IR 05000382/1986016

Insp Rept 50-382/86-16 on 860801-31.Violations Noted:Failure to Properly Implement Fire Protection Program & Security Procedures & Inadequate Execution of Insp Program

ML20215G459
Person / Time
Site:	Waterford
Issue date:	10/10/1986
From:	Bundy H, Constable G, Luehman J, Staker T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20215G433	List: IR 05000382/1986016 ML20215G430 ML20215G442
References
50-382-86-16, IEB-86-002, IEB-86-2, NUDOCS 8610200381
Download: ML20215G459 (13)

Text

. .

APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-382/86-16 License: NPF-38 Docket: 50-382 Licensee: Louisiana Power & Light Company (LP&L)

317 Baronne Street P. O. Box 60340 New Orleans, Louisiana 70160 Facility Name: Waterford Steam Electric Station, Unit 3 (W3 SES)

Inspection At: Taft, Louisiana Inspection Conducted: August 1-31, 1986

_/ -

Inspectors:kJ.T.Lushman,SeniorResidentInspector

/o[c[e(

Date~

/

m /* /51 T. 'TE ~ 5 taker', Resident Inspector Dat Iobe d AR.~F.Bundy,ProjectInspector, Project Date'

Section C, Reactor Projects Branch (paragraphs 4, 10)

Approved: _

/0/bM/

G. L. ' Constable, Chief, Project Section C Date Reactor Projects Branch Inspection Summary Inspection Conducted August 1-31, 1986 (Report 50-382/86-16)

38A288aBebygga G

. _ _ _

-

-. __ -_ _

. .

-2-Areas Inspected: Routine, unannounced inspection of: (1) Plant Status, (2) Licensee Event Report (LER) Followup, (3) Monthly Surveillance, (4) Monthly Maintenance, (5) Routine Inspection, (6) IE Bulletins, (7) Preparation for Refueling, (8) Potential Generic Problems, (9) Safety Injection Tanks,'and (10) Followup of Previous Inspection Item Results: Within the areas inspected, the following' violations were identified:

(1) failure to properly implement fire protection program (paragraph 7),

(2) failure to properly implement security procedures (paragraph 7), and (3) inadequate execution of inspection program (paragraph 10).

.

-

- . _ _ - _ . _ _ _

-___ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ - - _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _ -

_

.. .

-3-

DETAILS Persons Contacted Principal Licensee Employees G. W. Muench, Director, Nuclear Operations R. P. Barkhurst, Plant Manager, Nuclear T. F. Gerrets, Corporate QA Manager

• S. A. Alleman,. Assistant Plant Manager, Plant Technical Services

• N. S Carns, Assistant Plant Manager, Nuclear, Operations and Maintenan'ce J. N. Woods, QC Manager A. S. Lockhart, Site Quality Manager R. F. Burski, Engineering and Nuclear Safety Manager K. L. Brewster, Onsite Licensing Engineer G. E. Wuller, Onsite Licensing Coordinator T. H. Smith, Maintenance Superintendent, Nuclear P. V. Prasankumar, Technical Support Superintendent

• Present at exit interview In addition ~to the above personnel, the NRC inspectors held discussions with various operations, engineering, technical support, maintenance, and administrative members of the licensee's staf . Unresolved Items An unresolved item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviatio A previously identified unresolved item is discussed in paragraph 1 . Plant Status

~

The inspection period began with the. plant at full power. During the afternoon of August 4, 1986, the licensee entered TS 3.0.3 when the results of samples on the safety injection tanks (SITS) indicated that three of the four tanks contained less than the minimum boron concentration required by TS 3.5.1. A plant power reduction was begun at 3:59 p.m. (CDT). At 7:10 p.m., with SIT boron concentrations restored to the proper band, the licensee began the return to full power from the 75 percent level that had been reached while complying with the TS requirement .

