IR 05000317/1985013

Insp Repts 50-317/85-13 & 50-318/85-11 on 850506-0617. Violation Noted:Unnecessary Initiation of Automatic Plant Protective Sys Due to Personnel Errors

ML20129C984
Person / Time
Site:	Calvert Cliffs
Issue date:	07/05/1985
From:	Elsasser T, Foley T, Trimble D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20129C797	List: IR 05000317/1985013 IR 05000506/2006017 ML20129C792
References
50-317-85-13, 50-318-85-11, NUDOCS 8507160389
Download: ML20129C984 (12)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket / Report: 50-317/85-13 License: OPR-53 50-318/85-11 DPR-69

Licensee: Baltimore Gas and Electric Company

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Facility: Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Inspection At: Lusby, Maryland Dates: May 6, 1985 - June 1 85 Inspectors: f 4 . Foley, e#

M [4 esident Inspector M YS/Is

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. C. Trim e R

-) ent h Inspector W ifs /85 Date Approved: -

> /tMV 7/#/85 T. C. Elsassgmief, Reactor Projects Section 3C Date Summary: May 5-June 17, 1985, Inspection Report 50-317/85-13, 50-318/85-1 Areas Inspected: Routine resident inspection of the Control Room, accessible parts of plant structures, plant operations, radiation protection, physical security, fire protection, plant operating records, maintenance, surveillance, open items,-

and reports to the NR Total Inspection Hours 12 Results: A significant portion of this inspection was dedicated to followup in-spection of licensee actions with regard to diesel generator (DG) interpolar con-necting bar and pressurizer spray valve fastener problems. In both cases, the licensee demonstrated strong initiative and took conservative actions in resolving safety concerns. Some weaknesses were, however noted in the identification of cracks in and incomplete removal of DG interpolar connecting bar stubs. A licensee Quality Assurance audit is in progress which should better define these weaknesse The spray valve fastener problem showed inadequacies in maintenance personnel training in bolting practices and in administrative controls of bolt tensioning

. evolution One violation was identified regarding unnecessary initiation of automatic plant protective systems (Recirculation Actuation Signal on Unit 2 aad loss of shutdown cooling on Unit I due to closure of a return isolation valve) due to personnel

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. DETAILS

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Within this report period, interviews and discussions'were conducted with various licensee personnel, including reactor operators, maintenance and su'rveillance technicians and the licensee's manageaient staf . Summary of Facility Activities Unit 1 remained shutdown throughout the inspection period in an extended re-fueling outage. Startup, originally scheduled for the first week in June, was delayed due to the discovery on May 22 of low resistance readings on the main generator A phase stator windings. The projected staptt4 date is August 15, 198 j Unit 2 began the period operating at full power. Thefunit was shutdown from May 17 to May 22 to inspect and replace pressurizer spray valve fastener On May 23, with the unit at 100% power, an inadvertent Recirculation Actuation Signal (RAS) was generated due to technician error on performance of a sur-veillance test. The' actuation-cid not induce any transients of a safety or operational concer ,-

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On May 14 a'n interpolar connecting bar in the generator ' of #11. Diesel Genera-tor (DG) broke free within the generator and caused significant- damage to the stator windings. Further investigation revealed a large number of fatigue induced cracks in the interpolar connecting bars of all three'0G's. An engi-neering evaluation showe'd that the connecting bars were unnecessary and all bars were removed from #12 and #21 diesel generators. The generator for #11 DG was replaced with a new generator with connecting bar,s installed. Those bars will be removed at a later dat ,

'i Threeinspectionswereconductedb9regionalspecialistsinthefollowing subject areas: Integrated Leak Rate Testing (Unit 1), environmental monitor-ing, and DG interpolar connecting bar fatigue crackin . Licensee Action on Previous Inspection Findings (Closed) Inspector Follow Item (317/84-18-02) Install Signs <n Safety Related Outside Tanks Warning That Vents Should Not Be Obstructe The inspector confirmed that the subject signs had been installe (Closed) Unresolved Item (317/84-23-01) Review of Temporary Plant Modifica-tions for Unreviewed Safety Questions. As discussed in Inspection Report 50-317/85-09, 50-318/85-09, Section 3 (Item 317/84-18-01) on April 4, 1985 ;

the licensee made a. major revision to CCI 117, " Temporary Mechanical Device, i Electrical Jumper, and Lifted Wire Control", requiring appropriate review /

evaluation of temporary modifications. This item is close '

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(Closed) Violation (317/84-23-02) Failure to Follow the Requirements of CCI

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~ 117D, " Temporary Mechanical Device, Electrical Jumper and Lifted Wire Control".

