IR 05000245/1986003
| ML20197C121 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/05/1986 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20197C101 | List: |
| References | |
| 50-245-86-03, 50-245-86-3, 50-336-86-03, 50-336-86-3, NUDOCS 8605130236 | |
| Download: ML20197C121 (14) | |
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U.S. NUCLEAR REGULATORY COMMISSION l-
REGION I
-Report:
50-245/86-03; 50-336/86-03 Docket Nos:
50-245/50-336 License Nos.
OPR-21; DPR-65 Licensee:
Northeast Nuclear Energy Company Facility:
Millstone Nuclear PoweV Station,'Waterford, Connecticut Inspection at: Millstone Units 1 & 2 Dates:
February 25, 1986 through April 7, 1986 Inspectors:
Theodore A. Rebelewski, Senior Resident Inspector Geoffrey E. Grant, Resident Inspector Approved:
N bd,b sIslg4 E. C. McCabe, Chief, Reactor Projects Section 3B Date Summary:
50-245/86-03; 50-336/86-03 (February 25 to April 7, 1986)
Areas Inspected: This ' inspection included routine NRC resident inspection.(192 hours0.00222 days <br />0.0533 hours <br />3.174603e-4 weeks <br />7.3056e-5 months <br />) of plant operational verification, observation of surveillance and mainten-ance activities, ten year containment tendon surveillance (Unit 2), shroud head bolt crack review (Unit 1), fuel rerack program (Unit 2), isolation condenser valve anomaly (Unit.1) and a meeting with NRR and utility on the Integrated Safety As-sessment Program.
Results: An unresolved issue was identified at Unit 1 involving failure to submit a trouble report when an isolation valve in the isolation condenser system was shut by hand. On its next test, the valve failed to open automatically.
The licensee subsequently attributed the auto-actuation failure to excessive torque during the
~ closure by hand, which made the automatic actuation feature of the isolation con-denser system inoperable from January 8 until March 26, 1986.
The absence of a.
trouble report in this case represented a lack of input for management evaluation.
Otherwise, no unacceptable conditions were found.
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8605130236 860505 PDR ADOCK 05000245 l
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l TABLE OF CONTENTS Page 1.
S umma ry o f Fac i l i ty Ac t i v i ti e s........................................
2.
Plant Operational Safety Veri fication.................................
3.
Pressurizer Head Vent Isolation (Unit 2)..............................
4.
Tenth Year Containment Tendon Surveillance (Unit 2)...................
5.
Observation of Surveillance (Unit 1 and 2)............................
6.
Shroud Head Bolt Cracks (Unit 1).....................................
7.
Fuel Rerack Program (Unit 2)..........................................
8.
Isolation Condenser Valve Anomalie (Unit 1)...........................
9.
Maintenance (Unit 1 & 2)..............................................
10. Integrated Safety Assessment Meeting (Unit 1).........................
11. Unresolved and Other Items Involving Licensee Action..................
12. Management Meetings...................................................
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DETAILS 1.
Summary of Facility Activities Unit 1:
Unit 1 operated at 100% power during this report period, with minor reductions.
in power for surveillance testing.
Unit 2:
The plant operated at full power during the inspection period.
Reductions in power were made to accomplish surveillance testing.
Identification of pressurizer head vent valve solenoid leakage resulted in isolating the vent path (Paragraph 3).
2.
Plant Operation Safety Verification The inspectors observed and verified facility conditions in order to ascertain whether the licensee is operating the facility safely and in conformance with regulatory requirements.
Interviews and discussion with plant personnel, tours, observation of surveillance and maintenance work items, and an inde-pendent verification of selected safety system status were the bases for this inspection.
2.1 Control Room Activities (Unit 1 & 2)
The inspectors observed control room activities during daily visits.
Various shift crew changes were observed. At Unit 2, an indication of faulty reactor seal water flow to Reactor Coolant Pump No. 2 was observed on the back shift.
Operator response was professional.
The Shift Con-trol Operator and Shift Supervisor, who were both near the control board, provided the needed procedures for coping with this problem.
(In this case, the seal flow indication fluctuated for a short period and returned to normal by itself.
The cause was attributed to transient foreign ma-terial.)
The control room annunciators had a minimal number of alarms. The in-spector confirmed that operators were knowledgable of the cause of each.