At 3:36 p.m. on August 21, 1986, with the plant at 100 percent power the licensee again entered TS 3.0.3 when sample results indicated that the boron concentrations in 2 of the 4 SITS were below the minimum required by the TS. A slow power reduction was begun at 4:00 p.m. and terminated at

.. .

.

._ ________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

e' a

_4_

5:20 p.m. when further sample results indicated the SITS were back in specification following'a drain and fil The plant was returned to 100 percent power at 7:07 p.m. The problems with SIT boron concentrations are discussed in detail in paragraph 11 of this report. With the exception of a power reduction for surveillance testing, the plant operated at or about full power for the remainder of the inspectio No violations or deviations were identifie . Licensee Event Report (LER) Followup The following LERs were reviewed and closed. The NRC inspectors verified that reporting requirements had been met, that causes had been identified, that corrective actions appeared appropriate, that generic applicability had been considered, and that the LER forms were complete. Additional _ly, the NRC inspectors confirmed that no unreviewed safety questions were involved and that violations of regulations or Technical Specification (TS) conditions had been identifie (0 pen) LER 50-382/86-14, Due to Inadequate Work Instructions, Maintenance Personnel Inadvertently Opened Safety Injection Tanks 18 and 2B Isolation Valves. The NRC inspectors performed a preliminary review of this LER to determine adequacy of corrective action. The licensee corrective action does not address the fact that in investigating the original event, the isolation valves were opened a second time approximately 25 minutes after the first occurrence at which time most of the fluid was injected into the reactor coolant system. The NRC inspectors informed plant management that repeating events during the investigation is unacceptable. The licensee plant manager agreed to examine the adequacy of their accident investigation training prior to closing out this LE (0 pen) LER 50-382/86-15, Simultaneously Using Two Methods of Draining Reactor Coolant Syste.n Results in Loss of Shutdown Cooling. The NRC inspectors discussed this report with licensee management. Of particular concern was the licensee's reference to average temperature readings of 192.5 F. With the reactor vessel in a partially drained condition and no flow through the core after the loss of the shutdown cooling flow average temperature readings are probably not extremely useful even though average temperature is the temperature parameter recognized by TS.' A better explanation of the event could have been rendered if the wide difference of temperatures between the two essentially separate systems that existed (the active core submerged in the relatively cooler water and the steam / nitrogen mix covered by the relatively hotter head) had been more thoroughly explaine (Closed) LER 86-13, Improper Connection of Control Element Assembly Resulted in Reactor Trip. The NRC inspector has verified that MI-13-523,

" Control Element Assembly and Incore Nuclear Instrumentation Connection and Verification," has been changed to require that each connector be verified locked following reconnection of the control element assemblie _

. .

-5-(Closed) LER 85-40, Loss of 4.16 KV Bus Resulting in Automatic Start of

-

EDG and Reactor Trip. The NRC inspector verified that in addition to replacing the undersized cup valve for the 8 main feedwater' pump, the licensee inspected-the corresponding valve for theA' main feedwater pump to ensure it was satisfactor ~No violations or deviations were identifie . Monthly Surveillance The NRC inspectors observed / reviewed TS required testing and verified that testing was performed in accordance with adequate procedures, that test' instrumentation was calibrated, that limiting conditions for operation (LCO) were met, and that any deficiencies identified were-properly reviewed and resolve During this inspection period the NRC-inspectors verified, reviewed, or witnessed portions of the following surveillance procedures:

. MI-3-126, " Core Protection Calculator Functional Test"

. MI-3-372, " Control Room Outside Air Intake Isolation Radiation Monitor Functional Test"

. MI-3-380, " Plant Vent Stack Safety Channel A Particulate and Gaseous Radiation Monitor Calibration" No violations or deviations were identifie . Monthly Maintenance Station activities affecting safety-related systems and components were observed / reviewed to ascertain that the activities were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with T During routine tours the NRC inspectors found the following measuring and test equipment (M&TE), which was past due for calibration, in use or available in the plant: August 6, 1986 - MIET 025.081 (calibration due date August 5, 1986) August 8, 1986 - MIPT 030.119 (calibration due date October 2,-1985) August 8, 1986 - MIPT 030.133 (calibration due date November 20, 1985)

These are considered further examples of Violation 382/8615-0 F

. .