The subject event was the improper installation of a blank flange on the let-down line and subsequent overpressurization of a section of that piping. The requirements of CCI 117 and importance of good communications between the

[' maintenance and operations groups were discussed with maintenance and opera-tions personnel. Additionally, improvements were made to CCI 117D to require inclusion of a sketch or marked up drawing showing proposed locations of any blank flanges. This item is close (Closed) PAS Item (317/82-01-05) No Designation of Corrective Action and Other Records to be Reviewed by the Off Site Safety Review Committee (OSSRC). The inspector reviewed the OSSRC Manual, Technical Specifications 6.5.2.7 (re- j quired OSSRC reviews) and 6.5.2.8.1 (required OSSRC audits). The committee l

.has broad review responsibilities which include corrective actions (e.g. Non-conformance Reports)._ A large number of other records are required to be re'-

viewed (safety evaluations, LERs, Technical Specification Changes, violations, POSRC minutes, audit reports, etc.). Various audits conducted under the cog-nizance of the OSSRC encompass review of records applicable to the subject are This item is close (Closed) Violation-(317/83-26-03) Failure to Fully Implement a Surveillance Test Procedure (STP). This item concerned-an error in STP 0-7-1, " Engineered Safety Features, Logic and Performance Test", in that' direction was not given to return the service water inlet valve to #11 Spent Fuel Pool heat exchanger to its pretest conditio STP 0-7 has been revised to correct.the proble All operational STP's were reviewed by senior licensed personnel to ensure that similar conditions did not exist. The violation also concerned a failure of an operator to carry out a step of STP 0-7. The individual was counseled by the General Supervisor, Operations on the need for proper procedure im-plementation. The individual was required to draft an incident report which was distributed as required reading for all operations personnel. This item is close (Closed) Unresolved Item (318/82-27-04) Confirm Number of Letdown Isolation Cycles Do Not Exceed Maximum Allowabl The licensee completed an engineering evaluation to determine the number of thermal transient cycles allowable for the CVCS (Chemical and Volume Control System) nonregenerative heat exchanger, charging piping and charging nozzles. Of these, the most restrictive compo-nents were found to be the charging nozzles (180 cycles allowable). The in-

-spector confirmed with the licensee that the numbers of actual cycles experi-enced were less than the allowable. This item is close . Review of Plant Operations I Daily Inspection i

During routine facility tours, the following were checked: manning, ac-cess control, adherence to procedures and LCO's, instrumentation, recor-der traces, protective systems, control rod positions, Containment tem-L' __ _ _ _ _ _

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i perature and pressure, control room annunciators, radiation monitors, I radiation monitoring, emergency power source operability, control room j logs, shift supervisor logs, tagout logs, and operating orders.

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' System Alignment Inspection Operating confirmation was made of selected piping system trains. Ac-cessible valve positions and status were examined. Power supply and breaker alignment was checked. Visual inspection of major components L were performed. Operability of instruments essential to system perfor-mance was assessed. The following systems were checked:

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- Unit 1 Shutdown Cooling checked on May 30, 198 Service Water Alignment to #11 Diesel Generator checked on May 31, 198 Unit 1 Service Water in Service Water Pump Room checked on June 11, 1985.

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No' violations were identifie Biweekly and Other Inspections During plant tours, the inspector observed shift turnovers; boric acid tank samples and tank levels were compared to the Technical Specifica-tions; and the use of radiation work permits and Health Physics proce-dures were reviewed. Plant' housekeeping and cleanliness were evaluate Verification of the following tagout indicated the action was properly conducte Tagout 166691, SRW to #12 Diesel Generator checked on May 31, 198 No violations were identifie Other-Checks On May 14, 1985, Diesel Generator (DG) #11 was damaged when an interpolar connecting bar on the generator rotor broke free at one end, rubbed and abraded the stator winding and then completely separated by fractur The insulation on one stator winding was substantially damaged. The DG was undergoing over speed testing at the time of failure. The DG engine is manufactured by Colt Industries, Fairbanks Morse Power Systems Divi-sion, Model 38TD8-1/ The generator is a Fairbanks Morse, Typ.e TGZD The interconnecting bars are copper and are a part of the amortisseur, or damping, winding. They are attached between rotor poles. The winding serves as a speed and voltage stabilizer and would be most useful in situations where a DG is required to run in parallel with other machines to dampen speed changes or if the generator is required to carry single phase loads that can cause the three phases to become imbalance '

