IR 05000317/1986018

Insp Repts 50-317/86-18 & 50-318/86-18 on 861018-1130.No Violations Noted.Major Areas Inspected:Operational Events, Outage & Routine Insps,Physical Security,Lers,Maint, Surveillance & TMI Action Plan Items

ML20212B664
Person / Time
Site:	Calvert Cliffs
Issue date:	12/12/1986
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20212B590	List: IR 05000317/1986018 ML20212B585
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM 50-317-86-18, 50-318-86-18, GL-82-28, NUDOCS 8612290288
Download: ML20212B664 (16)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

i Docket / Report: 50-317/86-18 License: DPR-53 50-318/86-18 DPR-69

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Baltimore Gas and Electric Company Facility: Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Inspection At: Lusby, Maryland Dates: October 18, 1986 - November 30, 1986 Inspectors: T. Foley, Senior Resident Inspector D. Trimble, Resident Inspector Approved: A L. E. Tripp, Chief, Reactor Projects Section 3A 8 //////dd Date (

f Summary: October 18 - November 30, 1986: Inspection Report 50-317/86-18; 50-318/86-18 Areal.. Inspected: (1) facility activities, (2) licensee action on previous inspec-tion findings, (3) operation events, (4) routine' inspections including outage in-spectiens, (5) physical security, (6) Licensee Event _ Reports, (7) maintenance, (8) surveiH ance, (9) TMI Action Plan Items, (10) radiological controls, (11) re-ports to the NR Inspection Hours totalled 209 hour0.00242 days <br />0.0581 hours <br />3.455688e-4 weeks <br />7.95245e-5 months <br /> Results: Notwithstanding the abundant activity associated with the Unit 1 outage, no significant concerns were ishntified. Licensee efforts have resulted in close adherence to the outage schedule. Routine resident inspection, a regional team inspection of the maintenance area, and an ISI inspection all identified no signi-ficant problems. The ALARA group is demonstrating positive attributes in the area of radiological control ,

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DETAILS Within this~ report period, interviews and discussions were conducted with various licensee personnel, including reactor operators, maintenance and surveillance tech-nicians and the licensee's management staf . Summary of Facility Activities Unit Operating Experience Unit 1: The period commenced with Unit 1 at full power. On October 24 the unit shut down to commence the 10 year In Service Inspection and Refueling Outage. Outage activities to date have been routine with only minor depar-tures from the schedul Unit 2: The period commenced with Unit 2 at full power and, except for power reductions for maintenance and surveillance, the unit remained at full power throughout the perio Facility On November 21, recently qualified reactor operators and senior reactor opera-tors were presented license certificates by the NRC's responsible branch chie On November 23, the licensee conducted an "Open House" for all employees and contractors who war.. at Calvert Cliffs and their immediate famil Beginning November 17, two regional inspectors examined the In-Service In-spection program in process including Reactor Vessel Weld examination and, to a limited extent, the steam generator Eddy current testing. Another three regional inspectors examined the maintenance program and activities in pro-gress. These inspections are documented in Inspection Report 50-317/86-24; 318/86-2 . Licensee Action on Previous Inspection Findings (Closed) Unresolved Item 317/82-07-06; Amendment Request to Specify Reactor Coolant Flow During Modes 5 and 6. Flow specifications are now contained with the Technical Specifications for Mode 6 and all modes when reducing boron concentration. Flow requirements during Mode 5 are currently being reviewed by the licensee and NRR as part of a response to Generic Letter 86-13. This item is close (Closed) Violation 317/82-07-10; Operation with Inoperable Hydrogen Analyze This topic was reviewed and is addressed in Inspection Report 317/86-17; 318/86-17. This item is close (Closed) Unresolved Item 317/82-14-01; Complete Drawing Control Evaluatio The licensee has completed their review of the drawing control program and have revised it as necessary. No problems have been noted in this area within the previous two years. This item is close _ _ -

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-(Closed) Inspector Follow Item 317/82-30-02; Review Licensee Action to Minim-