-6-Activities observed during this inspection period included the shaft bearing adjustment on Emergency Feedwater Pump Work was performed per Condition Identification Work Authorization (CIWA) 02810 The NRC inspectors observed portions of the installation of TBC 19 and 23 and TBC 20 and 23 jumpers (part of auxiliary component cooling water pump start circuit) in Cabinets CP48 and CP49. Installation of.these jumpers was performed to conform with drawings PAC 8821D21 and 8821D2 The terminals were previously connected at the terminal strips. Work was performed per CIWA 2463 The NRC inspectors observed portions of ME-7-002, Revision 5, " Molded-Case

_

Circuit Breakers and Thermal Overload Relays," as performed on component cooling water makeup Pump "B".

No further violations or deviations were identifie . Routine Inspection By observation during the inspection period, the NRC inspectors verified that the control room manning requirements were being met. In addition, the NRC inspectors observed shift turnove" to verify that continuity of system status was maintained. The NRC inspectors periodically questioned shift personnel relative to their awareness of the plant condition Through log review and plant tours, the NRC inspectors verified compliance with selected TS and limiting conditions for operation During the course of the inspection, observations relative to protected and vital area security were made including access controls, boundary integrity, search, escort, and badgin On a regular basis, radiation work permits (RWPs) were reviewed and the specific work activity was monitored to assure the activities were being conducted per the RWP Selected radiation protection instruments were periodically checked and equipment operability and calibration frequency were verifie The NRC inspectors kept themselves informed on a daily basis of overall status of plant and of any significant safety matter related to plant operation Discussions were hel.d with plant management and various members of the operations staff on a regular basi Selected portions of

. operating logs and data sheets were reviewed dail The NRC inspectors conducted various plant tours and made frequent vi"'s to the control room. Observations included: witnessing work activities in progress; verifying the status of operating and standby safety systems and equipment; confirming valve positions, instrument and recorder readings, annunciator alarms; and housekeeping.

L

. .

-7-While in the control room the NRC inspector reviewed the response procedures for various annunciators associated with radiation monitor Operating Procedure OP-500-005, Revision 1, contains the annunciator response procedures for both " Circulating Water. Discharge Activity Hi" (A-19 on the E panel) and " Blowdown. Activity Hi" (B-19 on the E panel).

In each of these procedures control panel (CP) -14 is referenced as an indication location for the associated monitor but no such indication exists. These annunciator response procedures also reference CP-51 as an indication location. At one time CP-51 was a separate computer keyboard and CRT display. Subsequently, CP-51 was incorporated into CP-6 and no identification on CP-6 refers to CP-51. Response procedures for other annunciators associated with high activity (such as CCW for example)

reference CP-6 and not CP-51. It appears that references to CP-51 should be deleted as they may be misleading or confusing. Licensee operations personnel informed the NRC inspector that these two procedures are in the process of being revised and that the above comments will be incorporated in the change An inspection of the station batteries was conducted by the NRC inspectors and inconsistencies in the spacing at the end of the racks were observed in two of the three station battery rooms (3A and 3AB). It appeared that the adjustable end bars on the racks had been brought in as far as possible but a sizable gap between the racks and the cells existed. Th inspectors discussed these observations with the superintendent of technical services who initiated an evaluation of the situation. After measurements of the rack spacing were taken, the architect engineer (Ebasco Services, Inc.) as well as the battery manufacturer (Gould) were contacted. It was determined that the racks may not meet the design-specifications. The licensee has installed temporary shims to reduce the gap between the bars and the end cells in the rack The licensee has prepared a justification for continued operations and plans to restore the racks to the design standards during the first refueling outag Correction of the station battery spacing problems is identified as open item 382/8616-0 Violation 382/8528-02 concerned maintaining improper level in the wet mechanical draft cooling tower water basins due to improper calibration of the level indicators. When this violation was issued the licensee was requested to consider the possibility that such calibration errors could exist in other instruments important to plant safety. The response stated that "there is no reason for LP&L to believe that this concern would be a generic problem." Recently, the licensee has written two potentially reportable event (PRE) reports concerning inconsistencies in data used for level indication for the refueling water storage pool (RWSP) and the safety injection tanks (SITS). Though neither of the potential problems is identical to that found with the wet mechanical draft cooling tower water basin level indication, they are both very similar. The NRC inspectors discussed this concern with licensee management and suggested a rigorous inspection of data for other level indications important to