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Calvert Cliffs has three (3) EDG's (#11, #12 and #21). Each generator has sixteen (16) interpolar connecting bars; eight (8) bars on each en The seven bars remaining on the damaged end of #11 EDG were radiographed and six were found to contain cracks. All sixteen bars on #21 EDG were radiographed. Eleven bars had cracks and six bars were cracked completely through. A visual inspection of #12 EDG bars also revealed crack Metallurgical analysis of the failed bar determined the predominant cause of failure was high stress due to cyclic centrifugal loadin This problem appeared to be of a design nature and not due to a material de-fect. The licensee estimated that it took approximately 400-500 start cycles for the cracks in the failed connecting bar to initiate and an additional 500 cycles for failure. The licensee replaced the generator for DG #11 with a new generator (with similar interconnecting bars).

The connecting bars were removed from DGs #12 and #2 The licensee plans to remove the bars from DG #11 on an, as yet, undetermined schedul During a followup ' inspection on June 6,1985, a region-based NRC inspec-tor noted that several connecting bar stubs had been left attached to rotor poles on #12 and #21 DG's during bar removal. That inspector also noted, in a review of existing radiographs, indications of cracks near rotor poles. He expressed concern to the Plant Superintendent that cracks may be present in connecting bar stubs still attached to generator poles. The licensee did further radiographic examinations of the stubs and confirmed that several cracks were present. All remaining stubs were removed from the generators of #12 and #21 DG' The Facility Change Request, FCR 85-1025, which was used to remove the interpolar connecting bars specified that the cutting should be done "as close to pole face as possible". If this instruction had been closely followed, the later stub problem would never have arisen. The inspector questioned the General Supervisor, Operations Quality Assurance (GS-0QA)

why several stubs had been left in the generator during the initial cut-ting, particularly given the above instruction. The GS-0QA stated he also had several. questions about the planning and conduct of the bar re-moval and that he would initiate a special audit of this jo The in-spector will review this audit when complete CFR 21 reports were submitted to the NRC on May 29, and May 21, 1985 by Louis Allis Co. (current manufacturer of these generators) and June 5, 1985 by Colt Industries (former owner of the generator manufacturing firm). The June 5 report recommended removal of the interpolar connec-tor In general, the licensee's actions in response to this issue were prompt and conservative. ,

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6 Checks for Fatigue Cracking in the Salt Water System

'During the period, the inspector noted two potential sources of corrosion

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' fatigue problems in the Salt Water (SW) system and asked the Acting Plant Superintendent and. Principle Engineer, Plant Engineering Nuclear, to-evaluate whether or. not periodic checks for fatigue cracking should be conducted for the subject components. The endurance limits of ferrous materials can be lower in salt water. environments. Therefore, the pos-sibility of fatigue problems may be greater in the SW syste The disc in each butterfly valve of the SW system is fastened to its operating stem with*a tapered pin. The plant has experienced a failure of one of those pins and a rapid closure of the valve ensued. The in-ternals of butterfly valves _ located in headers shared by both SW loops, where a single valve failure could interrupt both trains of salt water flow,'were.later removed. However, the possibility of pin failures in remaining butterfly valves still exists, particularly in those~ valves that are exposed to greater turbulent flow due to their normally throt- ,

tied position. No periodic check of the pins is presently conducte The inspector noted the presence _of general corrosion and apparent pit-ting on exposed portions ~(outside the pump casing) on the carbon steel shaft on #23 SW pump. _ It is possible that similar corrosion ~ is-present

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.on the shaft internal to the pump casing. Since-(1) pits can act a sur-face stress concentrators and sites for initiation of cracks under cyclic stress and (2) these pump shafts have two additional stress concentrators-(shaft diameter change and a keyway) inside the pump casing, the inspec-

