IR 05000255/1995008

Insp Rept 50-255/95-08 on 950528-0703.Violations Noted But Not Cited.Major Areas Inspected:Operations,Engineering, Maint & Plant Support

ML18064A853
Person / Time
Site:	Palisades
Issue date:	07/27/1995
From:	Kropp W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18064A852	List: IR 05000255/1995008 ML18064A851
References
50-255-95-08, 50-255-95-8, NUDOCS 9508070154
Download: ML18064A853 (15)

Text

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

REPORT N /95008 FACILITY Palisades Nu~lear Generating Plant LICENSEE Palisades Nuclear Generating Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530 DATES May 28 through July 3, 199 INSPECTORS 0. Passehl, Resident Inspector R. Lerch, DRS Inspector

• S. Burgess, ORS Inspector M. Holmberg, DRS Inspector 0. Nelson, DRSS Inspector APPROVED BY

.* #r-2A AREAS INSPECTED A routine, unannounced inspection of operations, engineering, maintenance, and plant support was performe Safety assessment and quality verification activities were routinely evaluated.

9508070154 950727 PDR ADOCK 05000255 Q

PDR

RESULTS Assessment of Performance There was improvement in coordination and conununication of outage activities among all departments since the beginning of the refueling outage, whe *

series of errors occurred in a short period of tim These errors were described in. the previous inspection report No. 50-255/95007(DRP).

Positive observations were observed in ~lanning for major projects, reactor vesse disassembly activities, operator performance of nonroutine activities, motor operated valve testing, and in Engineering Department's implementation of the Alloy 600 projec The licensee exhibited adequate safety focus during this inspection perio This was evidenced by good management oversight of fuel reconstitution activities, the lowering threshold to identify and to correct pr.oblems in a timely manner, and in recognizing and planning for potential test problems associated with motor operated valve However, there were weaknesses in *

personnel attention to detail, a theme observed in the previous inspection report perio These weaknesses included:

control of dosimetry of personnel entering the radiological controlled area

mobile crane safety.

inattention that led to an error in s*etting the reactor vessel head

control of foreign material within the debris free zones of the reactor cavity and spent fuel poo *

Performance within the area of OPERATIONS was good (see Section 1.0). Since the beginning of the refueling outage, improvement was noted in communication, coordination, and management oversight of activities. This was evidenced in the Operations Manager's decision to stop fuel inspections based on the insp~ction crew becoming too narrowly focused on solving an immediate problem rather than the overall safety pictur Housekeeping in most areas of the auxiliary building was goo Housekeeping in the east safeguards component cooling water rooms and in the containment was adequat There were examples of material inappropriately stored on the landings of stairways and weaknesses were noted in the control of foreign material within the debris-free zone near the reactor cavit Radiological controls and postings in the auxiliary and containment buildings were all within regulatory requirement *

Performance within the area of MAINTENANCE was adequate (see Section 2.0).

The reactor vessel disassembly activities, including removal of the reactor vessel head, upper guide structure, and core support barrel was conducted wel During these major activities, the pre-job briefing packages were good;

the actual pre-job briefings were thorough and effective. Management involvement was evident with standards and expectations being reinforce Communication and coordination was also good. *Workers were cognizant of their respective responsibilities, and personnel access to the work areas was*

effectively controlle Lessons learned from previous problems were incorporated and were effectiv The licensee's control of contractors was goo A personnel error occurred during the removal of cables from the reactor vessel hea Maintenance workers mistakenly disconnected four core exit thermocouples (CETs) intended to be used as a backup indication for primary coolant temperature in the event of a loss of shutdown cooling. This event d~monstrated poor coordination and communication among work group However, the licensee's prompt identification and investigation of this error was viewed as a positive example of the licensee's lowering threshold to identify*

potential problem *

Prior to the loading of fuel, a weakness was observed during the interim setting of the reactor vessel head on the reactor vesse Personnel error on

• the part of the contractor crew resulted in unexpectedly lowering th~ head onto the vessel.*

