Text

Insp Repts 50-498/97-06 & 50-499/97-06 on 970810-1004. Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support
	ML20198F025
Person / Time
Site:	South Texas
Issue date:	12/08/1997
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20198E979	List: IR 05000498/1997006; ML20198E973; ML20198F016;
References
	50-498-97-06, 50-498-97-6, 50-499-97-06, 50-499-97-6, NUDOCS 9801090283
	Download: ML20198F025 (53)
	v • d • e

. . _ - . - . . . .. . . - . ~. . -. - .. . - . _ - . - .- - -. - _

!

.

,

.

j

.

ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

,

j i

.

Docket Nos: 50 498; 50 499 :

i

.

License Nos: NPF 70; NPF 80 [

i Report No: 50 498/97 06; 50-499/97-06 ,

Licensee: Houston _ Lighting & Power Company j

,

Facility: South Texas Project Electric Generating Station, i Units 1 and 2

,

-

Lonation: 8 Miles West of Wadsworth on FM 521-Wadtworth, Texas 77483 i

Dates: Augue,t 10 through October 4,1997 2~

inspectors: D. P. Loveless, Senior Resident inspector i J. M. Keeton, Resident inspector

W. C. Sifre, Resident inspector R. A. Kopriva, Project Engineer Accompanying Personnel: J. Edgerly, Resident !nspector Trainee l Approved by: J. l. Tapia, Chief, Project Branch A Division of Reactor Projects

J h

. -

-. ,- = - . . . - - . . . . . - - - - - - - - _ . - . - - - . ., ; - , , , - - , - _ - . . - - -

,

l l

'

.

l

!

EXECUTIVE SUMMARY South Texas Project, Units 1 and 2 NRC Inspection Report 50-498/97 06; 50-499/97 06 This resident inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers an 8 week period of resident inspection.

Operationg

In general, licensed operators were continuously aware of plant conditions and fully understood control panelindications (Section 01.1)

The operator's response to a lightning strike was excellent. The effects and potential impacts to plant equipment were thoroughly evaluated (Section 01.1).

Controls implemented to ensure continued core cooling throughout reduced inventory operation were excellent. Contingency and compensatory actions for midloop operations were well planned and in place (Section 01.2).

The use of an equipment clearance order caution tag to implement a change to a procedurafzed valve alignment resulted in the draining of the volume control tank to the low level auto makeup alarm setpoint. This was considered an example of a Technical Specification 6.8.1 violation (Section 01.3).

A licensed reactor operator f ailed to observe that the volume control tank level was decreasing following system valve manipulations (Section 01.3).

The f ailure to have written procedural controls covering a spent fuel pool filter replacement resulted in the inadvertent draining of approximately 3 inches of water from the spent fuel pool. This was considered a second example of a violation of Technical Specificathn 6.8.1 (Section 01.4).

Inadequate self verification techniques resulted in two minor examples of plant workers manipulating components in the wrong unit (Section 01.5).

An inadvertent modo change from Mode 4 to Mode 3 was caused by insufficient residual heat removal system flows to maintain constant reactor coolant system temperature during system initiation. The mode change was in violation of plant operating procedures. This nonrepetitive, licensee identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NBC Enforcement Policy (Section 01.6).

Licensee management was properly assessing and evaluating the impact of inoperable automatic functions on operator performance (Section 01.7).

Refueling and Equipment Outage 1REO7 was well run and controlled with a proper perspective on nuclear safety and shutdown risk (Section 02.1).

-- - .. - _ _ - . - . - . _ - - .- - - _ _ - - _ . -_-

!

., . !

,

+

2 t B

e Licensed operators were appropriately aware of the status of main control board .

instrumentation throughout the instrument normalization proces. Jtilized during Refueling Equipment Outage 1REO7 (Section 03.1).

o Licensed operators f ailed to ensure that the most restrictive Technical Specification !

was followed upon removing a station vital battery from service. This '

-

nonrepetitive, licensee identified and corrected Technical Specification violation is i

_ being treated as a noncited violation, consistent with Section Vll.B.1 of the E Enforcement Pollev (Section 08.2).

, e A violation was identified for the f ailure to verify certain containment penetration

'

isolation valve positions once per 31 days as required by Technical Specifications -

(Section 08.4).

e Licensed operators were cautious and conservative in implementing corrective ,

'

l actions for a bound control rod drive mechanism. These included a reac*or vessel

_

,

head lift and withdrawing Contro! Rod H 2 (Section E2.1). i Maintenanen ,

e Observed maintenance activities were well coordinated and implemented by knowledgeable technicians with appropriate levels of supervision and good support .

from engineering and operations personnel (Section M1,1). !

e The surveillance activities observed were performed in accordance with the applicable Technical Specifications. With one minor exception, testing was well controlled and properly implemented (Section M1.2).

o Independent calculations indicated that the component cooling water heat exchanger fouling f actor had been appropriately determined (Section M2.1). *

e The f ailure to maintain the outer containment personnel airlock door locked while the inner door seal was inoperable was in violation of Technical Specification 3.6.1.3. This nonrepetitive, licensee identified and corrected v!olation is being treated as a noncited violation, consistent with Section Vll B.1 of the E Enforcement Pollev (Section MB.2),

e A steam generator pressure instrument was found to be inoperable in violation of i Technical Specification 3.3.2. This nonrepetitive, licensee identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev (Section M8.3).

o Containment inspections conducted in accordance with Technical Specification

-

requirements were wellimplemented and provided additional assurance of

-

containment sump operability (Section M8.4), 4

_

_ __ _ _ _ __ _ ,_- u. _ ,. . ~ . __ _ . ... _._ _ - _

.

,

P

?4 3-

'

.

c , b e.meerina

$

..

.

Residua' heat removal system pump low amperage annunciators installed for (

, midloop operations protection had never been calibrated as recommended by the vendor. However, calibration activities indicated that the instruments had ternained within the required tolerances (Section 01.2).

In general, refueling outage related commitments in the engineering functional area had been properly implemented and procedurally controlled. However, the licensee had not completed internalinspections of standby diesel generator heat exchangers since 1993 (Section E1.1).

The analyses and practices related to spent fuel pool cooling and clean up systems prepared for Unit 1 Refueling and Equipment Outage 1REO7 properly verified that poolloading would remain within the licensing basis assumptions (Section E1.2).

Licensee engineers were conservative in the review and development of corrective actions related to a bound control rod drive mechanism. The unreviewed safety question evaluation properly assessed the evolution being proposed (Section E2.1).

The licensee's decision to perform service testing of station vital batteries at the f actory was nonconservative and did not meet the requirements of 10 CFR Section 50.59. However, upon discussic..s with the inspectors and other NRC personnel, the licensee performed service testing of the batteries after installation at the site in Refueling and Equipment Outage 1REO7 (Section E2.2).

A review of administrative controls over pressurizer heaters while the cold overpressure mitigation system was out of service disclosed that licensed operators failed to maintain the pressurizer backup heaters inoperable while water solid conditions existed in the reactor coolant system in January 1994. Current procedural controls governing systems that could result in a reactor coolant system mass or temperature increase remain limited (Section E3.1).

The failure to document a written safety evaluation of the installation of the advanced liquid waste processing system in September 1992 was a violation of 10 CFR 50.59. This failure also resulted in several deviations from Updated Final Safety Analysis Report commitments not being properly addressed, in addition, associated examples of licensing commitments not being implemented in the plara were noted. A further review of the system design will be necessary to evaluate its adequacy and acceptability (Section E8.2).

Bent Suonort

The radiological protection efforts related to divers entering the spent fuel pool were excellent. Other observed radiological protection performance was strong (Section R1.1).

!

,

. l l

The Technical Support Centers and Operations Support Centers in both units were readily available and maintained for emirgency operation (Section P2.1).

Daily security force activities were conducted in an appropriate manner (Section S1.1).

,

l

!

..

,

l

i (

.hble of Contents REPORT DETAILS ................................................. 1 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 Control Room Observations (Units 1 and 2) ....................1 O *i . 2 Reactor Coolant System Midloop Oparations (Unit 1) . . . . . . . . . . . , . . 3

,

01.3 Inadvertent Drain of the Volume Control Tank (Unit 1) . . . . . . . . . . . . . 5 01.4 Inadvertent Drain of the Spent Fuel Pool (Unit 2) . . . . . . . . . . . . . . . . . 7 01.5 Two Examples of Work Performed in the Wrong Unit (Unit 1) ........9 01.6 Inadvertent Mode Change (Unit 1) .......................... 10 01.7 Review of Inoperable Automatic Functions (Unit 1) ..............12 O2 Operational Status of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . 12 02.1 Pla nt Tout s (Units 1 and 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . 13 03.1 Operator Knowledge Related to the Status of Normalized Instruments . 13 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 08.1 Violation 50-498/96003-02: Out of Calibration Condition .........14 08.2 Licensee Event Report 50 498/96-001: Operability of St andby Die sel G enerator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 08.3 Inspection Followup Item 498;499/9303136: Continuous Management Representation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 08.4 Unresolved item 498:499/97005 02: Containment Penetrations not Surveilled ................................ 16 08.5 Violation 498/96004 04: Midloop Level Transmitters ............17 08.6 Violation 499/9700101: Level Sight Glass . . . . . . . . . . . . . . . . . . . 17 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 M 1.1 General Comments on Field Maintenance Activities . . . . . . , . . . . . . . 18 M1.2 General Comments on Surveillance Testing . . . . . . . . . . . . . . . . . . . . . 18

-.

_ . _ _ _ . _ . . _ . . . _. ___ _ .. .____.-_ _ . _ _ . . _ _ _ _ . _ _

>

,

.. .

,

I 2- !

i I

M2 Maintenance and Material Condition of Facilities and Equipment 20

. . . . . . . . . . .

-

!

M2.1 Component Cooling Water Heat Exchanger Performance Testing . . . . . 20 i i

i M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 l

!

M8.1 Licensee Event Report 50-439/97 006: Manual Reactor Trip . . . . . . . . -

21 l l

M8.2 Licensee Event Report 50 498/96-005: Personnel Airlock Inoperable . 21 l

i M8.3 Licensee Event Reports 50-499/96 003: Improperly !

'

Inst alled Jumper . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

!

M8.4 Violation 499/97002 01: Inadequate Containment inspection . . . . . . . 22 !

M8.5 Licensee Event Report 50 499/97 003: Loose Debris in C ont ainm e nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .- 24 ;

!

E Implementation of Outage Related Cornmitments . . . . . . . . . . . . . . . . 24 :

.

E1.2 bpent Fuel Pool Heatup Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 :

?

E2 Engineering Support of Facilities and Equipment ..................... 27 !

E Binding of Rod Control Cluster Assembly (Control Rod) during Rapid Ref ueling . . . . . . . . . . . . . . . . ........................... 27

.

E2.2 Station Vital Battery Bank Service Test Review . . . . . . . . . . . . . . . . . 29 t E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. 31 E Inspection Followup item 498;499/9320211: Backup i Pressurizer Heaters Not Controlled . . . . . . . . . . . . . . . . . . . . . . . . . . 31 E8.2 Unresolved item 498;499/96004 03: ALPS was not Described

. in t h e U F S A R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 E8.3 Unresolved item 498;499/96006 04: Seals for Spent Fuel Po ol G a t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 R1- Radiological Protection and Chemistry Controls . . . . . . . . . . . . . . . . . . . . . . 40 RI .1 Tours of Radiological Controlled Areas . . . . . . . . . . . . . . . . . . . . . . . 40 I

.t h

.,- - . - ,,-. , .,. - _ ...-- - -.- -u - -~., - _ _ .e _ , -.4,. - _ .. ._., . - - . , . - ~ - - .

_ _ . _ _ _ ____ _ ._ .. . . _ _ _ _ _ . _ - -

,

.. <

R7 Quality Assurance in Radiological Protection and Chemistry Activities . . . . . . 41 R7.1 Transportation of Contaminated Machine Tools . . . . . . . . . . . . . . . . . 41 P2 Status of EP Facilities, Equipment, and Resources . . . . . . . . . . . . . . . . . . . . 41 P Emergency Response Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 S1 Conduct of Security and Safeguardo Activities ...................... 42

S Daily Physical Security Activity Observations .................. 42 X1 E xit M e eting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42

- . . . ,. . -

- _ . _ ._ _ . . _ . ~

. . .