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tor felt the need for_a periodic check of SW pump shafts for fatigue cracks should be evaluated. No.such-checks are conducted at this tim The Principle Engineer, Plant Engineering Nuclear, indicated that conduct of these checks would be considered and could perhaps be done during routine preventative maintenance of the SW pumps'and when-SW valves are removed from the system for other maintenance' action ' Precautions to Prevent AFW Pump Steam Binding i:

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-During the period, the inspector reviewed licensee controls to prevent steam binding of Auxiliary Feedwater (AFW) Pumps. .Calvert Cliffs had previously experienced back-leakage from the steam generators. In re-L -

sponse they added an additional check valve in the steam driven'and motor driven pump discharge lines. These check valves are of a " tilted" disc design which provides for seating under low differential pressure condi-

tions. The licensee has had no recent check valve back-leakage ' problems.

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As a precautionary measure, weekly pyrometer readings are taken of AFW

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pump casings (PM 2-36-4-0-W). If temperatures exceed 10 degrees F above

ambient conditions, an. investigation for back-leakage is initiated.

. Operating Instruction 01-32, AFW System (Revision 22 for Unit 1 and Re-

[ vision 20 for Unit 2), contains a general precaution to " ensure AFW pump

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is vented properly if steam binding of the pump is evident before running the pump". No unacceptable conditions were identified by the inspector.

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. Events' Requiring Prompt Notification The circumstances surrounding the following events requiring prompt NRC noti-fication pursuant to 10 CFR 50.72 were reviewe ,

' Inadvertent Recirculation Actuation System (RAS) Actuation and Loss Of Shutdown Cooling

'On May 23, 1985, during performance of Surveillance Test Procedure (STP) M-

'220-2, ESFAS-Functional-Test, on Unit 2, instrument technicians encountered difficulty in testing a Refueling Water Tank (RWT) level switch. -At the time, the channel under test was in a trip or actuation state. The technicians

then, without first conferring with their' acting supervisor, removed the cover from a secondLlevel switch to make a physical comparison between switche Removal of the cove'r caused the actuation of the second channel resulting in RAS actuation. .The unit was at power operation at the time. No significant problems resulted from the actuation (i.e., Containment sump valves opened but due to the presence of in-line check valves, no water flowed to the sump from the RWT; Low Pressure Safety ' Injection Pumps were not in operation and

the trip signal sent by RAS, therefore, had no effect on them). The test was later. repeated by more senior technicians and all level switches (2-LS-4142A-D)

were found to be working properly. The inspector discussed the event.with-

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the technicians' Assistant General Supervisor. The technicians ~-failed to follow instructions given by their acting supervisor, regarding notifying him so he could be present for the test, and did not follow the directions and two general precautions in the ST Section V of_the STP requires removal of each RWT level switch channel, one at a. time, for testing. General Precaution B-emphasizes that only one channel is to be affected by the test ("be in a test mode") at any one tim General

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Precaution A directs technicians to " Proceed through each section of this

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(procedure) in sequential steps." If expected results are not obtained in a step, do not proceed with the testing of the affected channel until-the L required test results are obtained by applicable corrective actio At 10:45 a.m. on June 2, 1985, with Unit 1 in Mode 5, shutdown cooling was lost due to operator error. Reactor Coolant System (RCS) pressure was allowed H to increase to about 280 psia and as designed, automatic closure of a shutdown cooling suction isolation valve (1-MOV-652) occurred. The second pump used

, for shutdown cooling (Low Pressure Safety Injection Pump) was out of service l for maintenance. RCS pressure was quickly reduced by charging pump auxiliary l

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spray to the pressurizer, and shutdown cooling was restored at 11:08 Coincident with this event, the plant computer, which normally provides addi--

l .tional pressure indication and an alarm to warn operators that RCS pressure

'is approaching the 280 pound range, was inoperable (had failed earlier that

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day). Additionally, an annunciator warning of low shutdown cooling pump suc-l tion pressure did not properly alarm after the suction valve closed. The

plant was operating under Operating Procedure OP-1, Plant Startup from Cold Shutdown, Revision 26, at the-time. That procedure includes two precautions,

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under. General Precaution IE and Initial Condition IIA.4, stating that whenever the shutdown cooling system is in operation, RCS pressure shall not exceed 270_ psi The" licensee could not explain why the annunciator had not functioned. Its pressure' switch was-later bled down and annunciation occurred in the expected

. range of pressure (as measured by an indicator associated with the switch).