Also, a personnel error occurred when a mobile Grove crane was driven to the switchyard to assist in the 1 ift of a battery 1 oad bank off the back of a truc While enroute to the switchyard, the driver of the crane inadvertently seveted an overhead 220 VAC power line supplying lighting to the main parking

• lot and guard hous No personnel injuries were reported, and no other.

equipment was affecte The cause appeared to be due to inattention to detail. During the previous inspection period, a mobile crane collided with an overhead support structure in the plant protected area. The inspectors

. considered the 1 i censee' s contro 1 of mobile cranes to.be a weakness.. This was an Unresolved Ite Performance within the area of ENGINEERING was good (see Section 3.0).

Engineering's performance of the Alloy 600 project was goo The licensee's use of mockups to demonstrate and qualify nondestructive examination techniques used in the inspections was a strength, as was the use of

"enhanced'; ultrasonic inspection techniques.* Among weaknesses identified with

_the plan, the significant reduction in planned inspection scope, precipitated by errors in estimating radiation dose and ALARA concerns, had the largest impact on the pla Wo~k tb support testing of motor operated valves was performed in a well coordinated manner as evidenced by effective pre-test briefings and good support from operations and contracted technicians. Operations Department's dedicated support in such items as breaker and valve line-ups also contributed to the considerable amount of testing completio Engineering Department demonstrated good safety insights to potential test problem Performance within the area of PLANT SUPPORT was adequate (see Section 4.0).

Activities associated with removal of the core support barrel an~ the transfer of the incore detector cask were well planned and execute Both activities

demonstrated that lessons learned from difficulties with previous high radiological risk projects (dry cask storage for example) had been applied.

. However, there were examples where the licensee was not aggressive at reducing dose for these and other job Total outage dose, after compl~tion of approximately 33 percent of the outage scope, was approximately 135 person-rem (1.35 person sievert).

Thi~ total ~as approximately 17 perc~nt ov~r the goal for that stage in the outage. If th~

trend continued, the total outage dose would be approximately 310 person-rem or 24 person-rem over the original goal of 286 person-re This overage was due, tn large part, to the unexpectedly high doses accrued during the Alloy 600 projec For a multitude of reasons, the plant experienced difficulties in meeting its ALARA goal for the Alloy 600 projec A noncited violation was issued when an individual entered the radiologically

. controlled area without an electronic dosimete Several other incidents occurred involving the e~trance of plant personnel into the radiologically controlled area with dosimetry turned of Summary of Open Items Inspector Follow-up Item: none Unresolved Item: identified in section 2. Non-cited Violations: identified in section *

INSPECTION DETAILS 1.0 * OPERATIONS NRC. Inspection Procedures 71707 ~nd-92709 were used in the perform~nce of an inspection of ongoing plant operation The findings showed performance was goo.1 Performance of Operations:

Since the beginning of the refueling outage, there was overall improvement in communication, coordination, and management oversight of operations department activities. Operators performed many activities without errors, including

.

surveillances, various system realignments, drain down of the primary coolant system, and core offload. During reactor vessel disassembly activities, good leadership was demonstrated by shift supervisors. Operators were aware of the status of equip~ent important to shutdown safet In addition, operators exhibited a good questioning attitude while on round One example was noted when ~n auxiliary operator questioned the impact of power lines running through the containment hatch opening on containment closure time.*

However, some problems* indicated that improvement in coordination and communication among other groups with plant operators was still warrante One example occurred when operators stroked closed containment sump outlet isolation valve CV-3029 while workers were setting up inspection equipment

• inside. the piping near this valv Management stopped work and regrou~ed to

• develop an adequate work plan for the pipe inspection.2 Limiting Conditions For Operation (LCO) Board Annex The inspectors reviewed the "Limiting Conditions For Operation (LCO) Board.