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ize the Recurrence of Inoperable Emergency Diesel Generators. This concern

is addressed in several' resident inspection reports including 317/85-30, 317/86-16, and this report, 317/86-18. LThe licensee is currently overhauling-No. 11 and 12 Emergency Diesel. Generators. This concern is-resolved'and-close (Closed) Violation 317/83-03-01; Aperture Cards Not Controlled by' Formal Pro-cedur The licensee has created a formal control of aperture cards within-the Calvert Cliffs Administrative Instruction CCI-138-D " Aperture Card, Micro Fische, and Micro Film Control". .This item is close (Closed) Inspector Follow Item 317/84-19-02; Review of Non-Radioactive Waste Generated?in Controlled Area. Subsequent inspections by regional specialist have' reviewed this area and found no unacceptable conditions. This item is close (Closed) Unresolved Item 318/85-32-01; Numerous Minor Maintenance Problems with Emergency Diesel Generators, DG Rooms, System, Subsystem, and Supporting Auxiliaries. See Section 4 of this report regarding Emergency Diesel Genera-tors. This item is close (Closed) Inspector Follow Item 317/84-31-04; Licensee to Identify Root Cause of MSIV Failures.- The licensee has identified the cause to be failed bladders and has decided to replace the entire MSIV package. The MSIV's are currently being replaced with a less complex' system. This item is close (Closed) Violation 318/85-01-01; Inadequate Posting of High Radiation Area in the Overhead. The licensee has implemented a program to control scaffold-ing in order to control access to possible high radiation in the overhea Review of this action appears effective. This item is close (Closed) Inspector Follow Item 317/85-01-02; Licensee to Post Higher than Nor-mal Radiation' Areas as such, to alert personnel of the change in Radiation Area Fields. The licensee has responded to the inspector's concerns and now posts areas where the intensity of radiation is sufficiently high yet not a high radiation area. This item is close (Closed) Unresolved Item 317/85-01-03; Inadequacies during Testing of Person-nel Air Locks. Subsequent review of the test program indicate that the in-adequacies previously demonstrated are no longer evident. This item is close . Operation Events Unit 1 Outage Activities A description of the outage scope and schedule was discussed in Inspection Report 317/86-16; 318/86-1 l

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The outage consists of four principal areas of work as follows: reactor work, moisture separator reheaters (MSR), main steam isolation valves, and turbine

. Wo r The reactor work constituted the critical path. On day 12 of the out-age,_the critical' path began to experience delays primarily due_to the fuel transfer system hoist box / grapple interference and limit switches on the re-fuel pool / spent fuel pool trolley. Fuel transfer began on November 7 and the reactor _ core was entirely off-loaded on November 12, three days behind the scheduled dat Another significant unplanned impact occurred during steam generator (SG) eddy current testing. The licensee had planned on testing 256 tubes per SG to ful-fill the Technical Specification requirement of 3%-tube examination for Cate-gory C-1 requirements. Review of the results of this inspection determined that greater than 1% of the examined tubes were degraded. Based'on this, the Technical Specifications required that all 8500 tubes per SG be examine This determination required the licensee to reschedule work hours to go to a 24-hours a day, 7-day per week inspection proces Emergency diesel generator work also caused an impact on the outage. Testing of the reassembled No. 11 EDG determined that the generator end of the diesel had 7 mils of vibration, slightly exceeding the allowable tolerance of 5 mils provided by the vendor. This required additional alignment and testing and caused delays in starting the overhaul of No. 12 EDG. Because of the delays and possible further unanticipated testing delays, the repairs and testing of the No. 12 EDG may not be complete for several days after the scheduled date for commencement of fuel loading. Because this would eventually lead to delaying the start-up of the unit, the licensee requested an emergency Technical Specification change to refuel with only two of the three required EDGs available for_ emergency power on site. The licensee also committed to maintain other compensatory measures, including having the SMECO independent off site power supply available and the portable 1000 KW diesel generator described in Inspection Report 86-16 in plac Toward the mid point of the scheduled outage, the critical path had regained some of the time lost such that critical work was about one day behind sched-ule. All other work closely approximated the original schedul . Review of Plant Operation - Routine Inspections Daily Inspection 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-perature and pressure, control room annunciators, radiation monitors, effluent monitoring, emergency power source operability, control room logs, shift supervisor logs, tagout logs, and operating order No unacceptable conditions were note ,. . . _- . .. ~ .