Observations were made of instrumentation versus recorder operation parameters.
Only minor deviations were identified.
Unit Superintendents were observed to routinely be in the control room discussing activities with the shift supervisor and other members of the operating staff.
The inspector reviewed operation logs, night orders, bypass jumper log books, and daily reactor coolant leak rate data.
No inadequacies were identified.
Indication of increasing leak rate were addressed by the licensee and reviewed by the inspector as indicated in Detail 3 of this report.
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2.2 Operability of Engineered Safeguards Features Unit 1-The inspectors verified a number of valve positions, instrumenta-tion and electric power sources for the "B" Low Pressure Injection System.
No anomalies were identified.
Unit 2-The inspectors verified the lineups of the "B" Low Pressure In-jection System and valving and instrumentation requirements for two Motor-Driven and the Steam-Driven Auxiliary Feed Water System pumps.
No anomalies were identified.
2. 3 Tours of Facilities (Unit 1 & 2)
Tours of accessible areas, including exterior areas of the facilities, were routinely performed.
An assessment of conditions of equipment, radiological controls, security and safety were made.
Both units were notably absent of packing leaks. A minirum of waste material in properly placed receptacles was observed.
No fire hazards were observed.
2.4 Radiation Protection Controls The inspectors verified that numerous radiation surveys were made during Unit 2's Spent Fuel Pool Rerack Program (Paragraph 7).
No anomalies were identified.
Posting of High Radiation areas was reviewed and found satisfactory.
2.5 Additional Inspection Items 2.t.1 Problem Identification Reports (PIR)
The inspectors reviewed the Unit 1 PIR's, PR 1-86-11 thru PR 1-86-18.
The majority of items were problems encountered dur-ing preventive maintenance and/or Surveillance testing.
No unacceptable conditions were identified.
2.5.2 Locked Valves During the witnessing of a surveillance on a Unit 2 LPSI pump, the inspector noted that a number of locked valves were chained and locked in position, sut were secured with slack in the chains.
The technician in charge was asked to try to remove the chain from the locking posts.
He readily removed the locked chain such that the valve could be repositioned and the chain returned to what appeared to be a locked state.
The position of these valves was reverified by the licensee and the chains were properly placed to secure the valves.
This was reported to licensee management.
Both Unit 1 and 2 sur-veyed their locked valves for similar locked valve problems.
The methods used to secure locks were identified to operating shifts.
No additional locked valve problems were identifie.
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2.5.3 Snubber Problem The licensee identified that snubber MSS-6, a Pacific Scien-tific Mechanical Snubber on the Isolation Condenser System, was detached from its mounting.
The licensee's investigations indicated that minimum thread engagement was the cause.
This snubber had recently received Wylie Laboratory testing and had been reinstalled.
A review of drawings and observation of a similar snubber by the inspector indicated that up to one inch of insertion of the clevis to snubber attachment was available.
The procedures for plant installation of this and similar types have been revised to preclude similar problems.
The inspector had no further questions.
2.6 Plant Operations Review Committee (PORC)
The inspector attended the following Plant Operations Review Committee meetings.
Unit I - 2/25, 2/26, 3/31, 4/1
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Unit II - 3/5, 3/6, 3/11, 3/17, 3/19, and 4/4.
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The Technical Specification (TS) requirements for member attendance was met. Agenda subjects included procedure changes, design change packages, and proposed TS changes.
The subcommittees that were assigned design changes presented their findings to the committee.
Discussions were frank.
Comments and probing questions were provided on a number of is-sues.
Improvements to ensure radiological safety and clear precise pro-cedures were also discussed.
Not all items presented were approved.
No unacceptable conditions were identified.
2.7 Summary The inspector found that event resolutions, equipment, materials, proce-dures and personnel observed were acceptable.
3.
Pressurizer Head Vent Isolation-Unit 2 Reference: Technical Specification 3.4.6.2 The TS limit on unidentified Reactor Coolant System (RCS) leakage is 1.0 gpm.
In mid-February, RCS unidentified leakage began to gradually increase from a base level of approximately 0.15 gpm.
On March 2, 1986, the leak rate jumped from 0.24 gpm to 0.34 gpm.
It continued to gradually increase to 0.45 gpm and on March 15 it jumped to 0.52 gpm.
On March 10, the licensee made a containment entry at full power to inspect for the unidentified leakage.