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

.

'

,

.. .

f Renort Details

!

t Summarv of Plant Status .

.

At the beginning of this inspection period, Unit 1 was operating at 100 percent reactor power. On September 13, the unit was removed from service and the reactor was shut i down for Refueling and Equipment Outage 1REO7. On October 4, the unit was returned to service, ending Outage 1REO7, At the end of this inspection period, the Unit i reactor ,

was operating at 29 percent power with power escalation to 100 percent in progres .

Unit 2 operated at essentially 100 percent reactor power throughout this inspection perio LDoerationt 4 01 Conduct of Operations ,

01.1 Control Room Observations (Units 1 and 21 Insoection Scone (71707)

Using Inspection Procedure 71707, the inspectors routinely observed the conduct

_

of operations in the Units 1 and 2 control rooms. Frequent reviews of control board status, routine attendance at shift turnover meetings, observations of operator performance, and reviews of control room logs and documentation were performe in addition to full power operations, the inspectors observed portions of the following evolutions:

I e Unit 1 response to a lightning strike on site. (9/4)

e Shutdown of Unit 1 for Refueling and Equipment Outage 1REO7. (9/12) ;

e Control rod drop testing in Unit 1 as documented in Section M1.2 of this report. (9/13)

e Midloop operations in Unit 1 as documented in Section 01.2 of this repor (9/16 and 9/29) l

)

e Control room operator implementation of Plant Surveillance )

Procedure OPSP03 DG 0022, Revision 0, " Standby DieselInterdependence Verification" in Unit 1. (9/21)

e Unit 1 initial criticality following completion of Refueling and Equipment Outage 1REO7, (10/3) Qhservations and Findinas During routine observations and interviews, the inspectors determined that, with one exception, the control room operators were continually aware of existing plant l

l

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

,

.. -

conditions, in one instance a licensed opeiator f ailed to notice a decreasing volume control tank level caused by an inadvertent drain down until a low level annunciator alarmed. This event was documented in Section 01.3 of this report. Operators responded to annunciator alarms in accordance with approved procedure Annunciator alarms were promptly announced to the control room staff who, in turn, acknowledged by restating the announcement. The inspectors routinely observed shift turnover and attended shift turnover meetings. The on shift operators provided clear and concise information to the oncoming operator Oncoming operators routinely reviewed the control room logs, discussed current ,

plant conditions, and verified major equipment status during control board walkdown On September 4, Unit 1 was struck by lightning during a strong ,

thunderstorm. As a result, a 120 volt nonvital ac bus that supplied power to several radiation monitors and other instruments experienced a momentary ground fault. The inspectors responded to the Unit 1 control room and observed the operators' rosponse to the event. The voltage perturbation caused multiple annunciators to alarm in the Unit I control room. The momentary loss of power to the affected radiation monitors also caused them to enter an alarm state that, in turn, caused an automatic containment ventilation isolation actuation. The containment ventilation isolation caused reactor coolant system letdown to isolate. Reactor coolant system charging automatically stopped in response to the letdown isolation. The control room operators recognized the condition, promptly determined the cause, and restored charging and letdown within 12 minutes. The shift supervisor maintained good command and control as operators verified equipment status and critical parameters. The control room staff had good support from reactor plant operators who were verifying system condition. The operations manager, maintenance manager, and system engineer also responded to the control room and provided excellent support. An inspection and verification plan was developed to ensure that plant equipment was either unaffected or returned to an operable conditio On Septamber 12 and 13, the inspectors observed the Unit 1 shutdown and cooldown to Mode 3 for Refueling and Equipment Outage 1REO7 The inspectors ;

observed reactor operator trainees performing reactivity manipulations and removing main feedwater pumps from service with direct licensed operator oversight. The shutdown and cooldown was performed in cautious, controlled manner in accordance with approved procedure On September 21 the inspectors observed the Unit 1 control roem staff performing Plant Surveillance Procedure OPSP03 DG 0022. Revision 0, " Standby Diesel Interdependence Verification." This test involved the simultaneous start of all three standby diesel generators to ensure that the operation of a standby diesel generator did not interfere with the operation of either of the other standby diesel generator This test is performed on a 10-year frequency and this was the first performance

1

_ ______ _ _

.. . .. . .

,

.. <

3-since initial startup of Unit 1. The test was performed in accordance with the procedure and with good radio communication between the control room and the reactor plant operators located at each standby diesel generator.- The shift supervisor provided proper oversigh On October 3, the inspectors observed licensed operators withdraw control rods and dilute the reactor coolant system to achieve criticality. The evolution was well controlled with excellent communications between nuclear engineers and the .

operators. Reactor operator trainee's who were present were given appropriate levels of oversight by licensed operators, Conclucions in general, licensed operators conducted their duties in a professional manne Transfer of information at shift turnover was good. The response to a lightning-strike in Unit I was excellent. Control and instruction of reactor operator trainees were excellent. Operational procedures were properly implemented during

- reactivity manipulations and testing evolution .2 Reactor Coolant System Midlooo Ooerations (Unit 11 insoection Scoce (71707,92701)

^

The inspectors performed a review of the licensee's nractices and controls governing reduced inventory operations. The purpose of this inspection was to ensure that effective controls were in place prior to and during midloop/ reduced inventory operations planned for the seventh Unit i refueling outage. Procedures, planning, training, and oversight were evaluated prior to reduced inventory operations. Onsite NRC coverage was maintained throughout reduced inventory

. operations, in addit.un, licensee's corrective actions and lessons learned from three previous events were reviewed at this time. The following documents were reviewed during this inspection:

o Plant Operating Procedure OPOP03 ZG-0007, Revision 15, " Plant Cooldown" e Plant Operating Procedure OPOPO4 RH 0001, Revision 7, " Loss of Residual Heat Removal" e Plant Operating Procedure OPOP03 ZG 0009. Revision 17, "Mid Loop Operation" e Generic Letter 8817, " Loss of Decay Heat Removal" e Plant General Procedure OPGP03 ZO 0035, Revision 8, " Reduced RCS Inventory Operations"

.

. _ _ _ _ _ _ _ _ --

,

.. -

Plant Operating Procedure OPOP03 ZG-0010, Revision 15, " Refueling Operations"

Condition Report 9714481, " Duty Operations Manager not in place during mid loop operations" Observations and Findinai The inspectors reviewed the procedures and documentation related to the evolution. The procedures had been revised in preparation for the evolution. The inspector noted that these procedures implemented the corrective actions for Violations 498/90004-01 and 499/9700101. A midloop manager had been assigned to provide oversight of the preparations for reduced inventory operation The inspectors feviewed the contingency and compensatory actions planned and determined that they were appropriat From September 1517 nd again from September 29-30, the inspectors provided around the clock coverage of reduced inventory evolutions. The associatad procedures were properly implemented from the control room, and draining activities were performed in a deliberate and controlled manner Correlation between reactor vessellevelinstrumentation was continuously evaluated. All instrumentation functioned as expecte The inspectors reviewed the licensee's commitments related to Generic Letter 8817. Alllicensed operators and additional personnel had been properly trained. Allinstrumentation was in place and functional. The inspector questioned the operability of the residual heat removal system pump amp annunciators. Af ter a review, licensee personnel determined that the annunciators had never been calibrated. A calibration was performed, and the instruments were within tolerance. A preventive maintenance task was then developed to calibrate the instruments in accordance with the vennor recommended schedul The inspectors performed a walkdown of accessible portions of the reactor coolant system level sight glass and the hot leg narrow range water levelinstruments. The instruments were in the appropriate configuration and alignment for midloop operation in accordance with the reviewed procedures. The inspector noted that the temporary manifold, used in previous outages in the upper sight glass connection, had been replaced, and the vent valves that had previously contnbuted to a loss of indication had been eliminated. During this walkdown, the inspector also verified the prestaging of the residual heat removal system vent rig and necessary tools for installation. The inspector interviewed the reactor plant operator who was responsible for installing the vent rig, if necessary. He stated that the rig would be requirea if vortexing was indicated in the residual heat removal system. The operator was cognizant of his duties and familiar with the immediate actions necessary for venting the residual heat removal syste ~ ~. - _ ,

.. -

)

I In general, management oversight of reduced inventory activities was excellent. A management structure designed for additional support while the reactor coolant system was at midloop was in place and functioning. However, on one occasion, an operations manager was not delegated the responsibilities for onsite coverage, as required by Procedure OPGP03 ZO 0035. Condition Report 9714481 was written to document and correct this proble Conclusions Controls irnplemented to ensure continued core cooling throughout reduced inventory operations were excellent. Contingency and compensatory actions were planned and in place. Alllicensee commitments to Generic Letter 8817 were being implemented. However, the inspectors identified that the residual heat removal system pump low amperage annunciators had never been calibrated.

01.3 Inadvertent Drain of the Volume Control Tank (Unit M Insoection Scone (71707)

On August 20, approximately 550 gallons of water were inadvertently drained from the volume control tank to the spent resin storage tank in Unit 1 when a reactor plant operator opened the wrong valve while placing a cation demineralizer in service. The draining of the volume control tank went unnoticed by control room operators for 5 minutes until the "VCT Auto Makeup Required" annunciator alarmed. The volume control tank level trend was clearly indicated on a main control board panelin the Unit 1 control room. By itself, the valve manipulation would not have diverted volume control tank water to the spent resin storage tan However, a normally closed spent resin storage tank inlet isolation valve had been caution tagged in the open position and provided the remainder of the drain pathway. Licensed operators developed Condition Report 97 12966 and an avent review team was assembled to evaluate this event. The inspectors reviewed the evaluation and discussed the event with licensee managers, Observations and Findinos On August 20, Unit 1 chemistry personnel requested that operators place a cation domineralizer bed into service in thc chemical and voluma control system in order to adjust reactor coolant system pH. A reactor plant operator was sent to the mcchanical auxiliary building to plece the cation bed in service by opening the bed inlet and outlet valves and closing the associated bypass valve. The operator opened the inlet valve, then mistakenly opened the outlet isolation valve from the bed to the spent resin storage tank, and finally opened the cation bed bypass valv Ordinarily, the valve alignment would have isolated reactor coolant system letdow However, a norm:.lly closed spent resin storage tank inlet isolation valve, Valve 1 WL 0154, had been caution tagged in the open position. This condition

- . . -

,

..

enabled letdown flow to divert from the volume control tank to the spent resin storage tank. Approximately 5 minutes later, the primary reactor operator received a "VCT Auto Makeup Required" alarm in the control room and directed the operator to secure the cation bed. While securing the cation bed, the operator recognized his mistake and requested assistance in restoring the system to its normal alignment. The reactor operator then restored volume controlleve In their evaluation, the licensee vtermined that the spent resin storage tank isolation valve had been opened ;nder an equipment clearance order caution tag as an operational convenience because the valve was located in a locked high radiation area. Operators had written a procedure feedback form to request that Plant Operating Procedure OPOP02 WS-0002, Revision 0, " Waste Transfer to the Portable Solidification System," Lineup 1, be updated. This action, however, had bypassed the normal procedure revision and review process. The use of a caution tag instead of properly maintaining this station procedure was considered the first example of a violation of Technical Specification 6.8.1 (498;499/97006 01). In addition, the reactor plant operator he.d not met management's expectations for self verification, and the primary reactor operator had not met management's expectations for control board monitorin The licensee's corrective actions included the following:

Verifying that the spent resin storage tank did not overflow into the unit vent heade * Removing the caution tag on the spent resin storage tank isolation valve and closing the valv * Requiring dual verification or peer checks for chemical and volume control system cation bed evolution * Reinforcing control board monitoring expectations for control room personne * Evaluating the use of caution tags for configuration control, c. Cpaclusjons A reactor plant operator did not adequately verify the proper component identification when performing a valve manipulation The reactor operator did not adequately monitor volume control tank level while the evolution was ongoing. The use of a caution tag to implement a change to a proceduralized valve alignment instead of a procedure change was the first example of the violation of Technical Specification 6. _ .. . _ . . _ . ._ . __ . . . . ~ _ __ __ _ _ _ _ __ _ . _ _ __._ __ _ _ _ _ _ _

l

.. *

,

01.4 Inadvertent Drain of the Soent Fuel Pool (Unit 21

'

i Insoection Scooe (71707)

i i On August 31, approximately 0 inches of water levelinadvertently drained from the l

<

Unit 2 spent fuel pool. The draining was stopped prior to reaching the alarm leve ,

,

An event review team was assembled to evaluate the circumstances surrounding i i this event. The inspector reviewed the evaluation and discussed the event with !

licensee managers.- Observations and Findinag i

, On August 31, Waste Holdun Tank Purification Demineralizer Filter 2A and Spent i Fuel Pool Demineralizer Filter 20 had been removed from service to f acilitate the _

change out of Filter 28. The ta:,k had been accomplished using computer generated-test tags for the filter isolation valves. The unit supervisor had cautioned the

'

operators that the evolution could affect spent fuel poolinventory during a prejob briefing. A work risk assessment had been determined to be unnecessary because the task was being performed in accordance with established procedure The reactor plant operator walked down the equipment clearance orders and signed as the sole owner of the test tags. However, the procedure used for restoration, !

i Plant Operating Procedura OPOP02 FC-0001, Revision 15, " Spent Fuel Pool Cooling and Cleanup," did not contain a specific section for placing a filter back in service.