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5 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 as performed. Operability of instruments essential to system performance was assessed. The following systen:s were checked:

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Unit 2 Emergency Diesel Generator Breaker Line Up 4 --

Unit 2 Service Water *-

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Unit 2 High Pressure Safety Injection System

• For this system, the following items were reviewed: _The licensee's system lineup procedure (s); equipment conditions / items that might degrade system performance (hangers, supports, housekeeping, etc.);

instrumentation lineup and operability; valve position / locking (where required) and position indication, and availability of valve operator power suppl During review of the Emergency Diesel Generators Systems, while Nos. 11 and 12 EDG's were sequentially removed from service for overhaul, it was noted that No. 21 EDG appeared to be exceptionally well maintained. It is evident that appropriate attention is now being given to this safety

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' Report 317/85-32.

i During examination of the above Unit 2 HPSI system, it was noted that

. several welds'on the suction and discharge piping of the Unit 2 LPSI pumps appeared to be of a copper color. No other welds associated with the safety system appeared like these. The inspector discussed this with j, the Maintenance Manager who indicated that'a system engineer would look

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into i The inspector noted that the system was hydrostatically tested

last outage and the welds, except for the color, appear to be in excel-lent condition. Subsequently, the licensee determined that this was oxidation of the weld and was an expected phenomenon.

! The inspector also discussed two concerns with the Outage Coordinator

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and separately with the Maintenance Manager regarding (1) poor weld (s)

on No.11A Reactor Coolant pump oil reservoir level transmitter sensing

, line. Welds are not clean, look sloppy, are non-uniform, and have weld

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slag deposits remaining; and (2) the No. 118 Reactor Coolant Pump control

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bleed off pressure sensing line appears inadequately supporte The Maintenance Manager stated that these would be evaluated.

! No unacceptable conditions were noted.

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6 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. Area radiation and air monitor use and operational status was reviewed. Plant housekeeping and cleanliness were evaluate Verification of several tagouts indicated the action was properly con-ducte No unacceptable conditions were note Other Checks (1) Spent Fuel Pool Cooling System Capacity During the outage, the licensee off-loaded the entire fuel inventory into the spent fuel pool. In preparation for this, the licensee recognized that the decay heat load of the freshly spent fuel to-gether with the decay heat of the previously stored fuel might ex-ceed the design capacity of the Spent Fuel Pool Cooling System (SFPCS) for approximately 25 day Because of this concern the licensee wrote Change Report 86-248 which provides a supplement to Operating Instruction 24 " Spent Fuel Pool Cooling System" to provide operating instructions to augment the SFPCS with the shutdown cooling system should the decay heat load become excessiv As a result of this effort, the licensee determined the effective-ness of the spent fuel pool coolers by recording the cooler inlet and outlet temperatures. The efficiency of the SFPCS was traced /

trended during the core off-load until the predicted decay heat load was less than the design capacity of the SFPCS at 95 degrees Fah-renheit. The trending of SFPC efficiency demonstrated that it was not necessary to augment with shutdown coolin Although the augmentation procedure was not fully implemented, the planning and procedural development and licensee recognition of the potential for excessive heat loads in this mode is indicative of thorough prior plannin (2) Outage Inspections The inspector performed independent observations of work associated with the following activities and reviewed adherence to acceptable maintenance practices and adequate radiological and quality control .. _.- -. . ,

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(a) Inside Containment (i) . Reactor Work '

Observed disassembly, ~ removal of ventilation and shielding, replacement of excore neutron flux detectors, removal of in-sulation, setting of reactor cavity pool. seal ring and deten-sioning reactor vessel studs. . Verified setting of containment integrity pursuant to Technical Specifications 3.9.4. Subse-quently, reviewed completed procedures to ascertain require-ments were met during lifting of the reactor head and refueling operation Specific reviews were conducted to verify that the following Technical Specification requirements were met:

TS 3. Boron Concentration TS 3. Instrumentation TS 3. Decay Time TS 3. Containment Penetrations TS 3. Communications TS 3. Refueling Machine Operability TS 3. Crane-Travel TS 3. Shut-Down Cooling and Coolant Circulation TS 3. Containment Purge Valve Isolation TS 3.9.10 Water Level - Reactor Vessel TS 3.9.11 Spent Fuel Pool Water Level Observed defueling operations, fuel handling, and inspections (100% ultrasonic testing, partial eddy current and visual testing of fuel assemblies), and periodic observations of the removal of the core support barrel, instrumentation and in-spections of the reactor vessel welds utilizing the Babcock Wilcox Programmed and Remote device (PAR).

(ii) Steam Generator Work Observed local leak rate testing of the steam generator manways per STP-571-1 " Local Leak Rate Test"; preparation testing and

, installation of steam generator nozzle dams and supporting

! equipment; sluicing of the steam generators and associated l chemistry controls; and independent examination of internal

! upper head welds of the secondary side of the No. 12 steam l generator and the moisture separators and dryers.

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Through procedure reviews and discussions with the responsible

! cognizant engineer the inspector was kept informed of steam !

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Prior to commencement of the outage, licensee representatives presented to the resident inspectors a memorandum dated Novem-ber 4, 1986, with inspection plans detailing the specific number and location of tubes to be inspected for the purpose of satisfying the Technical Specification requirements. The plans specified the location of the tubes in the 3% numerical sample required by Technical Sepcifications, the location of all previously defective tubes, the sludge pile, plugged tubes and tubes with previously identified indication The licensee examined the 256 required tubes and determined that greater than 1% were defective, 7 in SG No. 11 and 8 in SG No. 1 Most of these were defective based upon a greater than 10% growth of previously noted indications. (The licensee performed a 100% examination of the SG's last outage.)

Based on these results, the licensee was required to inspect 100% of the tubes in each SG pursuant to TS 3.4.5 " Steam Generator Operability". The inspection effort is not planned to be completed until December During this steam generator eddy current testing program, the licensee used two zero-entry SM-10 devices to eddy current test

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the generator tubes and minimize personnel exposure. Along with the automated eddy current test device (SM-10), Zetec 560

" Bobbin Coils" magnetic bias probes were used to confirm indi-

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cations. Data from testing was reviewed by two separate groups independent of each other. The licansee's technique examines the tube from tube sheet to tube shee (iii) No. 12 Reactor Coolant Pump (RCP)

A review was conducted of the radiological controls associated with the removal and handling of the No. 12 Reactor Coolant Nmp rotating members, i.e. , shaft, impeller, and hydrostatic bearing. It was ascertained that training was conducted several weeks prior to the removal of the pump, and that ade-quate surveys, appropriate controls, and precautions were in-stitute (iv) Reactor Vessel Nozzle Plug Installation In order to provide isolation of the Reactor Coolant System (RCS) while performing maintenance on parts of the primary system including the Reactor Coolant Pump internals inspection, which requires draining down the RCS loops below the RV nozzles, two hot leg plugs and four cold leg plugs were installed in the reactor vessel nozzles with assistance from divers after

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the core cupport barrel was remove The plugs permit the reactor vessel weld inspection to proceed in parallel with the other primary wor Associated with this, an additional temporary level transmitter was installed on sensing line 1-PDT-12313, RCP 118 pressure sensing line, in order to provide an additional refuel water level indication. The inspector compared local tygon tube