Annex published during the period of June 19 to June. 30, 199 This annex listed the inoperable, non-conforming, or degraded equipment; the condition when th~ equipment was required to be operable; the applicable work orders or corrective action documents; and the applicable Technical Specification (TS)

or administrative requiremen During a review of the annex the inspectors identified the following errors pertaining to when the equipment was required to be operable which were discussed with the Operation's Department Superintendent:

Page 8 of the June 21 LCO Board Annex stated that eight of the steam generator main steam relief valves were removed on June 1 The plant condition was specified as "prior to critical" per TS 3.1.7. Sirtce *

heatup without the main steam relief installed would have been impractical, the inspectors questioned the appropriateness of this plant conditior The crew referenced the correct lCO, but the inspectors questioned if an integrated system picture had been used when specifying the required plant condition.. *

On Ju~e 23.the inspectors reviewed the controlled copy of the TSs maintained in Doctiment Control Center (DCC) and noted that TS amendmen dated June 5 revised TS 3.1.7. This amendment required operability of the main steam reliefs pr1or to leaving cold shutdow The amendment was faxed to the site from NRR on June 6, posted in the DCC's TS on June 23, and distributed to the other departments for posting in copies of the TS; The irispectors questioned the appropriateness of the posting delay because, on June 15, the onshift crew made an operability determination for the main steam reliefs using out dated TSs.

. The correct plant condition was referenced after discussion with the Operations Department Superintenden *

Page 11 and 12 of the June 26 LCO Board Annex listed six entries for equipment associated with containment coolin The equipment included component cooling water heat exchangers, containment air coolers, -and containment spray pump The required plant condition was specified as

• "prior to critical" per TS 3.4.1. This was the correct plant condition per TS 3.4.1 but the licensee had implemented a more restrictive operability condition for TS 3.4.1 per standing order (SO) 5 In this case SO 54 required operability of the equipment prior to 300 degree SO 54 was nrit referenced for these items but was correctly referenced in other entrie *

The general listing ~f components on the LCO Board Annex was hard to follow because they were randomiy listed and not listed by in an specific order such as by systemi TS, or any other Jogical orde Since the plant was in an outage and out of service equipment was controlled

. by the outage schedule, the items discussed above appeared to have been administrative in nature. However, the inspectors questioned the usefulness of the LCO Board Annex due to the error In addition, checks to assure proper TSs or sos reference appeared to have been lackin.3 Licensee Plans For Coping With A Strike The licensee's company-wide union contract the Utility Workers of America expired at midnight on May 31, 199 The inspectors reviewed the licensee's strike contingency pla The plan described measures the licensee would have taken in the event of a strike to maintain satisfactory control over plant activities. The inspectors determined the plan was adequat The inspectors specifically verified that minimum licensed operator staffing would have been maintained in the event of the strike. Prior to the ~eadlirie, the lic~nsee and union reached a tentative agreemen The strike contingency was still in effect at the close of this inspection period because the union membership was

• scheduled to vote on ratification in mid-Jul.4 Results of Plant Tours Housekeeping in most areas of the auxiliary building was goo However,*

housekeeping weaknesses were noted in the east safeguards and component

cooling water room Multiple examples of accumulated debris on the floors and pools of water spreading from contaminated drains* into adjacent clean areas were note In the east safeguards room, the inspectors noted improper securing of an argon gas cylinder to some scaffoldin In the component water

. cooli~g room, the inspectors noted the improper securing of ladders and that several workers not wearing hard hat The inspectors reported these find1ngs to cognizant personnel, and the problems were corrected by the.end of the inspectio Radiological controls and postings were all within the regulatory requirement Housekeeping in most areas of containment was adequat Some areas showed specific weaknesse For example, there were examples of material inappropriately stored on the landings of stairway An unsecured (not tied*

off) ladder was propped against the wall adjacent to the reactor vessel water level tygon tube. There were some weaknesses in the control of foreign

material within the debris-free zone near the reactor cavit The inspect6r found broken pietes of glass and broken pieces of signboard within the designated debris-free are Further, the licensee found multiple examples of unauth.ori zed materi a 1 in the debris free zones for both the spent fue 1 and

. reactor cavit The licensee* was aware of the weakness and initiated condition reports to evaluate the issu Radiological controls and postings in the containment were all within regulatory requirements. * MAINTENANCE NRC Inspection Procedures 62703 and 61726 were used to perform an inspection of maintenance and testing activities. There was one Unresolved Ite identified pertaining to use of mobile crane The findings showed maintenance was adequat.1 Reactor Vessel Disassembly The inspectors reviewed several activities associated with disassembly of the reactor vessel, performed in accordance with work order 2441316 The licensee performed the activities using procedure C-PAL-RFM-001, "Palisades Refueling Manual."