This action was not in accordance with management expectations as described in

! the conduct of operations document. Regulatory Guide 1.33, Revision 2, '

3 February 1978, Appendix A, recomn..,hds that written procedures be developed for the replacement of important filters. The failure to have procedures recommended by Regulatory Guide 1.33 is a second example of a violation of Technical "

'

Specification 0.8.1 (498;499/97006 01). '

I After performing the velve alignments in accordance with Procedure OPOP02 FC-0001, the operators determined that there was no flow in the system. The reactor plant operator referenced the system drawing and started

'

a valve alignment check of the system. He discovered that the filter outlet valve was still closed and tagged with a test tag that he owned under the equipment clearance order. When the valve was opened, flow was established in the system,

' and the operator left the are During the clearance restoration for Filter 28, the radwaste operator disecvered that Filter Vent Valve 2 FC-0026B and Filter Drain Valve 2 FC 00278 had been leh '

' open. This was the position required by the equipment clearance order. He 1 immediately closed the vent and drain valves, lif ted the tags, and informed the control room. The radwaste operator checked the spent fuel poollevel and ;

.

e e-- - --- -e *--,,-e,-r,~, n.w--w- ,- www , r+e ws.m ,s- - ,en- we,,-,w-smm.,,,,-- wn-m.--,m ,w v4 g- m-,y s w m, r v y. . -m. , -

-w-,, --g--A

_ _ _ _ _ . _- . _ _ . _ . . . . _ . _ _ . . _ . .

_ _ _ _.___ _ 3

_ __

i

,

..

-

l i

8 I

i

!

discovered that the level had decreased 3 inches from the previously logged measurement.- He informed the control room operators and verified that most of ;

the water had gone to the waste holdup tank.-

l The event review team identified the root cause of the event as improper use of ,

test tags to provide configuration control for filter replacements. The following

items were identified as contributing causes: i Failure to meet management expectations with respect to the Conduct of !

Operations document in that no procedural guidance existed when required and was not questioned by operations personnel for an extended length of time, o The lack of procedural guidance for fill and vent of Spent Fuel Poni Filter 2 .;

e Weak supervisory oversight and control which included:

' '

- The limited scope of th's prejob briefing

- Covering two jobs with the same briefing i

- Not completing one evolution before starting the second jo The lack of visible ownership of the evolutio The lack of visible supervision for the evolutio ,

o The failure to employ a questioning attitude when things did not go as

expecte .

e Failure to use self checking techniques during the manipulation of valves, d

The inspectors performed a valve alignment check of the system following the ,

event. During the documentation review, the inspector found that licensee ,

management assigned appropriate resources to investigation and to corrective l actions for prevention of recurrence of this event. The plant managers issued a i memorandum to the operations managers instructing them to implement interim guidance immediately and revise appropriate policies and procedures within 60 days from September 4,199 Corrective actions taken or proposed included:

e A comprehensive operations management evaluation of the use of test tags, e including a specific section on placing filters in service following replacement in Proceduro OPOP02 FC-000 ,

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

.

..

,

Reinforcing expectations for field supervisor oversight, peer checking, and prejob briefin * RNewing and implementing specific training needs for this evolutio The inspectors noted that, as documented in NRC Inspection Reg..ts 50 498/97 005; 50 499/97-005, Section 01.3, 50-498/95 023; 50 499/95-023, Section 2,6, 50 498/95 020;50 499/95 20, Section 7.1, a /96 004; 50 499/96 004, Section 2.5, the plant has experierwed four additional loss of spent fuel poolinventory events over the last 2 years. All of the:o events were at least partially attributed to a lack of adequat; procedural guidance. The failure to have written procedural controls governing the replacement of the spent fuel pool filter was in violation of Technical Specification 6,8,1. This nonrepetitive, licensee identified and corrected violation was net ccnsidered for discretion in accordance with Section Vll,B.1 of the NRC Enforcement Poliev because corrective actions for previous events have not been successful in preventing inadvertent losses of spent fuel pool inventor The inspectors reviewed the licensee's corrective actions related to this event as well as the response to previously issued Violation 498;499/97005-03. In that response, the licensee committed to conduct a multidiscipline, comprehensive rev3ew of the equipment clearance order procedure and the plant configuration program, Based on those commitments, no additional response is considered necessary because the proposed corrective actions should correct the problems identified in both examples of this violation, C.gpclusions The inappropriate use of equipment clearance order test tags was considered to be the cause of an inadvertent drain of the spent fuel pool Although specific identified causes were different, corrective actions taken following four previous inadvertent losses of spent fuel poolinventory proved ineffective in preventing such occurrences. The f ailure to have written procedural controls governing the filter replacement conducted on August 31 was considered a second example of th'

violation of Technical Specification 0.8.1, 01.5 Two Examnles of Work Performed in the Wrona Unit (Unit 1)

On August 18, instrumentation and controls technicians incorrectly removed the Unit 1 open loop cooling header pressure transmitter from service. The technicians were attempting to perform a calibration of the corresponding transmittar on Unit Condition Report 97 13387 was written to document the error. The cause of the event was determined to be inadequate self verification. The technicians were removed from power block work for the remainder of the day and were caunselled regarding the event, Briefings were conducted with maintenance crews end site

- .. - - - .- - . _ .-. . - - .. - . -..- ---.. . - - .

_

!

,

.. .

. ;

10- l i

wide during human performance enhancement day, A better method of labeling components at the circulating water intake structure was also being evaluate ;

On September 17, reactor plant operators incorrectly opened circulating water !

- system Siphon Breakers 2 CW 022" id 2 CW 0222 instead of the corresponding ;

'

va!ves on Unit 1. Licensed operato.. m the Unit 2 control room observed increased l

. system pressures and contacted reactor plant operators via radio to verify pump l

} discharge prassures. The Unit 1 operators overheard the radio transmission and j then identified and corrected the cause. Condition Report 9714561 was written to

'

.

review tbs event. Licensee management determined that the cause was a lack of j self verification and poor supervisor / briefing on this unique evolution. The causes i worn ap~orriately addressed by the corrective actions take !

The inspectort determined that both events were unique because they occurred in a ;

notomort eucture. The licensee did not have a recent history of wrong unit events 6nd Ne W.anificance of the events was considered minor because no impact to safe l

,

opeut!ons was presen [

'

01.0 ' Inadvertent Mode Chanae (Unit 1)

I insocction Scoce (71707)

.

During a routine review of the control room log, the inspectors noted an entry

'

indicating that reactor coolant system temperaturet may have exceeded Mode 3 entry e<,nditions after licensed operators had entered Mode 4. The circumstances surrour, ding this event were reviewed. Interviews were conducted with the licensed operators, the shift supervisor, and the operations manager. In addition, the following documents were reviewed:

i e Plant Operating Procedure OPOP03 ZG 0007, Revision 15, " Plant Cooldown" e Flant Operating Procedure OPOP03 ZG 0001, Revision 15, " Plant Heatup" f

Condition Report 9714188, "During the Process of Establishing Residual Heat Removal Pumps in Operation, Transient Reactor Coolant System j Temperature Exceeded 350 Degrees"

Observations and Findinas At 5:30 a.m. on September 13, licensed operators cooled the Unit 1 reactor to less

.

- than 350*F, entering Mode 4. In accordance with Procedure OPOPOLZG 0007, i licensed operators disabled one centrifugal charging pump and two high head safety injection pumps by tacking out the associated power supply breaker and tagging the

- breaker with an equipment clearance order. The reactor coolant system makeup -i flow une was then throttled to the Mode 4 required maximum valu ,

t e*s-e,- ,=v--e e r 2, -w , -,-- y.--- , ,. -+ .c,.a,-, -m -m., e,~ w w .g,-- ne- + - ,-m -- n

.- _ __ .. - = _

,

.. .

.

11- ;

,

At approximately 6:15 a.m., the residual heat removal system was placed in service and cooling via the steam generator power operated relief valves was terminate Reactor cociant system temperatures began to rise. At 6:17 a.m., reactor coolant system temperature increased above 350'F, constituting an inadvertent entry into Mode 3. Operators responded to cool the reactor back within the Mode 4 definition of less than 350'F. By 6:27 a.m., the reactor coolant system temperature was again less than 350*F h*ving reached a peak of approximately 353' The inspectors reviewed the Technical Specifications related to changing modes from Mode 4 to Mode 3. The affected specifications permitted licensed operators 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and/or 25'F to properly align systems for Mode 3. Therefore, Technical Specifications had not been exceeded. However, Provedure OPOP03 ZG 0001, Steps 6.20 through 6.28 required that numerous administrative reviews be completed prior to changing mode The inspectors determined that the inadvertent mode change at 6:17 a.m. on September 13 was in noncompliance with the licensee's adtninistrative controls and was, therefore, in violation of Technical Specification 6.8.1. This nonrepetitive, licensee identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Pollev (498/97006-02).

Condition Report 9714188 was written to document the event. Licensed operators found that the residual heat removal system heat exchanger flows from the system and/or from the component cooling water system had been insufficient to prevent the reactor coolant system temperature from increasing, onco the power operated relief valves were shut. An overly restrictive temperature band of 340 349'F and less than adequate communications among the control room operators were viewed as contributing cause The need to properly control reactor coolant average temperature and to adequately communicate plant status was discussed with the reactor operator. A procedure revision was initiated to ensure that the transfer of reactor coolant system ternperature control from the power operated relief valves to the residual heat removal system would be performed at a sufficiently low temperature to allow margin to the mode change in addition, the event will be reviewed by the licensed operator training curriculum review group as a potential requalification training subject.

c. Conclusiong An inadvertent mode change from Mode 4 to Mode 3 was caused by insufficient residual heat removal system flows during system initiation. The modo change was in violation of plant operating procedures. This nonrepetitive, licensee identified .

and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcernent Polie !

-

.

,

!

12 l

I 01.7 H_eview of Inocerable Automatic Functions (Unit il {

!

The inspectors reviewed the Unit 1 totalimpact assessment log for the week l ending September 29,1997, prior to the return of the unit to service. The purpose ;

of the log was to ensure that the impact of inoperable automatic functions and !

'

main control board instrumentation out of service was not excessive. The log appeared to be complete and accurate. All goals related to licensed operators were fully met and the remaining impact was considered minimal. Reactor plant-operators were impacted by some inoperable automatic functions. The assessment 1 of these functions was adequate, ar.d watch station impacts met management's :

goals, with 'the exception of the mechanical auxiliary building reactor plant operato ,

I The inspectors reviewed the impacts to the mechanical auxiliary building watch station. An estimated 75 minutes of the 12-hour shift was required to compensato i for the inoperable functions. This exceeded the goal of less than 60 minute i Station management routinely reviewed these goals during the daily communication ,

and team work meeting. _ Corrective actions had been initiated to correct the !

deficiencies, and compensating actions were In place to lessen the impact by ;

sharing the work with other watch stander *

O2 Operational Status of Facilities ana Equipment ,

,

02.1 Plant Tours (Units 1 and 21 Insoection Scone (71707)

The inspectors routinely toured the accessible portions of plant areas in Units 1 and 2. Areas of special attention during this inspection period included:

e Units 1 and 2 turbine g;er.creter buildings i e Units 1 and 2 mechanica auxiliary buildings l e Units 1 and 2 fuel handling buildings e Units 1 and 2 electrical auxiliary buildings I o Unit i reactor containment building ]

e Unit 2 Standby Diesel Generator 21 Observations and Findinog The inspectors observed that systems and components had beea maintained in good material condition in both units, in general, housekeeping was very goo The inspectors toured the Unit 1 reactor containment building throughout Refueling I and Equipment Outage 1REO7 Overall material condition of equipment in the reactor containment building was cood. A review of containment inspection - l activities was performed and is documented in Sections M8.4 and M8, I

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

,

.. ;

r

!