Two possible sources were identified.
The first, from a charging line check valve body-to-bonnet leak, was determined to be of minimal consequence.
The second, from a sparger common to both the pressurizer and reactor head vent
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systems, appeared to be the majcr contributor to the leakage.
Through a com-bination of indications, including temperature sensing elements, the licensee determined that the leakage source was through the solenoid-operated isolation valves for the pressurizer head vent.
During the week following identification of the leakage source, the licensee analyzed the short and long term effects of isolating the pressurizer head vent system.
By March 16, the leak rate was 0.64 gpm.
On March 18, with the rate at 0.72 gpm, the licensee made a containment entry to place a leakage collection apparatus under the sparger.
The apparatus yielded a 0.1 gpm col-lection rate, but was not collecting steam leakage.
On March 21, the licensee concluded that isolating the pressurizer head vent was an acceptable action and prepared for a containment entry.
The resident inspectors observed these preparations and questioned the adequacy of relying solely on oral briefings of personnel on the activity sequence and precautions.
The licensee subse-quently developed a special procedure covering all aspects of the entry and leak isolation.
This resolved the inspectors concerns.
On March 21, the resident inspector observed the containment entry and leak isolation evolution.
The pressurizer head vent was isolated by shutting Pressurizer Head Vent Isolation Valve 2-RC-239.
Subsequent unidentified leak rate computations have consistently yielded values less than 0.1 gpm.
The inspector had no further questions on this item.
4.
Ten-Year Containment Tendon Surveillance-Unit 2 References:
a.
Technical Specifications (TSs) 4.6.1.6.1.A, 4.6.1.6.l.B.
b.
USNRC Regulatory Guide 1.35 Revision 2.
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c.
Wisi, Janney Elstner and Associates Inc. Post Tensioning Report.
d.
The in service tendon surveillance program of the containment building post-tensioning system includes determination of tendon lift-off forces, tensile testing of wire specimens, chemical tests on filler grease, and inspection of corrosion of anchor hardware and sample wir's.
Three vertical, three e
horizontal, and three dome tendons were inspected.
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Information Notice 85-10, Rev 1, describes anomalies at facilities with similar tendons.
It was reviewed by the licensee and found to be not applicable due to difference in materials in the lift-off plates.
The inspector reviewed a licensee report that identifies the Millstone tendon anchor head material as type 4330 with a Rockwell C hardness of 16 versus the Farley No. 2 mate-rial of Rockwell C hardness of 42 12.
The material is more ductile at Mill-stone 2 and is less susceptible to hydrogen stress cracking.
Licensee visual inspection of lift off shims and button heads identified no anomalies.
TSs requires the inservice tendon surveillance program to assess containment system tensioning quality.
The surveillance program detects anomalies in lif t-off tensioning, filler grease condition, crack patterns, end cap cor-
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rosion, and tensile strength of selected wires.
The licensee developed data sheets to document the requirements of the TS. -Inspector's review of this documentation identified no inadequacies in procedure content or format.
4.1-Observation of Tendon No 12H07 Surveillance Test.
On March 19, 1986, the resident inspector observed the following activi-ties in relation to surveillance of Tendon No.12H07:
Surveillance equipment setup;
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Communications between the two tendon end anchorage locations; Use of test equipment;
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Adherence to the surveillance procedure;
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Repeated tendon lift-offs;
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Tendon shim removal and inspection; and,
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Tendon continuity pulls.
The inspector interviewed test personnel and determined that they pos-sessed in-depth knowledge of the above activities.
All observed para-meters were within required tolerances.
No unacceptable conditions were observed.
4.2 Observation of Concrete Crack Patterns At concrete areas adjacent to surveillance tendon bearing plates, crack patterns are documented on a one-foot grid. On March 24, the inspector examined one such crack pattern on either side of one end of a vertical tendon.
No cracks greater than 0.025 inches were observed.
A review of licensee documentation found this in agreement with licensee findings.
In addition, the data packages for the other eight tendons were reviewed.
Crack patterns appear to have settled out between 0.01" and 0.025", with little or no additional opening since the last surveillance.
No anoma-lies were identified.
4.3 Review of Detensioning and Retensioning Data The inspector reviewed the preliminary lift-off pressures and forces (detensioning and retensioning) data.