2.1.1 Palisades Refueling Manual Section 9.2.10. "Head Rem6val" The licensee's performance during lifting and removing of the reactor vessel head was goo The pre-job briefing was conducted wel The Westinghouse refueling supervisor satisfactorily covered the details, discussed precauti-0ns, potentials problems, and contingencies. Management invol~ement was evident; standards and expectations were reinforce The actual lift was performed without incident. Communication and coordination were goo.1.2 Palisades Refueling Manual Section 9.2.13. "Removal of Upper Guide Structure from Reactor Vessel" The removal of upper guide structure from the reactor vessel was performed wel The activity was well planned and execute Good coordination and communication among the various groups involved was note The licensee

demonstrated good contractor ~ontrol. The licensee performed a thoroug~

search and found no attached fuel assemblie.1.3 Palisades Refueling Manual Section 9.2.16. "Removal and R~installation

of Core Support Barrel" Removal of the _core* barrel was well planned and executed.. The briefing package was excellent; the pre-job briefing was thorough and effective in relating radiological contro1s concern The workers were cognizant of their respect i ~e res pons i bil iti es, and personnel access to the work areas w~s effectively controlled. Refer to paragraph 4.1.1 for a more thorough discussion of radiological control practices associated with the removal of the core barre.1.4 Core Barrel and Upper Guide Structure (UGS) Interim installation per paragraphs 9.2:16.D and 9.3 of CPAL RFM 001 The inspectors attended the prejob briefing for both activities. The briefingi were an interactive exchange of information between supervisors and

• crew member Items discussed included applicable procedural steps, critical*

work positions and who would man them, stopping points, contingency actions, and the radiological work permi *

The inspectors observed a portion of the UGS installation and noted that several lessons learned from previous UGS removal problems (fuel bundle remained attached and was lifted with the UGS) were incorporat~d. For example, the load cell calibration was current, the levelness of the UGS was confirmed, a remote submarine was used to verify cleanliness of the UGS seating flange,* and the submarine was used to assure that fuel alignment pins were not damaged during the mov Previously, the fuel alignment pins were gauged to assure straightnes However, during discussion with the engineers, the inspectors were not sure if the pins would be gauged after the UGS was removed from the vessel to facilitate core reloa The gauging was not mandated by regulatory requirements but implemented by the licensee subsequent to the last time a fuel bundle remained attached to the UG The licensee was encouraged to evaluate the benefits of the gauging prior to the final setting of UG.

During the UGS cleanliness inspection a piece of broken glass was noted on the flange which was remove Apparently the glass was the remains 6f a light bulb that broke during a previous UGS inspectio.1.5 Palisades Refu~ling Manual Section 9.3.2. "Setting of Reactor Head on Shims" Personnel error on the part of the crew performing this evolution cost the licensee additional dose and caused the redirection of resources to perform emergent inspections of the reactor vessel head and upper guide structure component While lowering the reactor vessel head onto the reactor vessel, the licensee intended to lower the head to about two feet above the reactor vessel flange and to hold for ISi inspections on the flang Conununication and coordination between personnel. in the reactor cavity and the crane

operator bioke dow At thii point, the crane operator lowered the reactor vessel head onto the* reactor vessel.fl~nge prior to the performance of ISi inspection The licensee has planned various inspections and te~ts to check for damage to susceptible component.2. Other Maintenance Observations 2.2.1 Work Order 24415662, Removal Of Cables F~om The Reactor Vessel Head This work activity was another example of poor coordination and convnunication among work group This example occurred early in the refueling outage and was discussed in the previous inspection report (IR95007(DRP)).