13- !

'

i

!~

The inspectors routinely toured the one stop shop and accessible plant areas throughout Outage 1REO7. Through observation, review d documentation, and l interviews with plant workers and outage management, the inspectors determined i that proper controls were in place to ensure a safe return to service of the uni I Schedule setbacks were aggressively addressed while ensuring that workers :

performed cautiously. Management oversight of the activities was evident and !

appropriate. The sbKt supervisor maintained cognizance of system status and ;

availabilit f Conclusions

,

Material condition and housekeeping in the areas toured was good. Refueling and Equipment Outage 1REO7 was well run and controlled with a proper perspective on nuclear safety and shutdown ris :

03 Operations Procedures and Documentation ;

03.1 Doerator Knowledae Related to the Status of Normalized Instrument . Insoection Scoce (71707)

t

. The inspectors reviewed the normalization process utilized during the Unit 1 l refueling outage. This involved a detailed review of the working copy of Plant i Maintenance Procedure OPMP08 SP 0001, Revision 5, "RPS/ESF System i Normalization." The inspector also discussed the process with licensed control

"- toom operators and instrumentation and controls technician Observations and Findinas Tha inspector observed the use of the normalization procedures and checksheets and found that they were correctly implemented. The licensed operators were asked if they had received specific training on the normalization procedure. They ,

stated that no specific training was rt;alved. However, the inspector found that all operators interviewed were f amiliar with the process and knowledgeable of which

,

instruments could be normaiired. The operators knew which instruments were active and which were normalized, Conclusions Licensed operators were appropriately aware of the status of main control board instrumentation with respect to normalization. Procedures governing instrument normalization were being implemented by the control room operator $

_ - w., . . . - .- _ ,c,v. . .,, .. .., ., ,,5.m.,.9,,, ..,m__w.,,w_.- ..,w. y _m.v.,..v.,,.wm .

..q..-,, v-

'

.

14-08 Miscellaneous Operations issues (92901,92700)

08.1 IClored) Viola: ion 50 498/96003-02: Failure to Properly Evaluate the Effects of an Out of Cahbration Conditio This violation cited that a differential pressure indicator was found out of calibration and that the condition was not evaluated and docurnented in accordance with the quality assurance plan. This resulted in continued acceptance of inaccurate baseline data fvr Residual Heat Removal System Pump 1C. The licensee found that there was no clear procedural guidance to direct the system engineers to evaluate out-of to!cranct 'itions on process instrumentation used as measuring and test equipment. The , wing corrective actions have been implemented:

An evaluation was performed on the out of tolerance condition. Other out-of tolerance conditions were reviewed and no additionalinadequate evaluations were identifie * Training was provided to the system engineers who were responsible for evaluating out of tolerance condition * Plant General Procedure OPGP03 ZM-0016, Revision 11, " Installed Plant Instrumentation Calibration Verification Program," was revised to d fine responsibilities and provide specific guidance for out-of tolerance evaluation The inspector concluded that the corrective actions addressed the concerns stated in the violation.

08.2 (Closed) 1.icensee Event Report 50-498/96-001: Failure to Verify the Operability of Offsite Power Sources When a Standby Diesel Generator was Inoperabl During planned maintenance on a Class 1E 125 vdc bus, the battery output breaker was opened and racked out. Technical Specification 3.8.2.1 was entered requiring restoration of the battery bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The technicians and licensed operators failed to recognize that, because this battery bus supplied control power and field flash for the associated standby diesel generator, entry into a more restrictive Technical Specification with a 1-hour action statement was require The battery bus w6s unavailable for 67 minutes. This nonte petitive, licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Poliev (498/97006 03,.

The root cause of this event was the failure to properly in.plement planning and preparation requirements. The preventive maintenance task instruction failed to reference the affect on the associated standby diesel generator. In addition, the work risk assessment process did not cause the review of contingencies required upon changes in plant system alignment . . - . _ - . - _ . - - - _ - - . - - . - - _ . . - - _ - ---

,

... . l

!

15- ,

!

'

The corrective actions that were taken to strengthen the barriers included: ;

i e Additional training for technicians and licensed operators, i e Evaluation of the adequacy of the work risk assessment for e Revision of the plant annunciator rest 9se procedures to include a reference ;

to the standby diesel generator operability Technical Specificatio Licensee managt went performed a self assessment of the corrective action effectiveness associated with this licensee event report. The findings of the ;

assessment were that the actions for revising the preventive maintenance tasks, for ;

revising the word process procedure, and for training the electrical maintenance planners and craft had not been fully effectiv Condition Report 97 8384 was initiated based on the above findings to ensure that '

barriers to prevent recurrence would be made effective. Additional corrective

actions identified included: ,

'

e Management expectations related to the work risk assessment process were discussed with the first line supervisors, o A formal training session for first line supervisors was develope e Work risk assessment forms were revised to remain with the work documentation as a permanent part of the documentatio o Condition Report 9712434 was developed to continue evaluation of the '

work risk assessment process following Unit 1 Refueling Outage 1REO i

'

Based on the additional actions taken and proposed, the inspector considered the actions were reasonable. The inspector also concluded that the licensee's self-assessment was critical and noteworth .3 (Closed) Insoection FoCowuo item 498:499/93031-36: Senior shift managers will

provide continuous management representation and presence during selected evolutions throughout the power ascension progra This item was related directly to the Unit 1 Operational Readiness Plan, Action Summary item 7. The purpose of the commitment was to ensure that the exercise 4 of command and control authority of licensed operators was not diluted by the E increased level of activitie Prior to the Unit 1 restart, inspectors reviewed issues related to operator staffing, 4

. . capability of licensed operators, and ability cf the operators to maintain proper focus. As documented in NRC inspection Report 50-498/93 041; 50 499/93 041,

,

l

.

..

.

16-observed control room command and control was very good, and management had been effective in reducing operator administrative burden. In addition, licensed oprator command and control was inspected and addressed as documented in NRL Inspection Reports 50 498/94-009; 50-499/94 009, and 50 498/94-010; 50-409/94 010. This item was addressed for Unit 2 as documented in NRC Insper tion Report 50 498/94 20; 50 499/94 20, Section 2.4.4. In addition, licensed operator command and control was inspected and addressed as documer.ted in NRC Inspection Reports 50-498/94 017; 50 499/94-017, and 50-498/94 024; 50 499/94 02 This item addressed an issue that was selated alely to the 1994 restarts of both units. In addition, the subject was reviewed and documented by NRC personnel on numerous occasions. Therefore, this item is administratively closed.

08.4 1 Closed) Unresolved item 498:499/97005 0.2: Manual Valves in Certain Containment Penetrations not Surveilled in Accordance with Technical Specification 4.6.1. This item was written to address concerns related to the surveillance requirements associated with manual valves in the containment isolation scheme of 10 containment penetrations. The inspectors had determined that the associated valves werc locked in the closed position. However, the licensee had not been verifying the valve position once per 31 days as required by Technical Specification 4.6.1.1.a. The subject valves were not capable of automatic closure and were required to be closed during accident conditions. This item had remained open to review the applicability of Technical Specifici. tion 4.6.1.1.a to the subject valves, in addition, the item trac" ' two penetrations that were not listed in Updated Final Safety Analysis Report (UFSAR), Figure 6.2.4 Licensee engineers stated that the va!ves were not subject to the requirements of Technical Specification 4.6.1.1.a because they were associated with penetrations that were not required to be closed during accident conditions and because the letter of the specification related directly to penetrations and not specific lines or valves. The inspector reviewed an internal NRC interpretation of the subject specification. The Office of Nuclear Reactor Regulation had documeated that valves, associated with penetration line isolation, that were not capable of being closed by automatic isolation valves were required to be verified closed and secured in accordance with Technical Specification 4.6.1.1.a, irrespective of the isolation capability of other lines associated with a common penetration, in addition, the inspector reviewed NRC Inspection Report 50 456/97-009; 50 457/97 009. In this inspection report, that licensee was cited under the

" General Statement of Policy and Procedure for NRC Enforcement Actions, NUREG 1600, for similar violations of Standard Technical Specification 4.6.1. Therefore, the inspectors determined that the f ailure to verify that valves associated with the 10 penetrations listed in NRC Inspection Report 50 498/97-005:

.

-

..

,

50 499/97 005, Section 02.2, was in violation of Technical Specification 4.0.1. (498;499/97006 04).

The inspectors determined that on September 11,1997, operators verified that the subject valves were secured in the closed position. This verification, and the need for final resolution of this issue were documented in Cc,ndition Report 9713071 for tracking and possible corrective actions. In additi , Condition Report 97 12022 was written to document that Penetrations M 71 and M 87 were not listed in UFSAR, Figure 0.2.41, and to track related corrective actions.

08.5 (Closed) Violation 498/96004-04: Midloop Level Transmitters Not Placed in Service This violation cited that two narrow range level transmitters had not been placed in service as required prior to entering midloop operation. The licensee identified inadequate procedural guidance as the cause of this event. The stated corrective action was to revise Plant Operating Procedure OPOP03 ZG 0000 to direct the instrumentation and controls technicians to place the narrow range hot leg level transmitters in service. Based on the reviews documented in Section 01.2 of this report, the inspectors concluded that the licensee's corrective actions had been implemented and addressed the concerns in the violation.

08.0 (Closed) ViolDtion 499/97001 01: Reactor Coolant System Level Sight Glass not Properly Aligned during Midloop Operations This violation cited that the reactor coolant system level sight glass was not properly aligned as required during midloop operation. A vent valvo in the sight glass upper connection was not accounted for by procedures and was left open, causing a loss of this indication uring the reactor coolant system vacuum fill evolution. The licensee identified inadequate procedural guidance as the cause of this even The licensee identified the following corrective actions:

Revise the reactor vessel vacuum fill procedur * Evaluate corrective actions for the vent valves on the temporary manifold in the upper sight glass connectio * Review procedures used for evolutions involving heatup, cooldown, and reduced inventory to ensure that configuration control issues do not arise during transitions between these procedure Based on the reviews documented in Section 01.2 of this report, the inspectors concluded that the licenseo corrective actions had been implemented and addressed the concerns in the violatio . _ _ _ _ . _ __ . __ ._ _ _ _ _ _ _

i

.. <

,

l l

18-l

.

'

II. Maintenance

'

M1 Conduct of Maintenance M1.1 Genert.1 Comments on Field Maintenance Activiting Inspection Scone (62707)

f The inspectors observed portions of the following work activities identified by their work activity numbers:

.