-

_ - - . _ __ ,_ _ _ - _ , .__ _ _ . . _ _ _ _ _

.

. .

-8-safety be made so that future in'dividual problems are avoided. Such an evaluation was already in progress and preliminary-results indicated no further. problems with tanks important to safet During a routine tour of the plant the NRC inspectors observed a total of:

six safety cans, each of which were at least partially full of combustible -

liquids, left unattended in areas containing safety-related component This is an apparent violation and is identified as 382/8616-02. The NRC inspectors have, during a number of previous inspections, made plan management aware that.the requirements of FP-1-017, Revision 2, " Transient Combustibles and Hazardous Materials," for placing unattended combustible liquid containers in designated cabinets (Sections 6.1.18 and 6.3) were not being strictly followe During this inspection the cans observed were in areas containing the main steam and main feed isolation valve Because of the continual need to add fluid (hydraulic fluid is defined as a combustible liquid by the licensee fire protection program) to valve hydraulic systems, the presence of some cans continues to be a proble During a tour of the control room on August 25, 1986, the NRC inspector noticed the trouble pushbutton on the front of Plant Protection Channel "A" illuminated. (The panel is located behind the main control boards.) The NRC inspector questioned one of the onduty licensed operators about the trouble light but he was unaware that it was illuminated. The operator subsequent 1y' cleared the light and explained that i.t was most probably left illuminated following testing of the plant protection system the previous da The NRC inspectors have observed this same condition on a number of control room tours during the past few months and have discussed the problem with licensee management. The existence of such lighted trouble indicators over a number of shifts indicates that proper review of plant indicators in the control room is not routinely being conducted by the operator Licensee management.had told the NRC inspectors that review of the applicable procedures would be done to ensure that the operator checks tn see that the trouble lights are out following performance of testin Following this occurrence of the problem, licensee management stated appropriate corrective action will be considered immediately. Although this problem reflects poor operating practices, it does not appear to be safety significant because a trouble condition of the Plant Protection Channels should also be annunicated at the main control boar The NRC inspector performed a comprehensive walkdown of the control panels on August 27, 1986. Two problems were discovered during this revie First, neither indicating light for SI-323B (SIT 1B vent) was illuminate Second, neither indicating light for CMU-131 (makeup water to RWSP) was illuminated. When questioned about the status of the lights and the corresponding valves, neither of the nuclear plant operators (NP0s) or the control room supervisor (CRS) was aware of any abnormal status. In both

'

cases the valves were verified to be in their proper positions (closed).

In the case of SI-323B, the green " closed" light circuit appeared to be inoperable and a CIWA was initiated. The breaker for CMU-131 was

-. - - . _ . - - - -

-. -_ _ --- . ,___ _--

- __ _ -_- _ _- _ _ _ ___- __ _ _ _ _ . __

..

. . .

..

.. .