Maintenance workers mistakenly disconnected four core exit thermocouples intended to be used as one backup indication for primary coolant temperature in the event of a loss of shutdown coolin The maintenance worker involved recognized an immediately reported his mistake to his supervisor and to plant operator The core exit thermocouples were not required to be operable per the Technical Specification~- and were reconnected shortly thereafter.. The licensee held an*

immediate corrective action review board with all involved personnel to discuss causes and corrective action.2.2 ~ork Order 24416029. Capacity Test Of Switch Yard Battery A personnel error associated with this activity occurred when a mobile Grove crane was driven.to the switchyard to assist in the lift of a battery load bank off the back of a truck. The crane driver inadvertently s_evered an overhead power line supplying lighting to the main parking lot and guard hous No perionnel injuries were reported, and no other equipment ~as reported to have been affecte The cause appeared to be due to inattention to detai Because a similar event had occurred during the previous inspection period when a mobile crane collided with an overhead *support structure, the inspectors considered the licensee's control of mobile cranes to be a weaknes The licensee took immediate action to investigate and to correct the ca'use of this recent even The 1 i censee agreed to respond within 60 days on causes and preventive actions for this even Pending review o the licensee's response, this was considered an unresolved item (50-255/95008-.*

01).

Activities associated with work order 24416287, Cleaning Of 1-2 DG Jacket Water And Lube Oil Coolers was observed with no concerns being identifie.0 ENGINEERING NRC Inspection Procedures 37551, 73051, 73052, and 73755 were used to perform an inspection.of engineering activities. The.findings showed pe~formance was goo Items which were "Closed" as a result of this inspection met the criteria established in the Inspection Procedure.1 Inservice Inspection (lnconel Alloy 600 components)

The licensee's Project Plan Alloy 600, Revision 1, represented a sound technical approach to managing primary water stress corrosion cracking (PWSCC)

• in I nc:one l A 11 oy 600 materials within the primary coo 1 ant system {PCS).

The nondestructive testing {NOE) techniques and acceptance criteria used were reasonable and con~istent with analysis and industry practice The NOE inspections completed this outage, albeit reduced in scope, still assured components in the most susceptible category, Group I, and the more susceptible components category, Group II, received ultrasonic {UT) or dye penetrant {PT)

and visual inspection (VT).

All identified Alloy 600 locations in the PCS received a visual inspectio~ as a minimum; Moreover, no PWSCC was found during inspections performed this outag.1.1 Program Review The Alloy 600 plan included as one of several goals, the development of an inspection program to* identify and to characterize PWSCC in Inconel Alloy 600 co~ponents. Toward this goal, the plan identified 251 Inconel Alloy components potentially susceptible to PWSCC in the plant. These components were grouped into three levels of inspection priority based on susceptibility to PWSCC, consequence of failure, detectability and ALARA consideration The original scope of planned insp~ctions was ~ubsequently reduced to save total project radiation dose expenditur Of the original 79 UT inspections planned, 27 were completed.* Of the original 66 PT inspections planned, 61 were complete The completed inspections covered all Inconel Alloy 600 compon~nts listed as the highest priority (group I) in the plan. Jhe licensee had not committed to perform inspections deleted for this outage ih the lower priority inspection categories listed in the original pla The following observations were not~d as weaknesses asso~iated with the planned inspection of lnconel Alloy 600 components:

Previously unidentified flow diverter plate in the pressurizer spray line prevented planned visual inspection of the internal weld surfaces on the pressurizer spray safe end;

The use of PT inspections on outside surfaces of components to locate PWSCC, ~hich would initiate from the interior surfaces of componentsi.

was at best redundant with UT or VT performe Dose expended on these PT inspections was not considered consistent with ALARA principles;

The underestimate of radiation dose expenditure associated with planned inspection activities, precipitated the significant scope reduction in actual completed inspection activitie The following observations.were noted as strengths associated with the planned inspection of lnconel Alloy 600 components:

The util izati_on of mockups to demonstrate and to qualify NOE techniques used in the inspections;