9501279 Reactor Coolant Pump 1D Motor Replacement e 102144 Replacement of Station Vital Battery Bank 10

115463 Mechanical 1.atch Repair on Refueling Transfer Cart Observations and Findinos

"

The inspectors found the work performed during the observed activities to be thorough and conducted in a professional manner. The work was performed by knowledgeable, qualified technicians utilizing approved procedures. Supervisors were observed providing an appropriate level of oversight. System engineers were

. observed providing quality technical support as needed. Prejob briefings were thorough and radiological controls were in place where applicable. Design change packages and evali'ations were reviewed with no observed discrepancies Vendor recommendatn.nr were incorporated into detailed work instructions. In addition,

-

during the replacement of the vital battery bank, craftsmen had developed techniques designed to improve consistency in installation and to reduce the unaval' ability of the battery, Conclusions Observed maint'snance activities were well coordinated and implemented by knowledgeable technicians with approoriate levels of supervision and good support from engineering and operations personne M1.2 General Comments on Surveillance Testina insoection Scone (61726)

The inspectors observed portions of the following curveillance activities:

Unit 1:

-

,

.. -

-19-

Plant Surveillance Procedure OPSP02-CM 4102, Revision 2, " Containment Hydrogen Analyzer Analog Channel Operability Test (A-4102)"

Plant Surveillance Procedure OPSPO6-DJ-0004, Revision 3, "125 Volt 1E Battery Service Surveillance Test"

Plant Surveillance Procedure OPSP10 DM-0003, Revisim 5, " Automatic Multiple Rod Drop Time Measurement'

Unit 2:

Plant Surveillance Procedure OPSP06 DJ-0006, Revision 5, " Battery Charger 8 Hour Load Verification"

Plant Surveillance Procedure OPSP11-MS-0001, Revision 9, " Main Steam Safety Valve Inservice Test" b. Observations and Findinns The inspectors found that the observed surveillance activities were performed in accordance with approved procedures. Reviews of procedures indicated that Technical Specification surveillance requirements were properly implemente I Limiting conditions for operation were properly adhered to throughout the testing evolution and tracked in the operability assessment system. The system engineers wore present during testing and properly evaluated test result During the performance of Procedure OPSP06-DJ-0004, the inspectors n:ted that ,

an equipment clearance order test tag had been utilized to remove the battery from l service. The inspector determined that the breaker tagout did not meet the procedural definition for usage of a test tag. In additon, the clearance was required to protect property and personnel, indicating that a danger tag should have been ;

used. The inspectors determined that the inappropriate use of a test tag was minor i because personnel safety was assured by the tag and by personnel actions take l In addition, as documented in Section 01.4 of this inspection report, the licensee is '

performing a comprehensive review of the equipment clearance order program to identify corrective actions to correct this type of proble On September 5, during the testing of main steam safety valves, two of the Unit 2 valves failed to open within the Technical Specification required tolerances. The licensee had previously identified problems with these valves. The most probable cause was attributed to oxide locking between the nozzles and disks. The licensee reported this condition in Licensee Event Report 50-498/97-009. This issue was i previously addressed as documented in NRC Inspection Report 50-498/97-003; I 50-499/97-003. The NRC continues to review the oxide-locking phenomenon, and I these specific failures will be further reviewed prior to closure of this licensee event I report. Although the safety valves lifted above the setpoint allowed by Technical l l

.. . ~ . -. - - - . . - - - . . - - - .

-.4 y

'

.,- .

, ;

, -- >

.

20-

. . . Il Specificatio'ns, the valve setpoints were reset within tolerance, in additionilicensee ,

Jengineers determined that the as found cond tion of the valves would have provided .

full design basis relief capabilit I

'

On September 13, during testing of rod cluster control acsemblies in accordance with Procedure OPSP10-DM-0003, six control rods failed to fully insert. This phenomenon was first Identified following a Unit 1 reactor trip on December 18, 1995. -The as found control rod configuration was determined to support operability and was well within previously calculated design basis bounding ;

conditions. The core reload conducted during Refueling and Equipment Outage

1REO7 was designed to correct allincomplete rod insertion conditions. This issue continues to be tracked and assessed by the Office of Nuclear Reactor Regulation

, under a review of the responses to Generic Letter 96-01, " Control Rod Insertion Problems." Conclusions The surveillance activities observed were performed in accordance with the

, applicable Technical Specifications. An additional example of inappropriate use of an equipment clearance order test tag was identified. However, this example was considered minor and corrective actions were being taken. Testing of the main steam safety valves and rod control cluster assemblies identified continued

-

'

examples of generic problems with these components. However, corrective actions taken during the Unit 1 outage were expected to conect both conditions.

,

M2 Maintenance and Material Condition of Facilities and Equipment M 2.1 Comoonent Coolino Water Heat Exchanoer Performance Testino (62707)

The inspectors observed performance testing of Component Cooling Water Heat Exchanger 1C in accordance with Plant Engineering Procedure OPEPO7 EW-0001, Revision 5, " Performance Test For Essential Cooling Water Heat Exchangers." The purpose of this test was to ensure that biofouling was eot present in the essential cooling water side of the heat exchangers sufficient to decrease the performance of the thermal exchang The inspectors verified that the prerequisites and precautions in the procedure had been met. _ Testing resistance temperature detectors were properly calibrated and were rated at the required accuracy for the test. Following system stabilization l the

'

inspectors independently collected data and calculated the fouling f actor for the heat exchange ' The system engineer conducting the test was knowledgeable of the system, system characteristics, and intricacies of the test. Data was electronically collected and reduced. The engineer then performed the associated calculations. The heat j exchanger fouling factor was well within the test acceptance criteri l l

.. . _ . __ __ _

.-.

.

.. -

21-M8 Miscellaneous Maintenance issues (92902; 92700)

M8.1 (Closed) Licensee Event Reoort 50-499/97-006: Manual Reactor Trip Initiated upon Malfunction of a Main Feedwater Rcgulating Valv This event was previously reviewed as documented in NRC inspection Report 50-498/97 03; 50-499/97-03. The previous review had included direct observations of operator rcsponse and the licensee's posttrip review. The response of the operations staff r.id the associated plant equipment had been considered excellen The inspectors d9termined that no new issues were revealed by this report. The corrective actions included the replacement of the f ailed circuit cards, alignment checks of ti'a circuit-driver cards in the other three main feedwater regulating valves, arJ a more generic evaluation of the use of this type of circuit-driver cards in other systems that could result in a reactor trip. In addition, the licensee comrotted to evaluate potential design changes to reduce the susceptibility of the me;n feedwater system to single point failures.

M8.2 (Closed) Licensee Event Reoort 50-498/96-005: Reactor Containment Building Personnel Airlock incorrectly Declared Operabl An inflatable seat on the reactor building side door of the personnel airlock had been replaced, the door was declared operable, and the mechanical auxiliary building side door was unlocked. A localleak rate test was then performed on the reactor building side door. This test failed, indicating that the door had been inoperabl Having one door inoperable and the other unlocked was contrary to Technical Specification 3.6.1.3 requirements. This nonrepetitive, licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (498/97006-05).

The violation existed for about 28 minutes on December 13,1996. Because of the short duration, the inspector concluded that the safety consequences were lo The licensed operators took immediate steps to return the containment integrity to compliance with the Technical Specifications upon discovery. The root cause of the ovent was determined to be poor communications between the craft personnel and operators. Corrective actions, including reinforcement of management's expectations for precise communications and the requirements for declaring components operable, were completed and were considered reasonable by the inspector.

M8.3 (Closed) Licensee Event Reports 50-499/96-003: Improperly Installed Jumper on a Main Steam Line Pressure Lead / Lag Circuit Car While performing an analog channel test on a Steam Generator 2D pressure instrument, the technicians had found Jumper TP3 for Card P03-0627 connected to

-

..

,

22-the " FIXED" pin instead of the " VAR" pin. With the jumper in that position, the lead / lag function had been removed from the associated channel contrary to the requirements of Technical Specification 3.3.2. This nonrepetitive, licensee-identified and corrected violation is being treated as a noncited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy (499/97006-06).

The cause of the event was determined to be inattention to detail by the technicians who had performed the previous test of that channel. A contributing cause was that the surveillance procedure had not provided for adequate posttest verification of the jumper position. Corrective actions included:

Enhancement of the test procedures that required verification of jumper positio * Review and modification of similar test procedures, as necessary, to preclude this type of even The inspector reviewed the safety impact evaluation related to having en inoperable lead / lag function for that channel. The safety consequences were considered to be very low. All corrective actions were completed and appeared to have been reasonable.

M8.4 (Closed) Violation 499/97002-01: Inadequate containment inspection regarding debris not removed that could affect containment sump operability This violation involved the licensee's failure to verify the operability of the emergency core cooling system sump when a visualinspection of areas affected by a containment entry was conducted in accordance with Plant Surveillance Procedure OPSP03 XC 0002, Revision 11, " Initial Containment inspection to Establish Integrity." The inspection failed to verify that no loose debris was present in containment. Specifically, plastic bags containing protective clothing and other loose debris that could have been transported to the containment sump and could have caused restriction of pump suctions remained in containment after the visual inspectio The licensee identified the following root causes for this event:

A f ailure to communicate management's expectations for the control of loose debris in the reactor containment building during restart operations following an outag * A lack of clear distinction in the surveillance procedure between the acceptance criteria for the Technical Specifications requirements and the station's housekeeping expectation .

,

-23 The licensee's corrective actions included:

Inspection for and removal of allidentified loose debris from the Units 1 and 2 reactor containment building * Revision of the surveillance procedure to require a formal briefing by the shif t supervisor or his designee of all containment entry team members on the acceptance criteria required for the control of loose debris in the reactor containment building in Modes 1,2,3, and * Assignment of a team leader for each containment entry team responsible for ensuring compliance wPh the requirements of Technical Specification The inspectors reviewed the licensee's implementation of their corrective actions, including Revision 16 to Procedure OPSP03 XC-0002. The inspectors also observed portions of the licensee's containment closecut inspections at the end of Refueling and Equipment Outage 1REO7. A three-step process was implemented to ensure satisfactory completion of the surveillance inspections. The first step involved maintenance crews performing inspections of their work areas. This was followed by area walkdowns performed by maintenance managers. Finally, the surveillance inspections were completed by licensed operators. The inspectors considered the licensee's process for completing the surveillance inspections goo The inspectors observed that there were several ongoing activiti' during the closecut inspections. The volume of supplies necessary for thes, tasks represented a potential for conflict with the containment inspection process, However, the observed results of the inspections were good. Operators performed well and conducted thorough inspections in the areas observe Based on these reviews, the inspectors concluded that the licensee corrective actions had been implemented and addressed the concerns in the violation.

M8.5 (Closed) Licensee Event Renort 50-499/97-003: Failure to meet the requirements of Technical Specification 4.5.2.c requiring an inspection of containment for loose debri This licensee event report documented the same event discussed and closed in Section M8.4 of this inspection report. Therefore, this item is administratively close _

.

_ _ _ _ _ _ _ _ _ _

,

.. -

24-lil. Enoineeririg E1 Conduct of Engineering E Imn!ementation of Outaae Related Commitments jnsoection Scone 07551)

in preparation for, and during the Unit 1 refueling outage, the inspectors reviewed outage-related commitments made by the licensee in the engineering functional area. Reviews of originallicensing basis documents and implementing procedures were conducted The inspectors verified licensee compliance with the following commitments:

e Ensure that controls were in place to operate the spent fuel pool within design temperature limits,

Continue to assess the essential cooling water system as delineated in the response to Generic Letter 89-13. Specifically,

-

Performance test or inspect heat exchangers

.

Inspect system for aluminum / bronze dealloying flaws

-

Inspect intake bays for biofoulin * Make committed progress in removing Thramo-Lag installations in containment, e if any residual heat removal pumps are disassembled, inspect the impellers for crackin * Perform a water balance of the assential enoling pond (ultimate heat sink) on a 5-year rotating basis, e Review the resctor coolant system volume determination with respect to reductions resulting from steam generator tube plugging, Observations and Findinos in general, the commitments reviewed were being implemented through approved station procedures. The spent fuel pool heatup analysis was reviewed as documented in Section E1.2 of this inspection report, inspections of the essential cooling water intake bays for biofouling and system piping for dealloying had been

'

conducted and documented. Modifications to remove Thermo-lag wrappings inside containment were on track to meet the December 1998 commitment date for having all Thermo-lag removed from containment. No residuai heat removal system l

l

____- __- ____ _-

-

..

,

-25-pumps were disassembled during this outage. However, as documented in NRC Inspection Report 50-498/97 05; 50-499/97-05, impeller inspections were being conducted as committed.

Changes in the reactor coolant system volume resulting from steam generator tube plugging were properly recounted for. The as-left Unit 1 volume was calculated to be 13,739 cubic feet, well within the Technical Specification 5.4.2 Limit of 13,8' 4

.100 cubic feet.