-9-t determined to be open, explaining the lack of valve indicating light The breaker had been intentionally left open because that path is not normally used to fill the RWSP and operations personnel wanted to minimize the possibility of an inadvertent RWSP dilution by opening the valve. A caution tag was placed across the valve indicating lights explaining the valve / breaker status. The NRC inspector has discussed the above problems with~ indicator status and shift turnover of such conditions with plant managemen On August-29, 1986, during a control room tour the NRC inspector observed material not related to the job being read in the secondary alarm station (SAS). The NRC inspector promptly informed the shift supervisor who informed the security shift supervisor. The security shift supervisor had the officers involved relieved while an investigation was conducte PS-90-101, Revision 3, " Security Department Organization, Duties, and Responsibilities," which in part implements the physical security plan details the duties and responsibilities of both the secondary alarm station console operator (SASCO) and the security officer (50). The NRC inspector informed the licensee that the reading of material not related to the job was not consistent with carrying out the requirements of PS-90-101 which is an apparent violation and is identified as 382/8616-0 . IE Bulletins (Closed) IEB 86-02, " Static 'O' Ring Differential Pressure Switches." In a letter dated July 25, 1986, the licensee notified the NRC that their investigation had determined that no differential pressure switches of the type described in the bulletin were installed or planned for use at W3 SE No violations or deviations were identifie . Preparation for Refueling During this inspection period the NRC inspectors reviewed the plant procedures in place to support maintenance, surveillance, and operations of the reactor during refueling. Within the scope of this review the NRC inspectors examined more than 60 procedures in the areas of operations, electrical maintenance, mechanical maintenance, instrumentation and control (I&C), and engineering. NUREG/CR-1369, Revision 1, " Procedures Evaluation Checklist for Maintenance, Test, and Calibration Procedures,"

was used by the NRC inspectors as a guidance document. Overall, the licensee's procedures in place to support refueling appear adequate. The NRC inspectors had the following comments based on their review: OP-2-006, Revision 6, " Fuel Pool Cooling and Purification System,"

needs to include the new heat exchanger presently being installed by Station Modification (SM) 10 OP-10-002, Revision 1, " Refueling," Step 4.12 only requires one startup nuclear instrument channel to have visual indication in the

. - .

.. .. _ _ - . _ _ _ _ ._ a

.. .

- 10'-

control room which is not consistent with TS 3.9.2 which requires two startup channels to have continuous visual indication in the control roo c. OP-901-27, Revision 3, " Fuel Handling Incident," is improperly referenced as OP-901-47 in Step 4.9 of OP-2-01 d. OP-2-006, Revision 6, " Fuel Pool Cooling and Purification system,"

has no annunciators referenced in the setpoint section of the procedure but applicable annunciators exist above CP- e. MI-3-364, Revision 2, " Fuel Handling Building Ventilation Airborne Isolation Radiation Monitor Safety Channel 'A' Calibration;"'

MI-3-365, " Fuel Handling Building Ventilation Airborne Isolation Radiation Monitor Safety Channel 'B' Calibration;" and MI-5-513, Revision 1, " Fuel Handling Building Emergency Filtration Unit E-35 Inlet Damper D29 Control Calibration," all reference the nuclear operations supervisor (N0S), a position that no longer exist f. There is inconsistency in the labeling of procedural hold points. In MM-08-18, Revision 2, " Removal of Upper Guide Structure," (as in most procedures) they are called " Inspection Hold Points." In MM-08-019, Revision 2, " Removal of Core Support Barrel," they are called

" Quality Hold Points," while in Attachment 10.1 to MM-8-13, Revision 3, " Reactor Vessel Head Removal," they are referred to as

" Independent Verification." Finally, in Attachment 10.1 to MM-8-028, Revision 3, " Reactor Head Stud Installation," hold points are labeled

"QC Witness."