, direct readings with those indicated by remote readouts in the Control Roo No inadequacies were identifie (b) Outside Containment (i) Main Steam Isolation Valve (MSIV) Replacement Due to poor reliability and excessive maintenance, the licensee chose to replace the existing Rockwell International /Greer Hydraulics MSIV's. The system is being replaced with Rockwell International A-180 gas-hydraulic actuators and a balanced disk globe valve interval. This modification required pulling cable, removing existing equipment, welding build-up, machining, honing and polishin Several tours of this work area were made during the reporting period. No inadequacies were note (ii) Component Cooling and Service Water Cooling Heat Exchanger Tube Replacement Work associated with both the Service Water and Component Cool-ing Heat Exchanger was observed at various stages of the total evolution. Channel head removal, pulling tubes, cleaning, alignment, re-tubing, and rolling of the tubes, and replacement of the channel heads was witnesse Component cooling channel heads were sent off-site to be rubber lined before reinstalla-tion. An independent examination of the internal surfaces of the No. 11 Service Water Heat Exchanger was conducted while the tubes were removed. Very little foreign matter was noted other than dried sil The internal supports appeared to maintain their original structural integrity, and welds, joints, supports and internal components appeared free of corrosion, cracking, or other impairmen (iii) Salt Water System and Sluice Gates Tours were made of each intake structure (whil. Crained with the travelling screens removed). Examinations of the following were conducted:

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Sluice gates

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Sluice gate lifting and holding apparatus

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Structural integrity of intake structure walls and flow partitions

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Circulating water pump volute (both sides

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Circulating Water System concrete piping and joints; and

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cleaning of marine growth from the surface The sluice gates'at the time of inspection were moderately corroded. New sluice gates are being installed later in the outage (FCR 84-0152). Lifting devices were being replaced at the time of inspection. Internal surfaces of the Circulating Water System piping were in generally good condition. Approxi-mately 1 - 2 inches of marine growth existed on the surfac Removal by hydrolasing was in progres No growth remained where hydrolasing was complet (iv) Main Turbine Work Conducted periodic inspections of the Main Turbine Control Valve and Stop Valves, and observation of Moisture Separator Reheater, Feed Water Reheater, Low Pressure Turbine, and Main Generator wor No unacceptable conditions were note . Observation of Physical Security Checks were made to determine whether security conditions met regulatory re-quirements, the physical security plan, and approved procedures. Those checks included security staffing, protected and vital area barriers, vehicle searches and personnel identification, access control, badging, and compensatory meas-ures when require No unacceptable conditions were note . Review of Licensee Event Reports (LERs)

LERs submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of cor-l rective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted on site follow up. The following LER's were re-viewed:

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LER N Event Date Report Date Subject Unit 1 86-05 10/01/86 11/03/86 Main Vent Wide Range Noble Gas Effluent Monitor, In-operable Due to Deficient Procedure 86-06 10/10/86 10/31/86 Reactor Trip Due to Turbine Trip from Loss of Condenser Vacuum Unit 2 86-07 Rev. 1 09/12/86 10/10/86 Manual Reactor Trip Due to Partial Loss of Feedwater Flow to Steam Generators Inoperability of Wide Range Noble Gas Effluent Monitors (WRNGM)

On October 1, 1986, the licensee declared the Unit 1 and Unit 2 WRNGM's in-operable from November 11, 1985, on Unit 1 and from December 11, 1985, on Unit 2. This action was taken when the licensee's systems engineer recognized deficiencies in the WRNGM surveillance test procedures (STP M-564-1 and M-564-2) which called into question detector calibratio STP M-564-1 was first conducted on Unit 1 on November 11, 198 During this test, the responses of the high and mid-range detectors were not within the required 15% of the expected count rates of the calibration sources. The technician performing the test was unable to adjust detector outputs to agree with the expected source count rate (15%).

A vendor representative advised the technician to align the discriminator, record the detector output and consider that output value to be the reference count rate for the source for subsequent calibration checks of the Unit 1 detectors. In fact, the reason that the mid and high-range detector outputs did not agree with the expected source count rate was that the detectors were inoperable. By following the vendor representative's recommendation, an in-accurate count rate for the source was recorded as a reference number and NBS traceability was los The STP's were modified to include the incorrect in-formation provided by the vendor representativ The incorrect STP was used for a December 11, 1985, Unit 2 WRNGM tes How-ever, detector outputs were within 15% of the expected source count rate and no improper adjustments were made to the monitors. Because an incorrect STP was used in December 1985 and because the licensee later (October 1986) iden-tified two component problems (remote pre-amplifier voltage drift and an out of specification mid range detector power supply), the licensee considers the Unit 2 monitor to have been inoperable since December 198 .