The inspectors reviewed the licensee's response to NRC Ouestion 241.05n related to the UFSAR. In their response, the licensee committed to conduct an ultimate heat sink seepage analysis every 5 years. Calculation CC 5112 documented the results of a seepage analysis conducted from September 31 through November 13, 1995. The test assumptions were conservative, and the results indicated that the essential cooling pond was cepable of providing design basis cooling throughout a 30-day acciden The inspectors reviewed the implementation of performance tests and inspections of essential cooling water heat exchangers. The licensee's response to Generic Letter 89-13, dated January 29,1990, documented the commitment to conduct performance tests on essential cooling water heat exchangers in accordance with Plant Engineering Procedure OPEP07-EW-0001, " Performance Test for ECW Heat Exchangers." The inspectors observed the testing of the component cooling water heat exchangers as documented in Section M2.1 of this inspection report. The licensee had committed to perform these tests once each fuel cycle for three cycles, in addition, preventive maintenance inspection and cleaning on small heat exchangers was considered as an acceptable alternate to performance testin The inspection of the remaining heat exchangers cooled by essential cooling water had been removed from Procedure OPEP07-EW-0001 and placed into heat exchanger specific preventive maintenance tasks. The component cooling water supplemental coolers and essential chiller condensers had all been inspected within the last 18 months. However, preventive maintenance tasks, documented in Plant Surveillarce Procedure OPSPO4-DG-0002, Revision 4, " Standby Diesel Generator 5 Year inspection," written to implement inspection of the standby diesel generator heat exchangers, did not specifically require internal inspection of the heat exchangers. As a result, the heat exchangers had not been inspected since 199 Licensee engineers developed Condition Report 97-14662 to document this problem. This document stated that the inspection nf standby diesel generator jacket water and lube oil heat exchangers had been inadvertently waived. The further review of this issue, including licensee compliance with commitments and applicable regulations will be tracked as an Unresolved item (498:499/97006-07).

-- _ - _ _ _ _ _ - _ _ _ - _ _ _ _ _ - . _ _ _ _ _ - - _ - _ _ - _ _ _ _ _ _ _ - _ _ _

- . . - - - - - . - - . ~ _ _ - . - . - . . - -

-

,

<.

26~ ,

i

.

' Conclusions -

i In general, refueling outage related commitments in the engineering functional area had been properly implemented and procedurally controlled. However, the licensee had not completed internalinspections of standby diesel generator heat exchangers since 1993. This was contrary to commitments made in response to Generic

,

Letter 891 E1.2 Soent Fuel Pool Heatuo Analvsis

'Insoection Scoce (375511 The inspectors reviewed the licensee's analysis of the postulated spent fuel pool heatup for Unit 1 Refueling and Equipment Outage 1REO7 as documented in Condition Report Eng;neering Evaluation 97-11202-2. The following ' additional documents were reviewed:

c o Condition Report 9711262, "SFP Heatup Analysia for 1REO7"

Calculation 97 FC-004, " Spent Fuel Pool Heatup For 1R507%

.

UFSAR Section 9.1, Revision 5, " Auxiliary Systems - Fuel Storage and

Handling"

. e Plant Operating Procedure OPOPO4-FC-0001, Revision 9, " Loss of Spent Fuel Pool Cooling Level or Cooling"

'

a Plant Operating Procedure OFOP08-FH 0009, Revision 14, " Core Refueling" Observations and Findinos

. As documented in NRC Inspection Report 50-498/96-003; 50-499/96-003, the

,

inspectors had determined that licensee engineers had not routinely ensured that peak spent fuel pool temperatures during a core offload would remain below the UFSAR design limit. In addition, calcu!ations had not predicted a time to boil in the spent fuel pool following a postulated loss of coolin The inspectors reviewcd the documents associated with spent fuel poolloading and cooling capability. Calculation 97 FC-004, concluded that, with core offload commencing 120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months after shut down, the licensing basis spent fuel pool normal

-

maximum temperature limits would be met provided the following major

--assumptions were maintained:

'

'* Core offload duration was gr,,ater than 45 hour5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months * Spent fuel pool heat exchanger flow was maintained greater than 3000 gp .

. _ _ . - _

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

. - . .. - .. -.. .-- - . . -. . ~. . - -..--

>

,

' .,

,

a

?

27- * Component cooling water flow to the heat exchanger was maintained ,

greater than 3600 gpm.--

Component cooling water temperature remai_ned less than or equal to 98' r The inspectors ensured that the core offload did not begin within the first 120

hours af ter shut down and verified that the major assumptions were met for

. Refueling and Equipment Outage 1REO7. Additionally, these items were covered

,'

by administrative controls and properly in place, as designated in -

Procedure OPOP08 FH 000 Calculation 97 FC 004 also predicted the best estimate time to spent fuel pool-boiling following a postulated loss of cooling. This was calculated for both conditions following a full core offload and for conditions following the reload. The assumptlons utilized appeared to be reasonable, and the maximum heatup rate

_

predicted provided for an adequate operator response to restore cooling or take appropriate corrective / compensatory actions. The heatup rate was also well within the safety evaluation report licensing limit Conclusions The analyses and practices related to spent fuel pool cooling and cleanup systems, prepared for Refueling and Equipment Outage 1REO7, properly verified that pool loading would remain within the licensing basis assumption E2 Engineering Support of Facilities and Equipment

.,

E Bindino of Rod Control Cluster Assembly (Control Rod) durina Raold Refuelina insoection Scooe (37531,71707)

On September '4,' Control Rod H-2 could not be withdrawn from the fully inserted position wher , ven a withdraw demand. The inspectors reviewed the licensee's

,

response to this occurrence. The following documents were reviewed:

Plant Maintenance Procedure OPMP07 DM-0003, Revision 4, " Rapid Refueling Rod Holdout Operation"

Condition Report 97-14212, "While Performing OPMP07-DM-0003, Control Rod H-2 would not withdraw"

Condition Peport Engineering Evaluation 9714212 3, " Evaluate Failure of the Rod Control System to withdraw Control Rod H-2"

'* UFSAR. Change Notice 2022, " Add insert to Section 9.1.4.2.2.1 Concerning Relieving CRDM Binding"

'f

-

,

.. -

28-

Plant Maintenance Procedure OPMPO4 RX-0019, Revision 17, * Rapid Refueling Mechanical Support"

License Compliance Review for Revision 17 of Procedure OPMPO4-RX 0019

Technical Specification 4.9.1.1.a, " Boron Concentration During Refueling Operations"

Technical Specification 3.9.10, " Water Level in Refueling Cavity"

UFSAR Section 9.1.4.2.2.1, " Refueling Procedure, Phase I - Preparation" In addition, the inspectors observed licensed operator activities to resolve the binding and the withdrawal of Control Rod H 2.

b. Observations and Findinas During the performance of Procedure OPMP07-DM 0003, licensed operators withdrew Control Bank C. However, Control Rod H-2 would not withdraw. The withdrawal was being conducted as part of a control rod lockout operation designed to allow the integrated head package to be mo/t d from the reactor vessel flange to the rapid refueling head stand with the control rods withdrawn into the integrated head packag Licensee engineers, with the vendor's assistance, determined that the control rod drive mechanism movable gripper latch was bound against the drive rod. Control rod drive mechanism coil current profile recorder traces were obtained from the stationary, movable, and lif t coils for Control Rods H 2 and B-8. The differences between the signatures of these rod drive mechanisms were compared for both withdraw and insert cycles. As a result of this review, engineering personnel applied mechanical agitation to the movable gripper latch mechanism in an attempt to free the latch. These attempts were not successful. The cause of the binding was determined to be the repeated cycling of the insert function of Control Rod Drive Mechanism H 2 during attempts to fully insert Control Rod F-10. Following the reactor shutdown on September 13, licensed operators had performed rod drop testing in accordance with Technical Specification, as documented in Section M of this inspection repor Engineering personnel determined that detensioning of the reactor head would eliminate the binding. The head was then lifted approximately 2 inches to relieve the frictional forces and allow the movable grippers to release. The head was set back down, and Control Rod H 2 was exercised, withdrawn, and locked into the rapid refueling positio ,

.. -

The inspectors reviewed Change Notice 2022, written to revise UFSAR procedures to permit this evolution. The change was specific to the control rod mechanism binding observed and provided appropriate options for future corrective actions that may be necessar Licensee engineers prepared Unreviewed Safety Question Evaluation 97-0034 in accordance with 10 CFR 50.59 to address Change Notice 2022. The fuel vendor determined that neither an assembly drop from 2-4 inches nor a 2 4-inch fuel assembly crush would result in fuel pin ruptur The inspectors observed the implementation of the corrective actions from the main control room. Procedure OPMPO4 RX-0019 had been revised to permit the head lift with Control Rod H-2 inserted. Licensed operators were cognizant of potential problems and remained in charge of the evolution. Containment integrity had been established and was maintained throughout the evolutio Conclusions Licensee engineers were conservative in the review and development of corrective actions related to a bound control rod drive mechanism. The unreviewed safety question evaluation properly bounded the evolution being proposed. Licensed operators were cautious and conservative in implementing corrective actions and withdrawing Control Rod H-2.

E2.2 Station Vital Batterv Bank Service Test Review Insoection Scone (37551)

On August 25,1997, licensee engineers informed the inspectors of their intent to perform Technical Specification required testing of the 125 Volt Class 1E battery replacement cells at the factory. The inspectors reviewed the regulations and licensing basis for this change. The following documents were reviewed:

Technical Specification 4.8.2.1.d and related basis document

IEEE Standard 450-1980, "lEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations"

Unreviewed Safety Question Evaluation 97 0024, " Proposed Preoperational Testing Plan for Class 1E 125 vde Batteries"

Plant Surveillance Procedure OPSP06 DJ-0004, Revision 3, "125 Volt Class 1E Battery Service Surveillance Test"

.. .. - . . - . .. - ._ .-

.

- '

,

I -

. Qblervations and Findinos Technical Specification 4.8.2.1.d requires that:

Each 125 volt battery bank and charger shall be demonstrated

'

operable at least once per 18 months, during shutdown, by

'

verifying that the battery capacity is adequate _to supply and .

maintain in operable status all of the actual or simulated ESF loads for the design duty cycle when the battery is subjected to a battery service tes The Technical Specification basis document states that:

The surveillance requirements for demonstrating the operability

~

'

of the station batteries are based on the recommendat!ons of

,

Regulatory Guide 1.129, " Maintenance Testing and

,

Replacement of Large Lead Storage Batteries for Nuclear

Power Plants," February 1978, and IEEE Standard 450 1980,

"lEEE Recommended Practice for Maintenance, Testing, and

Replacement of Large Lead Storage Batteries for Generating Stations and Substations.'

IEEE Standard 4501980 provides the following statement regarding battery service testing:

A service test of the battery capability may be required by the user to meet a specific application requirement t.pon completion of the installatio Licensee engineers had performed Unreviewed Safety Question Determination 97-0024 to justify changing commitments in the UFSAR related to IEEE Standard 400-1980. Engineers proposed that testing the batteries at the factory was equivalent to performing a service test upon completion of the battery installation at the site. The inspectors expressed concern regarding a change to standard industry practice and to the description in the basis to the Technical Specifications.10 CFR 50.59 permits licensee's to modify their plant provided that the proposed change does not involve a change to the Technical Specification As a result of the inspector's questions, the licensee performed service tests of the replaced battery banks upon completion of battery installation in Refueling and Equipment Outage 1REO7. However, engineers maintained that service testing of the station batteries at the f actory was an acceptable alternative that may be

- utilized in the future. Discussions with Region-based and NRR personnel indicated that performing the subject service tests at the factory was not permitted under 10 CFR. 50.59 and would, therefore, require prior NRC approva ,

.. -

31- Conclusions The licensee's decision to perform service tests of station vital batteries at the factory was nonconservative and did not meet the requirements of 10 CFR 50.5 However, upon discussions with the inspectors and other NRC personnel, the licensee performed service tests of the subject batteries upon completion of installation during Refueling and Equipment Outage 1REO7.