g. MM-8-028, Revision 3, calls for a hold point for Step 8.4.7; however, on the procedure checklist (Attachment 10.1) the hold point is indicated for Step 8. h. MI-5-569, Revision 4, " Fuel Handling Building Emergency Filter Train A or B Inlet Air Temperature Loop Calibration," Section states that both the "Lo Temp Diff Alarm" and "Lo Temp Diff Fan Interlock" have the same setpoint of 7 F. The interlock setpoint should be 5 i. MI-3-370, Revision 2, " Fuel Handling Building Ventilation System Emergency Exhaust High Range Noble Gas Radiation Monitor Channel Functional Test," and MI-3-371, Revision 2, " Fuel Handling Building Ventilation System Emergency Exhaust High Range Noble Gas Radiation Monitor Channel Calibration, each contain two data gathering steps (8.3.2, 8.3.5 and 8.8.22, 8.8.25 respectively) which are followed by procedural notes concerning stabilization prior to collecting and recording the dat j. A comparison of NE-1-001, Revision 4, "New Fuel Shipping Container Operations," to NE-1-004, Revision 1, " Control Element Assembly Receipt, Intpection, Storage, and Docurrentation," shows that NE-1-001

.

+

,. .

-11-is much more detailed as to exactly how the newly arrived materials get into storage in the fuel handling building. Additionally, though a mention of security inspection upon arrival at_the gate is made in NE-1-004, both procedures need to further reference security to ensure proper control of fuel. handling building doors and #1oor plugs is maintained while transfer of the materials takes plac HP-2-220, Revision 2, " Health Physics Support of Fuel Receipt,"

requires the responsible technician to inform the health physics supervisor "as soon as practical" if, for example, contact readings of greater than 200 mr/hr are detected on the newly arrived fuel shipping containers. UNT-06-010, Revision 1, " Event Evaluation and Reporting," classifies such ar. occurrence as a one-hour report pursuant to 10 CFR 20.205. W erefore, "as soon as practical" should be clarified to ensure the shtft supervisor is informed soon enough to make any required report The followup of the correction / inclusion of these comments is identified as open item 382/8616-0 No violations or deviations were identifie . Potential Generic Problems o

The NRC inspectors provided the licensee a copy of 10 CFR Part 21 report on cracking of Limitorque limit switch rotors issued by Sacramento Municipal Utility District (SMUD) on June 10, 1986. SMUD found cracks on two rotors out of ten motor operators inspected. The cracking problem appears to be limited to white (melamine) rotors in that no cracks have been identified in brown (fibrite) rotors currently being supplied by Limitorque Corporatio Because of problems identified at the Dresden facility, the NRC inspectors inspected the anchorage of a number of control and switchgear panels. At W3SES, the majcrity of the panels are welded to plates embedded in the concrete floors. Most of the welds for control room panels are currently inaccessible. Therefore, the NRC inspectors confirmed that installation inspection records had been completed. The anchor welds for panels in other rooms were physically inspected by the NRC inspectors and appear to conform to installation drawings with the following exceptions: No installation inspection records could be identified for the electric hydrogen recombiner power supply Panel B anchor weld Because of epoxy paint and general inaccessibility of the required welds, neither the licensee quality control (QC) inspector nor the NRC inspectors could physically verify their existence. The licensee subsequently uncovered the joints and discovered that the front and rear welds had not been installed. The licensee prepared a Justification for continued plant. operation and initiated action to install the required weld .. .

-12- No installation inspection records could be identified for 120v nuclear instrumentation Static Uninterruptible Power Supply (SUPS)

3 MB The torque inspection for anchor bolts was marked "not applicable." However, it appeared to the licensee QC inspector and the NRC inspectors that the unit may be anchored only by expansion bolts in that no anchor welding could be positively identified because of epoxy paint and general inaccessibility of the welds, if they existed. Also, a Field Change Request (FCR) E-630. existed which allowed deletion of the nut on the anchor bolt on the rear of panel adjacent to the wall. This would leave seven torqued anchor bolts for the unit. The licensee QC inspector and NRC inspectors observed that, for the remaining seven anchor bolts, the two nuts on the front half of the unit adjacent to the wall appeared to not have adequate thread engagement. The licensee's investigation determined that the panel required only bolting and inspection records were availabl The licensee further determined that one of the nuts in question did not have adequate thread engagement. The licensee committed to reperforming the seismic analysi The failure to perform an adequate installation inspection for this equipment is an apparent violation (382/8616-05).