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The Unit 2 WRNGM was returned to service on October 18, 198 At the close of the inspection period, the licensee was awaiting receipt of a coax cable prior to declaring the Unit 1 monitor operable. Unit 1 is currently in a re-fueling outage and monitor operability is not required in Mode 5 or 6 opera-tion. Receipt of the cable is expected prior to the scheduled Unit 1 start up date. Licensee Event Report 86-05 dated November 4, 1986 describes the WRNGM inoperabilit No unacceptable conditions were note . Plant Maintenance The inspector observed and reviewed maintenance and problem investigation activities to verify compliance with regulations, administrative and mainten-ance procedures, codes and standards, proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualifications, radiological controls for worker protection, fire protection, retest requirements, and reportability per Technical Specifications. The following activities were included:

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PM 2-12-I-RQ4-3, #22 Salt Water Air Compressor Oil Level and Temperature Switches

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PM 1-58-I-R-15, RPS Power Supplies

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PM 1-36-1-R-14, Turbine Driven AFW Pump Control I/P Controller

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MP-RV-20, Removal of Core Support Barrel

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M0-206-297-362A and 363A, Replacement of Main Steam Isolation Valve in-ternals with Rockwell A-180 Hydraulically Operated Valve

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M0-206-172-634A, Replacement of 1AFW 1-CV-4531 Diaphragm for the AFW Turbine Driven Pump Steam Isolation Valves (4530 and 4531)

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PM 1-36-1-R-14, Turbine Driven Pump Flow Control PM on I/P Controller No unacceptable conditions were note . Surveillance The inspector observed parts of tests 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 tests were reviewed:

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OP-5 Change 86-232, Plant Shut Down from Remote Shut Down Panel

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STP M-571, No. 11 Steam Generator Manway Penetration 11-1 No. 12 Steam Generator Manway Penetration 12-1 No. 12 Steam Generator Manway Penetration 12-2 Hydrogen Sample Valve SV-6507 Penetration 49-C Refueling Transfer Tube Penetration 42

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STP M-525-1, Steam Generator No. 11 and 12 Line Pressure Transmitter Calibration Operating Procedure 5/ Operating Instruction 32, Change Report 86-232 Reactor Coolant Cool Down From IC043 During the normal plant shut down after Unit 1 was subcritical, the licensee conducted an approximate 25 degrees Fahrenheit cooldown from the Remote Shut Down Panel (1C043).

The purpose of the procedure was to verify that an RCS cooldown could be ac-complished from outside the Control Room. For time and safety considerations, a maximum of 25 degrees Fahrenheit cooldown was conducted from IC4 The procedure was to be a one time operatio Training on the remote shutdown panel was performed just prior to this evolu-tion and, in addition, will be incorporated into the simulator training pro-gra During the procedure, the licensee demonstrated a transfer of control from the Control Room to 1C043; adequate communication and coordination be-tween IC043 and the Auxiliary Feed Pump Rooms and adequate control over the atmospheric dump valves and auxiliary feed pump Both resident inspectors observed this evolution. The licensee also performs a Performance Evaluation 1-102-1-0-R Safe Shut Down Panel Operability Verifi-

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cation during Modes 5 or 6 which verifies operability of each controller and indicator on panel IC043 independentl No unacceptable conditions were note . Licensee Action on NUREG 0660, NRC Action Plan Developed as a Result of the TMI-2 Accident The NRC's Region I Office has inspection responsibility for selected action plan items. These items have been broken down into numbered descriptions (enclosure 1 to NUREG 0737, Clarification of TMI Action Plan Items). Licensee letters containing commitments to the NRC were used as the basis for accept-ability, along with NRC clarification letters and inspector judgment. The following item (s) was/were reviewe NUREG Item II.F.2, Inadequate Core Cooling Instrumentation. NUREG 0737,