The use of "enhanced" UT techniques, which included angle beam (shear and longitudinal wave mode) search units, automated data collection and

an~lysis and scanning equipment custom designed arid built specifically f~r Palisades component.1.2 Observati~n of Work Activities Personnel from B&W Nuclear Technologies (BWNT) performed the inservfce

• inspection (ISi) of Inconel Alloy 600 components in accordance with Palisades'

Alloy 600 Project Plan and Alloy 600 specification revision The NRC inspectors observed work activities and had discussions with BWNT and licensee personnel during ISi activities. These observations included:

BWNT personnel performing UT on pressurizer spray line safe end and shutdown cooling outlet connection safe end welds;

BWNT personnel performing PT on the pressurizer surge line safe end welds, hot leg loop pressure piping safe end welds and hot leg drain

• line safe end welds;

BWNT personnel performing UT on the replacement power operated relief

. valve (PORV) safe end upper wel ~1~3 Procedure Review ISi procedures used in support of the Alloy 600 Project Plan were reviewed by the NRC inspector The ISi procedures were fourid to be acceptable and in

accordance with ASME section V, 1989 edition as modified by requirements of the Alloy 60~ Inspection Specification revision The inspectors reviewed qualifications and certifications of all BWNT personnel performing ISi, verifying conformance with licensees' Alloy 600 specification revision 0 *

requirement *

3.1.4 bata Review The UT and PT examination data ~eviewed was found to be in accordance with the applicable ISi procedures and the Alloy 600 Specification, Revision Radiographs of the upper PORV safe end weld were reviewed by the inspectors and found to be in conformance to ASME section V, 1986 addenda requirement Radiographic films taken for the lower PORV safe end weld could not meet code requirements for geometric unsharpness and 2T hole penetrameter image quality requirement The licen~ee stibsequently performed ultrasonic inip~ction of this weld to meet ASME section Ill, 1986 addenda requirements for volumetric examination.. No reportable indications were found for the UT inspections performed or for the radiographic inspections of the upper PORV safe end weld.

. Four comp.onents had reportable surface indications disclosed through PT testing. These indications were subsequently removed and follow up PTs verified no further surface indications to be presen.2 MOV Testing to Support GL 89-10 Commitments Work to support testing of MOVs was performed in a well coordinated manner due to effective pre-test briefings and good support from operations and

contracted *technicians. Operations Department's dedicated support in such items as breaker and valve line-ups also contributed to the considerable amount df testing completion~

Engineering Department demonstrated good safety insights to potential test problem Fo_r example, prior to performing test T-352, "HPSI Loop Isolation MOVs M0-3007, 3009, 3011, *3013 Differenti~l Pressure Test," the test coordinator anticipated a potential overthrust condition for M0-3009 becaus of the flow-over-the-seat valve desig The valve vendor was contacted and indicated that the valve weak link thrust limit of 15,151 pounds was conservative and a thrust up to 20,000 pounds for several hundred cycles was acceptabl During actual testing M0-3009 experienced approx 13,181 pounds of thrus *

3.2.1 Fuel Inspecti6n and Reconstittition per I-FC-942-01 The licensee performed an inspection of the fuel bundles that will be reused during the next fuel cycl The purpose of this inspection was to identity any cladding failures and to determine if there were any grid strap relaxation problems with fuel bundles used for vessel beltline weld shielding. Cladding

. integrity was confirmed by ultra-sonic inspection. Grid strap tension measurements required removal of the top plate and pull test on selected pin During the grid strap inspection, the operations manager observed the crew's recovery from a stuck too The manager alertly observed that a "red" mark lo~ated on the tool almost came out of the wate This mark was used to assure that sufficient water shielding was maintained between the crew and the attached* componen Apparently, the crew was narrowly focused on the recovery activities. It was watching the underwater camera and not monitoring the "big picture".