The. inspector recommended that the switch's calibration be checked. The lic-

.ensee stated this would be don ,

The failure to follow procedural guidance was the principal cause of the above events (RAS actuation and loss of shutdown cooling). Failure to follow ap-propriate procedures'is a violation (317/85-13-01).

Control Room Ventilation HEPA Filters During a surveillance test of both-trains.of the post accident Control Room (Loss of Cooling-Incident) ventilation system, both trains of HEPA filters i failed pressure-drop, flow rate and D.0.P. tests. The filters were replaced and a-Licensee. Event Report (LER) was submitted. The filters were replaced which corrected the efficiency and pressure drop problems. The flow rate problem was resolved by a damper. adjustment. The licensee attributed the P inability to pass the D.O.P. and pressure drop tests to the accumulation o dust / dirt of the . filters' due to large scale building renovation activities in the room. The flow rate will be checked monthly for two months and again l 'six months later to assure the damper adjustments were adequate.

I 6. ' Observations of imy>1 cal -Security

L Checks were made to determine whether security conditions met regulatory re-L quirements, the physical security plan, and approved procedures. Those checks l included security staffing,Lprotected and vital area barriers, vehicle l: searches and personnel identification, access' control, badging, and compen--

satory measures'when required.

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No violations were identifie : Review of Licensee Event Reports (LER's)

l-E LERs submitted to RI were~ reviewed to verify that the details were clearly Lreported, including accuracy of the ' description of cause and

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adequacy of corrective action. The inspector determined.whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup. The following LER's were reviewe ]

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LER N Event Date- Report Date Subject

~ Unit 1 85-03 04/06/85 05/03/85 Shutdown for Refueling Outage

'85-04 04/06/85 05/02/85 ESFAS Occurred During Surveil-lance Testing 85-05 04/06/85 04/29/85 Inadvertent Initiation of Steam Generator Isolation 85-06 04/18/85 05/16/85 UGS Removal Without Fuel Handling Supervisor Present 85-07- 04/15/85 05/15/85 'HPSI Injection Legs' Flow s Imbalanced Unit 2 85-01: 04/25/85 05/23/85 . Manual Trip Caused by Degradation-of 21A Reactor Coolant Pump Shaft Seal

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~85-04* 05/18/85 06/13/86 Control. Room Post LOCI Filter Sy' stem Inoperable 85-07 04/15/85 05/15/85 HPSI Injection Legs' Flow Imbalanced

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• See.Section 5 for further detail . Plant Maintenance The inspector observed and reviewed maintenance and problem investigation ac-tivities to verify compliance with regulations, administrative and maintenance

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procedures,-codes and standards, proper QA/QC involvement, safety tag use,.

equipment alignment,-jumper use, personnel qualifications, radiological con-trols for worker protection, fire protection,-retest requirements, and re-'

. portability per Technical Specif.ications. The following activities were in--

,cluded.

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M0 205-151-300A, #12. Diesel Generator MCC Bus Bar Inspection observed

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.on May 31, 198 MO '203-003-257A, Tack weld of bolts to lube oil flange for #12 Diesel Generator observed on May 31,-1985.

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PM 1-24-F-R-12, #12 Diesel Generator Control Relays observed on May 31, 198 Inspection of #21 LPSI pump coupling performed on June 11, 198 No violations were identifie . Surveillance Testing The inspector observed parts of one test to assess performance in accordance with approved procedures and LCO's, test results (if completed), removal and restoration of equipment, and deficiency review and resolution. The following test was observed:

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STP 0-5, Unit 2 Auxiliary Feedwater Pumps-test observed on June 13, 198 The inspector reviewed documentation for the following completed Surveillance Tests:

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STP M-2-1, Pressurizer Safety Valve setpoint check for 1-RV-200 and 201 reviewed on June 12, 198 STP M-8-1, Component Cooling Relief Valve RV-3827 setpoint checked re-viewed on June 12, 198 STP M-11-1, Unit 1 Snubber Functional Test reviewed on June 12, 198 No violations were identifie . Radiological Controls Radiological controls were observed on a routine basis during the reporting period. Standard industry radiological work practices, conformance to radio-logical control procedures and 10CFR Part 20 requirements were observe No violations were identifie ,