II.F.2, REG Guide 1.97 and Generic Letter 82-28 all address the require-ments for inadequate core cooling instrumentation. In response to these the licensee proposed a system for detecting and monitoring inadequate core cooling (ICC) conditions including a subcooling margin monitor (SMM),

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core exit thermocouples (CET) and a reactor vessel level monitoring sys-tem (RVLMS) for Calvert Cliffs Units 1 and 2. The NRC staff and its contractor, Oak Ridge National Laboratory (ORNL), reviewed the BG&E sub-mittals dated March 10, 1983, October 14, 1983, and April 10, 1984 de-scribing the proposed system and determined the proposals to be accept-abl The following is a brief description of each portion of the overall system:

Reactor Vessel Level Monitoring System The RVLMS is a differential temperature measurement concept similar to the generic Ccmbustion Engineering (CE) heated junction thermocouple (HJTC) system except for the control room display. The BG&E RVLMS in-cludes a qualified light array in the control room for each HJTC channel to indicate the HJTC sensor being covered or uncovered. Class IE trend recorders are also provided for vessel inventory tracking capabilit BG&E considers the control board mounted Class IE display as a backup display with the primary display being the plant's safety parameter display system (SPDS).

Subcooling Margin Monitor (SMM)

The subcooled margin monitor consists of temperature and pressure sensors, associated cabling and connectors located inside containment, and redund-ant dedicated digital subcooled margin calculators and continuous digital displays located outside containment. The subcooled margin monitors were installed utilizing existing reactor coolant system instrumentation channel The subcooled margin monitor has been incorporated into the

, emergency operating procedures and operator training has been completed under the revised procedures. The temperature inputs are from RTDs (1 cold leg and 2 hot legs) ranging from 212 degrees Fahrenheit to 705 de-grees Fahrenheit and the pressure inputs are from two pressurizer pres-sure sensors ranging from 15 to 3208 psia. These inputs are from safety related sensor Core Exit Thermocouples (CET)

The Calvert Cliffs incore instruments currently utilize Gulton connectors at the cable-to-incore instrument guide tube flange termination point These connectors were originally supplied under the NSSS scope of supply as safety related components. However, qualification for post accident service is required for use in the core exit thermocouple monitoring syste The licensee is replacing the Gulton connectors with a type that is already qualifie _ _ _ _ _ - _ _

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15 BG&E has determined that the incore instrument cables and containment j electrical penetrations are qualified for post accident, in containment applicatio During the current outage core exit thermocouple cabling, and hardware are being installed. Other plant computers and SPDS equipment are also being installed, however, will not be complete until the 1988 refueling outag BG&E has procured a new main frame computer system for installation at Calvert Cliffs which will replace the existing plant computers. Associated with the plant computer replacement project is the installation of a safety parameter display system (SPDS) utilizing the human factored CRT displa The schedule for completing these upgrades (data collection, processing, and display) is now during the 1988 refueling outage for each uni No unacceptable conditions 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 10 CFR Part 20 requirements were observe Independent surveys of radiological boundaries and random surveys of non-radiological points throughout the facility were taken by the inspecto During the Unit 1 outage the ALARA aspect of radiological controls appeared to be continuously evident, and factored into every evolution taking place within the controlled area boundary. The ALARA group has set a goal of 290 man rem for this ten year Ir Service Inspection outage. This appeared to be a very ambitious schedule, however, to date the outage has expended apprcxim-ately 150 man rem and it appears that the goal will be me @ No unacceptable conditions were identifie . Review of Periodic and Special Reports Periodic and special reports submitted to the NRC pursuant to Technical Speci-fication 6.9.1 and 6.9.2 were reviewed. The review ascertained: inclusion of information required by the NRC; test results and/or supporting information; consistency with design predictions and specifications; adequacy of planned corrective action for resolution of problems; determination whether any in-formation should be classified as an abnormal occurrence, and validity of reported information. The following periodic reports were reviewed:

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September and Octobei,1986 Operations Status Reports for Calvert Cliffs No. 1 Unit and Calvert Cliffs No. 2 Unit, dated October 15 and November 15, 1986, respectivel No unacceptable conditions were identifie _ - -

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1 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 inspection.

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