The manager's -Observation prompted a shutdown of this activit until the crew was briefed on the procedure and equipment.*

While observing the ultra-sonic inspection, the inspectors discussed the scope of the inspection with the fuel enginee The scope included approximately

330 bundle This number included the bundles scheduled to be reused for the next. fuel cycle and those selected for dry cask storage. During this conservation the inspectors was informed that the licensee's practice has been to confirm the integrity of the fuel bundles selected for dry cask* storage by performing the.mandated visual inspection and additional inspections such as ultra-sonic or sipping to confirm the integrity of the fuel pin cladding.*

3.2.2 Service Water Enhancement per I-FC-959-01 The enhancement required isolation of cooling water to the spent fuel pool (SFP) heat exchanger To provide SFP cooling, a temporary cooling system was installed. The inspectors reviewed the installation and operation procedure Once the system was installed, the inspectors observed system performance and confirmed that adequate SFP cooling was available and maintaine The i nsta 11 ati on and operating procedures addressed many contingencies, provided minimum operating temperatures and maximum SFP temperatures, specified when and how normal SFP cooling shall be reestablished, and ensured

. equipment reliability of the teniporary equipment through pre-certification operability run When the system was placed in service, SFP temperature remained essentially stabl. Follow-up on Non-Routine Event NRC Inspection Procedure 92904 was used to perform a review of written reports on non-routine event The following item was closed:

(Closed) LER 50-255/93009: Through wall cracking, caused by primary water stress corrosion cracking (PWSCC) in the heat affected zone of the PORV line to pressurizer Inconel Alloy 600 safe-end weld~ The cracking was circumferential and initiated from the inside diameter progressing *

intergranularly with the final 5-10 percent of crack growth ~being transgranula *

To prevent reoccurrence of this event, the licensee replaced the lnconel Alloy 600 pressurizer PORV safe-end with a type 316L stainless steel safe-end and used mechanical stress improvement techniques on select components to reduce susceptibility to PWSC In addition the licensee implemented an inspection of susceptible Alloy 600 components in the plant and performed a fracture mechanics analysis for Alloy 600 components to demonstr~te that they would fulfill their inservice lifetimes. This item is close * Follow-up on Previously Opened Item NRC Inspection Procedure 92904 was used to perform a ~eview of previously

• opened items (violations, unresolved items, and inspection follow-up items).

No problems were identified, and the following item was closed:

(Closed) Violation (50-255/94015-01):

Failure to take prompt and adequate corrective actions for the review of a 10 CFR Part 21 notice, and the

evaluation of completed MOV tests, which included overthrust and overtorque condition In response, the licensee developed an MOV Program Recovery Pla that established MOV program directives, provided prriper operability acceptance criteria for MOV test procedures, and prioritized available resource In addition, the licen~ee revised engineering manual procedure EM-28-01, 11Motor Operated Valve Program, 11 to clarify the evaluation of completed test dat The 10 CFR Part 21 notice was evaluated and results were.

incorporated in the MOV torque/thrust calculations.* This item* is close.0 PLANT SUPPORT (IPs 71750 and 83750)

Total outage dose, was approximately 135 rem (1.35 Sievert (Sv, with about 33 percent of the outage scope complete This total exceeded the dose goal by about 17 percent, largely owing to difficulties with th~ Alloy 600 project (section 4.2). Several poor radworker practices were identified during reviews of outage work (section 4.1), and a non-cited violation was issued for inadequate personnel monitoring (section 4.3). Overall, the radiation protection (RP) department performance was considered adequat.1 Removal of the Core Support Barrel and the Transfer of the Incore Detector Cask The inspectors observed two radiologically significant outage activities; the removal of the core support barrel and the transfer of the incore detector cas Although both projects used "lessons learned" from similar work, sevetal concerns were identified regarding radworker practi~e A lead shield was used to lower dose rates around the barrel to around 60-100 mrem (0.6-1.0 mSv) per hour (from> 1 rem (10 mSv) per hour).

