security force member On January 11, the licensee reported, pursuant to 10 CFR 26.73, that a contractor supe~visor tested positive during unannounced fitness-for-duty testing:
The individual's site access was revoke The licensee provided the site inspection st~ff with a list of work activities that the individual was performin The inspector.
verified, by review of selected work activities, that operability checks (pre-planned as part of the work activities) were performed and that the components worked as intende This information was provided to Region III fitness-for-duty specialist On January 23, the licensee. informed the resident inspector that a plant employee was arrested for possession of a contrtilled substanc The individual subsequently tested negative during fitness-for-duty testing. This information was provided to Region III fitness-for-duty specialis Any additional questions will be relayed by separate communication No violations, deviations, unresolved or open items were identifie.
Design Changes (37700) Testing was performed to confirm the adequacy of the system restorations performed during the changeout of the steam generators, included a hydrostatic test of the secondary system in accordance with Technical Specification 4.0.5 and ASME Section XI requirement The inspector reviewed the licensee's procedure for the rn~in steam and feedwater hydro, R0-701, 11Main Stearn and Feedwater System Hydrostatic Test, 11 Revision 0, prior to the performance of the tes The procedure was found to be acceptable overall, with minor concerns consisting of the following:
-*
(1) Assurance of guidance to operators in case an isolation of the Decay Heat Removal System 6ccurs during the tes (2) Step 5.2.15.b - the.minimum flow value called for in the procedure was not measurable.
. (3) Pressure fluctuations in the test pressure gauges caused by the positive displacement hydrostatic test pumps, should be dampened by use of snubbers or accumulator The licensee adequately addressed these concerns through procedure changes or operator briefing note The test was performed *on February 10, 1991, and was observed, in part, by the resident inspector staf The test documentation was reviewed following completion of the test. Minor problems, such as lack of correspondence of readings between the two calibrated precision pressure gauges du~ing_the p~essurization phase of the test; inadequate communication between hydrostatic test control operators and control room operators regarding secondary pressure effects due.to primary coolant temp~~ature changes, which led to a sever~l hour delay in stabilizing at the re qui red test pressure; and inadequate anchoring of the test manifold, resulting in movement of the ~anifold due to pulsations caused by the hydrostatic test pumps, were observed during the test. These problems were identified by and satisfactorily resolved by licensee personnel.
The four-hour*hold period at the test pressure was marked by stable pressure within the allowable band and less-than-expected leakage through boundary valves and packing leaks (less than eight gpm).
The inspector reviewed Work Order 257039 for pre-test and post-test calibration records for.the precision pressure gauges used in the test, and found that the gauges exhibited minor calibration deviations*
when post-test calibrations were performe These variations were not at the test pressure range and the minimum test pressure was maintained during the tes * During an inspection of the main condenser hotwell, the inspector found several downcomer pipes which were attached to the hotwell floor and others that were no The downcomers are int~nded to be free to vibrate to prevent damage to the floo The licensee stated that several were tack welded to the floor because of a misunderstanding of the work order instruction. This problem had already been identified and would be resolved after the condensor hydrostatic test. During a subsequent inspection the inspector verified that the downcomers were. free to mov The inspector had no additional question No violations, deviations, unresolved or open items were identifie * Management Interview (30703)
The inspectors met with licensee representatives - denoted in Paragraph 1 -
on February 22, 1991 to discuss the scope and findings of the inspectio In addition, the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection was also discussed.. The licensee did not identify any such documents/processes* as proprietar *
Highlights of the exit interview are discussed below: Strengths noted: (1)
Involvement of the engineering department in the resolution of problems (paragraphs 3.f, 11Maintenance, 11 4.h, 11Surveillance,
5.a, 11 Refueling 11 ).
(2)
Involvement of the Quality Assurance department in problem identification (paragraph 4.i.(3),
11Surveillance 11 )
(3) The integrity of the isolation boundary for the main steam hydrostatic test (paragraph 7.a, "Design Changes").
Weaknes~es noted:
(1). Actual work activity exceeded the scope of the planned work activity (paragraphs 3.d, 11Maintenance 11 and 7.b, "Design Changes").
(2) Poor communication method between the field and the control room (paragraph 4.i.(2), "Surveillance).
(3)
(4)
(5)
Inadequate instructions to control testing performed at a vendor facility (paragraph 4.j, 11Surveillance 11 ).
Administrative ~ontrols not in effecf to ensure foreign material was removed from the reactor vessel (paragraphs 5.a *
and d, "Refuel ing 11 ).
Personnel error that resulted in a mispositioned fuel bundle (paragraph. 5. c, "Refuel i ng 11 ). The fitness-for-duty problems were briefly discussed with the closing statement that the information was provided to Region II Any additional questions will be handled by separate communication (paragraph 6, 11Security 11 ).
14