11. Pressurizer Spray Valve Fasteners On May 1,~1985, during a Unit 2 outage, the licensee found that one of eight body to bonnet stud fasteners was broken on pressurizer spray valve 2-CV-100 The remaining fasteners on that valve were ultrasonically tested (UT) and no indications of cracks were found. The broken fastener was replaced with the specified 17-4 PH (ASTM A564, type 630) stainless steel material. A decision was then made to check the condition of the fasteners on the Unit 1 spray valves 1-CV-100E and 100F. Three bolts in each of these valves were rejected (by UT)) du'e to cracks or, in one case, due to fracture (a stud on valve 1-CV-100E). The. rejected studs for 1-CV-100E were further examine Two had indications of stress corrosion cracking (SCC) and the fractured bolt was apparently made of the wrong material. On May 17 and 18, the licensee reduced ;

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power on Unit 2 and entered Containment to verify that the Unit 2 spray valve fasteners were made of the correct material (by magnet and eddy current tests).

The licensee's decision to do this was also in part based upon a discovery

.that on May 3,1985, a mechanic had tightened and very possibly over-torqued the body-to bonnet bolts on 2-CV-100E to stop a leak. No torque specification had been provided to the mechanic. The checks on Unit 2 showed that 3 fast-eners on each valve were made of the proper material and 5 on each were of,

-improper material-(type 316 stainless steel which has a lower yield strength than 17-4 PH). One crack was found in a 17-4 PH fastener on 1-CV-100 Due to an unavailability of 17-4 PH material, the licensee replaced the 316-stainless steel bolts and the cracked 17-4 PH bolt with A193 type B7 carbon steel bolts (Facility Change Request 85-28) and resumed power operation. They

. intend to replace the carbon steel bolts with 17-4 PH bolts at the next re-fueling outage (Fall 1985). A193 material is used in a number of fasteners in primary system valve No indication in plant maintenance history could be found that would indicate that any pressurizer spray valve studs had been replaced. Therefore, the valves could have been received from the vendor-(ITT Hameldahl for all four-valves) with-improper fasteners installed. A review of all other ITT valves was conducted to verify correct stud material. Incorrect material (was sup-posed to be 17-4 pH) was found on 3 of 4 pressurizer spray bypass valve These studs were replaced by studs of an approved material. A documentation review was undertaken to determine if valves supplied by two other major ven-dors specified 17-4 pH studs. No such valves were identified as being in-stalled in either plant in other than non-critical application I&E Bulletin 82-02, Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants, discussed industry problems with SCC and emphasized the'need for maintenance crew training on proper bolting practice In their bulletin response, the licensee indicated administrative and main-

'tenance procedures had been reviewed and updated where necessary to ensure training in proper bolting / stud practices, detensioning and retensioning practices and gasket installation and controls. However, the mechanic who tightened the fasteners on Unit 2 valve 1-CV-100E was apparently not aware of the problems that can result from over-torquing. This coupled with the fact that no torque specifications were given to the mechanic indicates that the licensee's training / controls are wea Through discussions with licensee personnel, the inspector learned that torque specifications are not always provided by vendors. An industry group is try-ing to better define threshold torque valves below which SCC should not occu The Plant Superintendent stated that he was very concerned about the apparent lack of maintenance personnel knowledge / training on bolting practices and felt a significant weakness was expose The General Supervisor, Maintenance and Modifications, discussed the event with maintenance personnel and instructed them to not torque fasteners in any pressure boundary without first ensuring that appropriate torque specifica-

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tions are given (either by Technical Manual or from the engineering group).

The licensee is currently scheduling a training program for maintenance per-sonnel-on bolting practice Identification of proper torque specifications for critical fasteners, estab-lishment of controls to ensure those torque values will be included in main-tenance procedures, and completion of proper training for maintenance person-nel will be reviewed by the NRC during routine inspection of maintenance ac-

'tivitie . Exit Interview Meetings were periodically held with senior facility management to discuss the inspection scope and findings. A summary of findings was presented to the licensee at the end of the inspectio .