TS requirements were met in detensioning.
One wire was found parted in Tendon 1023.
Tendon lift off acceptability was recalculated and found satisfactor.
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The licensee determined that a decrease in total lift-off tension
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was to be instituted based on the following,
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c Permanent concrete cracking has taken place over the ten year period.
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The permanent set due to compression of concrete has been completed.
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Surveillance tendon wirss have been stretched, removing any relaxa-
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i tion in wire tension over the ten year period.
Retensioning forces, which averaged approximately 1540 Kips, have been decreased to 1470 Kips.
The reduction in overall forces was in line with maintaining containment integrity acceptable for forty years.
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suance of the Surveillance Report (due in August, 1986), the licen,yecis conclusions will be subject to further NRC review.;
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4.4 Corrosion Review i
Three wires were removed for-surveillance.
Licensee visual inspection on site found no degradation.
These wires were sent to an offsite laboratory for tensile strength analysis.
Results will be reviewed when
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4.5 Grease Anomaly During tendon surveillance, one tendon's filler grease appeared to be darker than the rest.
Two samples were sent to laboratories, one from i
this tendon and a second from another tendon under test.
The results, received on March 20, 1986, indicated no essential differences in grease makeup.
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4.6 Additional Comments
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Corrosionofbeari.ngplatesandcapinteriorsrangedfrom0.006"to-liss
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than 0.010" in pit denth.
This was found acceptable by the licensee;>
A program for regreasing additional below grade tendons where water was found on removal of end caps is in progress and will be reviewed during a future inspection.
5.
Observation of Surveillance (Unit 1 & 2)
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Unit 1
Manual Scram Functional Test On March 18, 1986, the inspector observed the Scram Discharge Volume High Water Level Scram Testing per SP408D in the control room and in the discharge volume area.
Testing isolates and exercises one set of level sensing switches on one discharge volume at a time and results in a system one-half scram (no
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rod motion) which must be cleared prior to testing the other set of switches.
Test personnel exhibited awareness of the implications of this testing.
An operator monitored the solenoid-operated scram pilot valves for abnormalities following the reset of the one-half scram and prior to testing the second set of switches.
Test parameters were within specifications.
Test personnel ex-hibited procedure and system familiarity and expertise.
No unacceptable con-ditions were identified.
The inspector also observed the ATWS Recirculation Pump Trip / Alternate Rod Insertion Calibration Test per SP409B from both the control room and the in-strument enclosures.
Minor procedural problems were found: two areas of the procedure needed clarification to ensure correct operator action during the test.
These were identified to the licensee.
Corrective action was promptly taken.
The inspector had no further questions.
Additional Tests observed were:
Unit 1 - No. 668.1-1 Diesel Generator Operational Readiness Demonstration.
Unit 2 - 2604A HPS1 Pump Test; 2604B HPS1 Pump Test; 2604D LPS1 Pump Operability Test; and, 2602A Reactor Coolant Leak Rate.
No deficiencies were identified during these tests.
6.
Shroud Head Bolt Cracks-Unit 1 During the last refueling outage, Millstone I was informed that cracking of shroud head bolts (SHBc) had occurred at five other similar BWR plants.
At that time, the licensee had completed the torquing of SHBs to greater than or equal to the 50 ft. Ib. force recommended.
General Electric SIL No. 433, issued on February 7, 1986, states that complete bolt failure would be de-tected in bolts that could not be torqued to 50 ft. lb. force.
Licensee re-view concluded that SHB non-destructive examination (NDE) need not be per-formed until the next scheduled outage.
The inspector noted that the licensee had reviewed the need to purchase addi-tional spare head bolts.
Shroud head bolt NDE test results will be reviewed incident to routine inspection.
7.
Fuel Rerack Program-Unit 2 7.1 General License Amendment No. 109 authorizes an increase in spent fuel pool (SFP)
storage capacity fram 667 to 1112 fuel assemblies. A change to assure availability of a laop of shutdown cooling for SFP cooling if pool tem-perature rises above 140F is to be submitted.
The new fuel racks (17
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of 19) have been received and the old fuel racks are being removed from the SFP.
The following NRC inspections were made to address the various phases of this progam.
7.2 Receipt Inspection of New High Density Racks The licensee's (NUSCO) Construction Quality Control receipt inspection of the new spent fuel racks was reviewed.