E8 Miscellaneous Engineering issues (92903)

E 8.1 (Closed) Insnection Followuo item 498:499/93202-11: Administrative Controls for Backup Pressurizer Heaters While Cold Overpressure Mitigation System was Out of Servic As documented in NRC Inspection Report 50-498/93 202, 50-499/93 202, the operational readiness assessment team had observed that pressurizer heater hand switches were not maintained in the "off" position via an equipment clearance order when the pressurizer power-operated relief valves were inoperable. This item had been designated in the reprrt as Observation 93-202-0 This issue was later addressed as documented in NRC Inspection Report 50-498/96 022; 50-499/96 022. This issue remained open at that time in order to address apparent c'.nflicts between the licensee's submittal requesting the issuance of Technical Specification Amendments 31 and 22 and the related NRC safety evaluation report, Through discussions with the original reviewer for Amendments 31 and 22, the inspectors ascertained that the licensee's commitment to maintain the pressurizer heaters inoperable during water solid operations had been considered material to the license amendment revie Technical Specification Action Statement 3.4.9.3.c requires that, with both power-operated relief valves inoperable, the reactor coolant system be depressurized and vented through a vent of at least 2 square inches within 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months This requirement is applicable in Modes 4,5, and 6 when the head is on the reactor. A footnote states that:

This action may be suspended for up to 7 days to allow functional testing to verify power-operated relief valve operability. During this test period, operation of systems or components wich could result in an RCS mass or temperature increase will be administratively controlled . . .

The inspectors reviewed the licensee's December 21,1990, submittal requesting that Technical Specification 3.4.9.3 be modified to add the subject footnote, in this submittal, the licensee defined the administrative controls that would be

,

.. *

f 32-implemented to assure that the potential for a low temperature overpressure event was minimized during plant heatup and relief valve testing. One of the controls documented that:

The Pressurizor Heaters will be inoperable during water solid operations to minimize the potential for a heat input overpressure transien The inspectors reviewed the control room logs associated with the restart of Unit On January 11,1994, at 11:17 p.m., operators closed the pressurizer vents and entered Technical Specification 3.4.9.3 action statement. The control room log indicated that the specification required that the pressurizer power operated relief valves be declared operable within 7 days. However, the log failed to document the required administrative controls. On January 12,1994, at 2:52 a.m. operators documented that pressurizer pressure was stabilized at approximately 100 psig, indicating a water-solid reactor coolant system. The system remained in a water solid condition until January 14, at 9:33 a.m. when a pressurizer bubble was established. During this time, as documented in NRC Inspection Report 50 498/93 202;51-499/93 202, the pressurizer heaters had remained operable. Although the plant operating procedures provided limited administrative controls over systems that could result in a reactor coolant system mass or temperature increase, the f ailure to maintain the pressurizer backup heaters inoperable while at water solid conditions occurred. The inspectors reviewed current procedural controls and determined that they remain limited and do not specifically state, and therefore require, that pressurizer heaters be inoperable during water solid operations. Th> licensee has committed to modify Plant Operating Procedure OPOP03-ZG-0001, Revision 16, " Plant Heatup," to include a checklist which willinclude procedural steps that will assure implementation of the administrative controls required by Technical Specification 3.4.9.3.

E8.2 (Closed) Unresolved item 498:499/96004-03: Advanced Liquid Waste Processing System (ALPS) was not Described in the UFSAR Insnection Scope (92903)

This item addressed deviations from the UFSAR description of the originally designed and installed liquid waste processing system. The inspectors reviewed the system history, licensee design activities, and the licensee's resoonse to the open inspection items. The following documents were reviewed:

e Temporary Modification Request TI WL SO28, " Processing Liquid Radwaste through Temporary Demineralizer Skid Prior to Discharge to Waste Monitor Tanks" e 10 CFR 50.59 Screening Form for Chem Nuclear Procedure DM-OP-022

,

.. .- -. - .- - - .. . - . .

i

. . .

, .. .

_ .3 3 .

!

,

'

e Design Change Pacliage 94-4232 3,'" Design Review of the Advanced Liquid Processing Systems"-

e= Condition Report 9615062, " Revise and Update UFSAR 11.2 - Liquid Waste Management Systems"

,

e UFSAR Change Notice 2119, "More Accurately Describe Liquid Waste Processing Options and Philosophy" e Condition Report 97-7826, " Nuclear Regulatory Commission Resident inspector has identified Two Concerns with the UFSAR LWPS Descriptions"

<

e UFSAR Change Notice 2160, " Integrate the Description of the Mobile Filtration / Demineralization Unit into UFSAR Section 11,2" e Unreviewed Safety Question Evaluation 97 0018, " Change Notice 2160 -

Liquid Waste Processing System Mobile Filtration /Demineralizer System" ,

i e Plant Operating Procedure OPOP02 WL-0026, Revision 6, " Operation of the Advanced Liquid Processing System" b. System Deslan and Characteristics in September.1992, the licensee connected a mobile filter /demineralizer system via hoses to the liquid waste processing system. The system consisted of large

. shielded filter /demineralizers that were designed to process radioactive effluent prior to release. The system equipment housed approximately 130 gallons of water and 200 cubic feet of demineralizer resins. The equipment was located in the mechanical auxiliary building truck bay and connected to existing liquid waste processing system fitting The ALPS was originally intended to be a temporary system, as documented in Temporary Modification Request TI-WL 92-028. This document estimated that the L original system would be restored by December 6,1993. The demineralizers had been leased from the vendor, and the operation of the system had been the responsibility of the vendor. The truck bay was provided with curbs sufficient to-contain the tank contents ar.d with the capability to return spillage to the permanent liquid waste processing system. The licensee has since purchased the system and trained reactor plant operators to control the system components. The -

system, originally designed to augment the installed equipment, is frequently used instead of the original portions of the liquid waste processing syste . c. Backaround-The licensee originally determined that the temporary use of the ALPS system did

, not involve a change to the f acility as described in the safety analysis report. This

. -- _

. . - -

.. . -_ . - . . ,

- .- . - , . - .. . - - .

<

.

,

. 3 4. .

finding was based oathe classification of the system as nonsafety related_and the i

'

capability of the truck bay curbs to contain any system spillage. Additionally, the-reviewers determined that the equipment and hoses would not have changed the existing system reliability.

'

In May 1995, licensee engineers again reviewed the ALPS design as a result of severalinternal audits. However, no new engineering effort was accomplished at that time and engineers only reiterated the positions noted in August 1992. At that

,

time, the reviewers did acknoviledge that the system had been purchased and vendor supplied drawings were added to the document control syste As a result of the inspector's origine concerns, licensee representatives documented the concerns in Condition Report 96 7788 and initiated UFSAR Change

Notice 2035. This change notice added a statement that a mobile filtration /

domineralizer unit existed and processed potentially r' % active liquid wastes for discharge to the main cooling reservoir. The licenses gain found that the change-did not involve a change to the facility as described in the safety analysis report-because no physical changes were made to the facility at the time that the evaluation was performed.

'

Following further discussions with the inspectors, licensee engineers issued Condition Report 97-7826 and UFSAR Change Notice 2160 was initiated. This document added and integrated additional description of the Mobile Filtration / Demineralization unit (ALPS) into UFSAR Section 11.2 and into the general plant arrangement drawings located in Chapter 1 of the UFSAR.

Unreviewed Safety Question Evaluation 97-0018 was developed to ensure that the requirements of 10 CFR 50.59 were being met, Observations and Findinas The inspectors reviewed Unreviewed Safety Question Evaluation (USOE) 97-0018 and associated UFSAR Change Notice 2160. These documents properly documented the de:ign and current usage of the ALPS and were responsive to the inspector's questions posed in NRC Inspection Report 50-498/96-04:

50-499/96-04. The questions related to the f ailure of the ALPS to meet the codes >

and standards utilized to design the liquid waste processing system. The inspectors reviewed the response to each deviation from UFSAR commitment as documented-below:

(1) The components of ALPS were not contained in Seismic Category I structure Change Notice 2160 updated the UFSAR to document that the ALPS was

, located in a nonseismic Category I structure, in USOE 97-0018, licensee -

engineers documented that the strength of the mechanical auxiliary building

. - - - - - _

. . . _

. - .

- - - . _ . . . - . . - .

.

- F 35-E truck bay did, however, exceed the increased loading that would occur during an operating basis or safe shutdown earthquake.

-

(2). A spill _ from the ALPS would not be returned to the LWPS by connecting '

floor drains because licensee personnel maintained the building drains covered with plastic taped to the floo Operators had previously removed the plastic from the building floor drain Licensee engineers documented in USOE 97 0018 that the licensing basis for the system was for spill retention and not spill recovery. Curbs had been provided around the perimeter of the building.' These would retain, at a minimum, the spill of the entire contents of the ALPS. Spills larger than the contents of the system, including continued supply of waste to a broken

, system, were also reviewed. Engineers concluded that the consequences of these postulated events remained bounded by previous Chapter 15 accident analyse (3) Materials for the ALPS did not meet the specifications for materiallisted in Section ll of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,197 Licensee engineers documented in USOE 97 0018, that an alternate method of compliance was utilized. Hoses were operationally leak tested anytime they were disconnected and reconnected to the plant systems. The

,

inspector reviewed Procedure OPOP02 WL 0026. The opera *;onalleak checks were documented in Section 8.0. This change was documented in Change Notice 216 (4) The ALPS was not provided with instrumentation for verifying differential pressure across the demineralizers and tank levels were not indicated in the radioective waste control room.

.

Change Notice 2119 updated the UFSAR to change the term " pressure !

differential" to " pressure." However, it was not clear that this provided

,

appropriate protection for indicating degradation of the demineralize Enginects also stated that the ALPS vessels were not tanks. Therefore, indication in the radioactive waste control room was not necessar (5). The design of the liquid waste processing system components and the l seismic design of the structures were found to be material to the original

NRC licensing determinations and findings, as documented in NUREG-0781

" Safety Evaluation Report related to the operation of South Texas Project, Units 1 and 2."

Licensee engineere documented in USQE 9'7-0018 that there wue no quality i group classification changes associated with ALPS as compared to the

l

!

. , - - ~ - ,

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

+. .

.-

-3 6-remainder of the liquid waste processing system, in addition, the engineers

-

stated that the mechanical auxiliary building truck bay met the requirements-of Regulatory Guide.1.143 and, therefore, was acceptabl On April 15,1997, the inspectors toured the mechanical auxiliary building truck bay .;

e in Unit 1. The inspectors noted that the solid waste processing system had been in service. However, the truck hay doors had been open and the continuous air monitor had not been in service. The inspectors reviewed the licensee's response to Safety Evaluation Report Confirmatory item 22, dated April 22,1987. This .

response indicated that the truck bay would be maintained at a slightly negative pressure. Additionally, a letter dcted May 20,1987, related to the same confirmatory item stated that the area would be monitored and locally alarmed by a continuous air monitor during operation of the solid waste processing syste Licensee personnel stated that these commitments were also applicable to ALPS

. processing. These concerns were appropriately addressed in Condition

-

Report 97 7239 written at that time. However, the failure to properly control the use of the continuous air monitor and maintain a negative pressure in the truck bay during solid waste processing operations constituted additional examples of a deviation from the licensing basis document CFR 50.59 states that the holder of a license authorizing operation of a utilization facility may make changes in the facility as described in the safety analysis report, without prior Commission approval, unless the proposed change involves a change in the Technical Specifications or an unreviewed safety questio Paragraph (b)(1) further states that the licensee shall maintain records of changes in the facility and of changes in procedures made pursuant to this section to the extent that these changes constitute changes to the facility as described in the safety analysis report. These records must include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety questio The inspectors found that the ALPS was a vital portion of the liquid waste processing system that, over time, had replaced portions of the original processing equipment, without clearly meeting the associated UFSAR commitments. Licensee personnel had continued to troat the system as temporary, although no plans or work; documents were outstanding to return the original equipment to servic Additionally, this change to the facility was not evaluated to determine lf the change involved an unreviewed safety question because plant workers had indicated that ALPS did not constitute a change to the facility as described in the safety analysis report. The f ailure to provide a written safety evaluation of the ALPS from its installation in September 1992 untli USOE 97 0018 was documented on August 20,1997, and was in violation of 10 CFR 50.59 (498:499/97006 08),

, . , - , -- - - ,. - .

. - . . - . . . . . _ . ..n__ - --- -

!

,. .. -. ;

37-

.