During the moving of the barrel, several workers were observed loitering around the barrel, and the RP technician was moving frequently from behind the shield to inspect the barrel. Similar observations were made during the incore cask removal, where workers were observed loitering in area dose rates between 5-20 mrem (0.05 and 0,20 mSv) per hour. * ln neither case (core barrel or incore cask) were the workers' behavior challenged, suggesting a lack of

aggressiveness by the licensee in reducing individual dos * Alloy 600 Project The initial scope of the Alloy 600 project involved the examination of 79 primary loop penetrations using.dye-penetrant (PT) and ultrasound testing (UT) in an effort to detect primary water stress corrosion cracking (PWSCC) on the interior of the penetration Based on the vendor's estimate of the time needed to perform the examinations using.remote UT equipment, the plant developed an ALARA goal of 10 person-re However, from the beginning of the work, the plant experienced difficulties in meeting its ALARA goal for this jo Due to clearance constraints, the vendor was unable to use remotely o~erated * automatic UT equipment, and had to rely on manually operated equipment that required the operator to be in close proximity to the piping being examine Radiation levels near the piping were much higher than the levels expected for the use of remote equipmen The higher radiation levels, thus, increased the committed dose for the project. *

Aside from the inability to use remotely operated UT equipment, the erroneous use of the vendor's time estimate to complete the Alloy 600 ISi examinations made the largest impact on the licensee's ability to establish an accurate ALARA estimat The time estimate provided by the vendor was in clock-hours, rather than person-hours. This significant difference in time was not effectively communicated between the licensee and the vendo Although the vendor's "cl~ck-hour" estimate ~as a f~ctor of three different from the licensee's estimate of the time required to complete the examinations, the licensee did not question the reasoning behind the differenc The licensee accepted the vendor's time estimate, and believing the time was expressed in person-hours, used that time in developing the ALARA projection for the work packag * The lJcens~e documented the UT equipment and time estimate problems in an ln-progress Review report, and indicated that long term corrective actions and lessons learned would be included in a post job review planned for after the outag *

• External Exposure Control On June 20, 1995, an individual violated st~tion procedure by entering the radiologically controlled area (RCA) without an electronic dosimeter (ED).

A review. of selected licensee tondition Reports identified several other incidents involving EDs set on "pause" (i.e. not logged into the computer).

Station Protedur~ Number 7.04, revision 14, states, in part, that individuals shall 1 og in. on the. Management lnformat ion System (MIS) with a Secondary Dosimeter (ED) prior to entering the RC The procedure would have been violated if an individual entered the RCA with an ED that was improperly set on "pause~ or an individOal entered. the RCA without an E For each event, the workers were counseled by RP personnel and the respective supervistirs were notifie The access of the worker involved in the June 20th incident was revoked for an unspe~ified amount of tim The licensee was de~eloping a formal process for taking action against individuals who violate RP procedures and practices. Although the~e events, in the aggregate, were a violation of the. above station procedure, it was identified by the licensee and immediate corre~tive actions were take Ther~fore, it will not be cited as the criteria specified in Section VII.B.2 of the "General Statement of Policy and Procedures for NRC Enforcement Actions", (Enforcement Policy, 10 CFR P~rt 2, Appendix C), were me *

• PERSONS CONTACTED AND MANAGEMENT.MEETINGS The inspectors contacted various licensee operations, maintenance, engineering, and plant support personnel throughout the inspection perio Senitir personnel are listed belo At the conclusion of the inspection on July 6, 1995, the inspectors met-with licensee representatives (denoted by*) and summarized the scope arid findings of the inspection activities. The licensee did not identify any of the documents or processes reviewed by the inspectors are proprietar *R. A. Fenech, Vice President, Nuclear Operations
• T. J. Pa1misano, Plant General Manager
• K. P.. Powers, Engineering and Modifications Manager R. Swan~on, Director, NPAD
• D; W. Rogers, Operations Manager D. P. Fadel, Engineering Programs Manager
• J. P. Pomaranski, Deputy Maintenance Manager H. L. Linsinbigler, Project Management and Modifications Manage S. Y. Wawro, Planning Manager
• K. M. Haas, Safety & Licensing Manager
• R. B, Kasper, Maintenance Manager R. C. Miller, Deputy Engineering and Modifications Manager
• C. R. Ritt, Administrative Manager R. M. Rice, System Engineering Manager 15

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