Receipt inspection includes:
preliminary cleanliness, visual examination, vendor material certifica-tion, review of nonconformance reports, dimensional inspection, and cell gauging.
The inspector observed the gauging of two racks in the SFP area.
In addition, the fit of the fuel entrance cone was verified.
A sample of receipt inspection documentation for three of the new fuel racks was re-viewed.
The resident also reviewed all deviations from contract require-ments (DCRs) for Spent Fuel Rack 9.
No unacceptable conditions were identified.
Other Deviations from Contract Requirements (DCRs) were reviewed.
The DCRs included the following items.
9430139-1-Cell Blocking Device-4-Foot Hold Block Dimension-8-Chrome on 4 attachment screws-10-Depth of adjustment hole-13-14-Additional screws-chrome plated-18-Change on Cell Foot-29-Overall cell dimension-33-Gauging at cell corner All of the above were accepted by NUSCO.
The inspector questioned the acceptance of an overall slightly large spent fuel rack.
The licensee reviewed this concern and accepted a tighter spent fuel rack cluster.
The inspector had no further questions on this rack.
In addition, documentation including the following were reviewed.
1)
Certificates of Conformance 2)
Receipt Inspection Checklists 3)
Cell Gauging Inspection Check Lists 4)
Technical Change Requests 94030139, 1, 3, 6A, 8, 9 and 10 Technical Change Request No. 10 accepts the construction tack welds used to hold cell units together prior to placement of 48" welds at cell seams at the top and bottom of individual cells.
The licensee based their acceptability conclusion on the limited number of such tack welds and on gauging of cells to demonstrate no fuel cell interference with the tack welds.
Similar welds were identified and accepted on other racks.
The inspector had no further questions on this are _
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7.3 Additional Receipt Inspection Items At the conclusion of this report period, the licensee had the following items to resolve to conclude the receipt inspection procedure.
Spent Fuel Rack CE No. 1, NUSCO N14, has not been received and re-
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inspected.
Three Nonconformance items are to be resolved: 1C2-86-214 (N15),
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215 (N15), and 218 (N18).
These address gauging and a deformed bottom support beam for one of the new racks.
Resolution of these items will be reviewed during routine inspection.
7.4 Additional Receipt Inspection The licensee's request for high density fuel pool racks states that 100%
of the neutron attenuator material placement in the Region I racks (to be used for core off load) would be verified.
The licensee had made only a sampling inspection.
After the inspector questioned how the lic-ensee met his commitment, the licensee expanded the receipt inspection program to verify all neutron attentuation material is in place prior to fuel rack placement in the SFP.
The inspector had no further ques-tions in this area.
7.5 Spent Fuel Pool Area Radiological Controls Barriers enclosing the spent fuel pool areas were established for the work.
Entrance to the areas was controlled by the health physics or-ganization.
The licensee protected the divers by providing physical barriers in the pool, by requiring additional dosimetry, and by utilizing communication equipment for each diver.
Work areas were monitored by a TV camera and direct visual observation.
Tool Material Control was established at the entrance to the pool work area.
The inspector reviewed the material entry log, records of personnel ex-posure and surveys of the pool bottom, top, and sidewall areas.
Two health physics surveys were observed.
The exposure records for about 20 divers showed a quarterly exposure of less that 250 mR for each; pre-vious exposures at other sites were factored into the totals.
No deficiencies were identified in the items inspected.
7. 6 Other Spent Fuel Pool Re-rack Activities Other Spent Fuel Pool (SFP) Re rack Project activities observed included:
Plant Operations Review Committee (PORC) meeting concerning approval
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of project procedures;
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New spent fuel rack QA receipt inspections;
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Spent Fuel Pool area preparations;
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Diving operations in the Spent Fuel Pool;
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Interference removal operations; and,
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Initial old spent fuel rack removal.
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These inspections were preceded by review of supporting engineering documentation ar.d procedures.
Licensee preparations were well planned and adequate.
Of particular note were the precautions taken during the initial rack removal to minimize personnel radiation exposure in a po-tentially high airborne areas.
The expected levels were not reached.
The rack was removed from the SFP and placed in the cask washdown pit.
for cleaning by hydrolazing.
No unacceptable conditions were identified.
Inspector observations will continue as the remaining racks and old flooring steel is removed and the new racks are installed.