No further violations or deviations were identifie . Safety Injection Tanks (SIT)

On August 4, 1986, and again on August 21, 1986, the licensee had to. enter

.TS 3.0.3 (limiting condition for operation is not met) because the boron concentrations in more than one SIT dropped below the minimum required by TS 3.5.1. The licensee attributes these situations to tank stratification and the inability to mix the tank before sampling. Further, because or the proximity of the fill / drain lines for the SITS to their. respective sample connections, after additions to the tanks, samples routinely indicate a concentration very close to that of the water added. Again, this problem is also influenced by the inability to mix the tank In order to minimize the problems with the SIT boron concentration, the licensee now samples both the refueling water storage pool (RWSP) and the recirculation line on the high pressure safety injection (HPSI) pump to be used for the addition, before each additio The NRC inspectors have reviewed the problems and the licensee's action Though tank stratification, inability to thoroughly mix the tanks, and the physical relationship between the addition and sample lines do appear to be problems, there appear to be steps the licensee can take to minimize the problems. First, TS 3.5.1 allows SIT boron concentrations between 1720-2300 ppm; yet the licensee fairly consistently operates with the boron concentrations below 1900 ppm. If concentrations in the range of 1900-2100 ppm were maintained, the effects of stratification or obtaining satisfactory sample results should be minimized. Secondly, remote sampling capabilities for the SITS are needed. Because the SIT samples

--

-

n 1

.

. . . .

.

"

-13-

_'

have to be obtained by entering the' containment at power, licensee personnel tend to defer increasing both boron concentration and level as-long as possible because additional containment entries would be required-to verify additions. Routine 31 day boron samples.taken on two of the SITS on_ July 21, 1986, were-1724 ppm and 1726 ppm, yet no action was taken-to increase the concentration further above the minimum concentration of 1720 ppm even.though the tanks would potentially go another month without a' sample. On a number of control room tours the NRC~ inspectors have questioned licensee _ operations personnel about operating with an SIT low:

level alarm illuminated. The operators indicated that they realized alarms were in and that they had verified the level was above that required by TS 3.5.1. When-the NRC inspectors, asked why action wasn't taken to clear the alarm the response was that they were waiting until borated water has to be added to the SIT to keep it above the TS minimum level or addition to more than one-SIT is necessary. . Again, these were

~

efforts to minimize containment entries because TS 4.5.1.b requires a tank sample any time a 1 percent volume increase is made. The above observations have been discussed with licensee maaagemen No violations or deviations were identifie . Followup of Previous Inspection Items (Closed) Violation 382/8602-04, " Failure to Specify and Record Results of Required Operability Testing Performed Prior to Returning Chiller Unit A to' Service." The NRC inspector has reviewed the licensee's response'to this violation which is contained in a letter dated May 16, 198 _

Additionally, OP-100-010, Revision 2, " Equipment Out of Service," has been reviewed to verify the changes to the procedure properly implement the licensee's specified corrective action ~

(Closed) Violation 382/8613-02, " Failure-to Take Prompt Corrective Action-on Identified Valve' Wiring Problems." This specific violation is

-

considered closed, however, the larger issue of environmental qualification of Waterford 3 Limitorque operators for motor operated valves will be addressed by Unresolved Item 382/8615-0 (Closed) Deviation 382/8613-03, " Failure to Properly. Implement Response to IE Bulletin 85-01,_' Steam Binding of Auxiliary'Feedwater Pumps'." -The NRC inspector has reviewed the licensee's response to this deviation which is contained in a letter dated August 22, 1986. Further, the NRC inspector has reviewed OP-903-001, Revision 3, " Technical Specification Surveillance Logs," to ensure the described changes adequately resolved the issu . Exit Interview i

The inspection scope and findings were summarized on September 4, 1986, l with those persons indicated in paragraph 1 above. The licensee

, acknowledged the NRC inspectors findings. The licensee did not identify

'

as proprietary any of the material provided to or reviewed by the NRC inspectors during this inspection.

-

I

, s L_