Additional areas to be reviewed on a sampling basis include:
Transportation of removed fuel racks;
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Methods used for fuel assembly relocation;
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Leveling of new racks;
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Overall plan for fuel assembly placement; and,
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Radiological controls for liquid waste from hydrolazing of the old
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racks.
The inspector had no further questions.
8.
Isolation Condenser Valve Anomaly Reference:
Technical Specification 3.5.E 8.1 General:
The Isolation Condenser System provides a method of reactor core decay heat removal.
On March 26 at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />, the Isolation Condenser was declared inoperable during surveillance testing (SP 627.5) when conden-sate return motor-controlled val.*e 1-1C-3 failed to respond to an "open" signa.
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8.2 Description of Event:
Licensee investigation found that the motor-operator failed to perform its function due to high torque.
The previous surveillance, on January 7,1986, indicated a satisfactory opening of Valve 1-1C-3.
Discussion with the licensee indicated that, upon completion of restoration of iso-lation condenser after the last test, leakage was identified past valve 1-1C-3.
The valve was manually torqued and leakage subsided.
If a motor-operated valve is torqued, a trouble report should be prepared (as dis-tinct from "shall be prepared") in accordance with Departmental Instruc-tion 1-0P-6.15.
No trouble report was written and the valve remained manually over-torqued from January 7 to February 26.
The valve, which is outside containment, could have been reached and manually opened in about two minutes in case Isolation Condenser operation was required.
8.3 Licensee Immediate Corrective Action The licensee restored the valve to operability by moving it off its seat manually, readjusting the closing torque switch, and satisfactorily test opening the valve on March 26.
8.4 Findings Management control of hand torquing of motor-operated valves important to safety, including but not limited to the preparation of a trouble report, is an unresolved issue (50-245/86-03-01).
9.
Maintenance Condenser Tube Degradation - Unit 1 The licensee has experienced a number of main condenser outages due to tube leakage.
Unit tube plugging status review identified the following conditions:
Condenser Bay Tubes Plugged A
7.3%
B 10.0%
C 6.5%
D 9.5%
Total tubes per bay are approximately 10,098.
Discussions with licensee in-dicate that the normal guideline for retubing is 10% plugging.
The licensee has identified the peripheral and direct impingement tubes as having had numerous failures, Presently, the licensee plans to review continued opera-tions, based on the maintenance and chemistry programs' ability to control condensate acceptability, until a 15% tube loss is reached.
This item will be reviewed during routine inspection.
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10. ' Integrated Safety Assessment Meeting - Unit 1 The NRR staff has issued a Safety Evaluation for the topics to be considered in the Integrated Safety Assessment Program (ISAP) for Millstone 1.
On March 12 the resident inspectors attended an NRR/ Licensee staff meeting at the Berlin, Ct., corporate offices on ISAP topics.
During this meeting, the NRR and licensee staffs discussed the status and proposed resolution of approximately 80 ISAP topics.
Those topics were classified as warranting further licensee action, satisfactorily resolved, or removed based on the Probablistic Risk Safety Study (PSS). The PSS identified areas where minor or no safety benefit was to be realized, manpower considerations, and economic feasibility aspects.
Topics related to recent occurrencess at Millstone 1 were discussed, including stud failures on steam line restraints, the offsite power loss during Hurri-cane Gloria, and Gas Turbine anomalies.
The above related to ISAP Topic 1.01, Gas Turbine Generator Start Logic Modification, ISAP Topic 1.16.3 & 4 Alter-nate Cooling for Shutdown and Powering of Shutdown Equipment, and ISAP Topic 2.17, Plant Electrical Distribution.
An on-site meeting was conducted on March 13.
Discussions on ISAP related operations, maintenance and hardware changes were held. Observation of on-going work items were discussed.
NRC Region I observations on motor-operated valve performance and Loss-of-Power problems were also discussed.
-The resident inspectors will monitor PSS-related design changes incident to routine inspection.
11.
Unresolved Items Unresolved items on which additional information is required are identified in Details 4.3, 4.4, and 7.4.
12.
Management Meetings At periodic intervals during the course of this inspection, meetings were held with senior plant management to discuss the scope and findings of this in-spection.
No proprietary information was identified as being in the inspec-tion coverage. At no time during the inspection was written material provided to the licensee by the inspector.
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