The insper* ors will continue to review the adequacy and acceptability of the ALPS

.

design. Ms review willinclude a thorough evaluation of Charse Notice 2160, the licensee's response to this violation, and other information related to the design of

_

the syste , Conclusions The inspectors concluded that the installation of ALPS in September 1992 _

!

constituted a change to the facility as described in the safety aralysis report. The 1 licensee's failure to document a written safety evaluation of the ALPS was a ;

'

violation of 10 CFR 50.59. Associated examples of failure to ensure that design related commitments were implemented in the field snd deviations from UFSAR

'

commitments were identified. A further review of the ALPS design will be

necessary to evaluate its adequacy and acceptability.

E8.3 (Closed) Unresolved item 498:499/96006-04: Review Implications of not Providing Safety related Seals for Spent Fuel Pool Gate ,

a, 'Insoection Scoce 192903)

This item documented cuncerns related to testing of nonsafety-related spent fuel pool gate seals prior to th' nplementation of certain corrective actions associated with a previous loss of pod, inventory event. The item discussed several additional i questicns related to the design of the spent fuel pool gate seals and the acceptability of the this existing design. The inspectors reviewed each of the questions raised in the unresolved item, associated documentation, and a licensee letter dated December 16,1996, addressing the subject, in addition, the inspectors performed a field inspection of the spent fuel pool area, Backaround As documented in NRC inspection Report 40-498/95 20; 50 499/95 20, on-July 18,:1995, problems with the control of spent fuel pool gate seals had caused a loss of inventory in the Unit 2 pool. Deficiencies in the licensee's followup and corrective actions related to this event resulted in a special inspection documented in NRC Inspection Report 50-498/95 21;50-499/95-2 During this special inspection, the inspectors determined that the implementation of engineering specified actions were not achievable under the circumstances. At that time, licensee engineers were evaluating options for providing an additional barrier netween the spent fuel cool and the unfinished cask handling area. However, the -

Inspectors had concludsd that the design was adequate because Section 9.1 in NUREG 0781, " Safety Evaluation Report Related to the Operation of South Texas -

Project,_ Units 1 and 2/ dated April 1989, stated that the gates between the storage pools and cast loading pool were acceptable, u _ _

._ -. - _ _ _ - _ . _ _ . _ _ _ _ _ _ _ _ .._

<

.

,, .: -;

,

i-38 :

!

,

I In July 1996, the inspectors further reviewed this design. A recent d5covery of a.

.

- - -

licensee operating their f acility in a neanr.ar contrary to t's UFSAR description had' ,

highlighted the need for an NRC focused review of UFSAR descriptions.> Th l

, -inspectors determined that the UFSAR indicated that the spsnt fuel pool gates were

'

safety related; yet was silent on the class 1(ication of the gate seal. In acdition, the-general plant drawings in the UFSAR inoirted that a third watertight ba rier, tha i

cast channel gate, was in place. Documents discussing the status of tho cask handling creas, at the tima of licensing, had not indicated that his barrier was not ;

d-installe As documentedin NRC inspection Report 50 498/96-06;50 499/96 06, '

misunderstandings concerning the scope and status of licensee commitments in this area were identified. Additio. Tally, a lack of timeliness in constructing the third barrier between the pool and the cask handling areas was noted.

4 Observation and Findinos ,

Following the issuance of NRC Inspection Report 50 498/96 06; 50-499/96-06, the

licensee accelerated the schedule to fabricate end install a permanent, passive,

stainless steel wall between the spent fuel pool and cask handling areas in each

-

unit, in a response letter dated December 16,1996, the licensee documented that the installation of the walls was complete. The inspectors verified, by visual >

inspection, that the walls were in place and appeared to be sound. . After this

=

installation, no immediate safety question remained. The licensee's response letter adequatelv addressed the misunderstandings associated with prior commitments,

'

in addition, each of the remaining questions were addressed as followed:

,

e What design criteria was utilized for the seats during the originallicensing review?

, .

Although this question was posed to address the original NRR safety evaluation, the licensee stated that they had implemented Regulatory

Guide 1.13, " Spent Fuel Storage Facility Design Basis." The inspectors

.

agreed that the failure of the spent fuel pool gato seals would not result in-fuel uncovery. However, the results of the failure would have been considered unacceptable prior to the installation of the walls. The installation of the walls has rendered this question mut * What would be the final water levelin the spent fuel pool should both

,

nonsafety related seats fail?

'In their response, the licensee stated that the resultant water levelin the.

spent fuel pool following a postulated catastrophic failure of the spent fuel pool to cask handling area gates would be approximately 47 feet. This was spproximately 10 feet above the active fuel. This calculation assumed that allleak paths in the cask handling area were isolated and fluid r 'ssorption by

7

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

- - - - _ _ . _ _ _ _

- - .. .- . . . ~ _ . _ - - - - , - - . - - - ..

,

'o

.

39-utilined concrete _was negligible. Based on the as found configuration of the area drain system, and the incomplete status of the area, the inspectors had questioned the validity of these assumptions. However, these questions were also rendered mute by the installation of the third barrier wall. The wall would contain such a failure in a small stainless steel clad area.

t

-* Was the cask connecting channel gate considered a critical part of the .

originally accepted design?

In their retiponse, the *!cersee stated that the cask connecting channel gate was only designed to be used in preparation for offsite transport of spent fuel. During these activities, this gate would be required to be installed and *

capable of maintainir.g water inventory in the spent fuel pool and cask handling area. Further, since no federal repository existed at the time of license review, South Texas Project did not consider the cask connecting channel gate to be a critical part of the original or current spent fuel pool desig Again, this questien was posed regarding the material components of the original NRR safety evaluation, as opposed to the licensee's design inten This question was, also, rendered mute with the installation of the barrier wal * Has long-term exposure of the unfinished cask handling area surfaces to spent fuel pool boric acid affected the integrity of the reinforcing steel?

In their response, the licensee stated that the unfinished areas of the cask

. handling area have not been subjected to long term exposure to boric acid.

The cask connecting channel has experienced limited boric acid exposure .

during nate/ seal maintenance or testing activities. These exposures have not resulted in damage to any exposed concrets due to the extremely INted exposure and evaporation time frames. No deleterious effects of the expasure have been noted during engineering inspections of these area Tha inspectors observed that there was not currently boric acid contacting ,

any of the unfinished surfaces and that the new barrier wall prevented spent fuel pool water from contacting the unfinished surfaces of the cask handling

,

areas,

Has the licensee performed a flooding analysis considering the potential for loss of the spent fuel pool gate seals?

In their response, the licensee stated that South Texas Project personnel had thoroughly inspected all portions of the cask handling area and found no equipment that would be affected following postulated flooding from the

-. . _ . .

-- . . . _ _

. _ . .

- - . - .. .. , - - . - - . - - - - - - - -, .

k 4 eg .

l l

-40

_

spent fuel pool. The cask handling area was essentially a series of concrete vaults with no installed equipment required for plant operation, Based on ,

- these inspectionsi further flooding analysis is not warrante The inspectors further reviewed the general plant layout drawing Postulated flooding beyond the cask handling area would have spilled into the fuel handing building truck bay and out of the bay doors, No critical components would have been affected. Likewise, flooding from a postulated uncontrolled drain system would have been contained within a watertight vault beneath the cask handling area. However, the potential for flooding

,

was also made mute by the installation of the barrier wal d, Conclusions The inspectors concluded that the licensee's corrective actions, including the f aNication and installation of a permanent, passive, stainless steel wall between th cask connecting channel and the remainder of the unfinished cask handling areas

, resolved any open immediate safety issues. The remaining questions regarding the original design were rendered mute by this action, ly. Plant Support R1 Radiological Protection and Chemistry Controls f

R1,1 Tours of Radioloalcal Controlled Areas

, Insnection Scone (71750)

The inspectors routinely toured the mechanical auxiliary and fuel handling buildings in Units 1 and 2. The inspectors also toured the Unit 1 reactor containment building dering Refueling and Equiptnent Outage 1REO7. These tours included observation of work, verification of proper radiological work permits, sampling of locked doors, reviews of area surveys and postings, and observation of personnel entrance and egress from contaminated areas and the radiological controlled areas, Observations and Findinas

' Radiological housekeeping in the areas toured was very good. Doors required to be locked in accordance with Technical Specification 6,1,12.2 and the licensee's

. radiological program were properly secured. Obse:ved work was performed in accordance with proper radiological work permits. No entrance / egress discrepancies were identifie The inspectors rorninely toured the Unit 1 reactor containment building during Refueling and Equipment Outage 1REO7. The licensee established radiological

_ _

_ . _ .

. . . - . .- . -- - -_ - .

,

.. . e .. >

-41

- control check points manned by health physics technicians at the 68 foot elevation near the entrance to the reactor containment building and at the 19 foot elevation near the bloshield entrance. The technicians at both check points were cognizant of radiological conditions in their designated areas. The inspector observed the technicians as they provided thorough radiological protection briefings to crews-preparing to work in the reactor containment building. The briefings included reviews of radiological surveys that were frequently updated to provide the latest >

information. The inspector observed technicians as they closely monitored work in contaminated and high radiation area On September 24, the inspectors observed health physics technicians providing support for the repair of the fuel transfer cart in the Unit 1 fuel handling buildin This job required divers to work on the cart at the bottom of the fuel transfer cana The diver had several radiation detectors attached to his suit to monitor radiation exposure during the dive. The detectors provided real time data to the health physics technicians on the fuel handling deck. The health physics supervisor maintained continuous oversight of the activities, Conclusions The inspector considered that the radiological protection efforts related to divers 4 entering the spent fuel pool were excellent. Other observed radiological protection performance was stron R7 Quality Assurance in Radiological Protection and Chemistry Activities

'

R7.1 Transoortation of Contaminated Machine Tools {71750)

On September 19, licensee representatives received a call from a vendor informing -

them that a snubber testing machine and associated tools had left South Texas Project contaminated with low levels of fixed contamination. The machine had arrived at another nuclear power plant the previous day. During a receipt

,

inspection, that licensee's heath physics technicians had detected the contamination. The highest reported contamination was reported to be 800 counts per minute on a vise. Indications are that the vendor was not licensed to possess byproduct materials. This event will be reviewed further oy NRC radiation 2 protection inspectors and will be tracked as an inspection followup item (498:499/97006 09).

P2 Status of EP Facilities, Equipment, and Resources P2.1 Emeraenev Resoonse Facilities (71750)

During area tours, the inspectors observed that the Technical Support Centers and Operations Support Centers in both units were readily available and maintained for

' emergency operatio . . --

.

-. . . -- - - . - -

,. .. ,.

'

S1 Conduct of Security and Safeguards Activities 1

- S Daily Physical Security Activity Observations (71750)

.

On a daily bssis, the inspectors observed the practices of security force personnel and the condition of security equipment. . Protected and vital area barriers were in good condition. Persennel access measures and equipment searches for ,

contraband were observed on a daily basis. The inspectors concluded that daily security force activities were conducted in an appropriate manne V. Manaaement Meetinas

- X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee <

management at the conclusion of the inspection on October 7,1997. The licensee acknowledged the findings presented. Licensee representatives stated that they did not agree that the f ailure to perform surveillance tests on valves associated with certain containment penetrations was In violation of the Technical Specifications because the specification specifically exempted certain penetrations, However, management agreed that the verification of the valve positions was a good practic In addition, engineering representatives stated that performing service tests of station vital battery replacement cells at the factory instead of upon completion of the installation was still considered a viable alternative. The inspectors informed those members of management orasent that the NRC would take their disagreements under adviseme The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

.

w- ri-

- . . - . .-. - - . . - - -. .

'

,

8.. ,

s ATTACHMENT

SUPPLEMENTAL INFORM ATION PARTIAL LIST'OF PERSONS CONTACTED

,

Licensen i T. Cloninger, Vice President, Nuclear Engineering J. Cook, Supervising Engineering Specialist

. W. Cottle, President and Chief Executive Officer B. Dowdy, Manager, Plant Operations, Unit 2 J. Groth. Vice President, Nuclear Generation E. Halpin, Manager, Maintenance, Unit 2 S. Head, Senior Consulting Engineer K. House, Supervising Engineer, Design Engineering Department

.M. Kanavos, Manager, Mechanical / Civil Design Engineering M.- Lashley, Manager, Reliability Engineering D. Leazar, Director, Nuclear Fuel and Analysis

'

B. Logan, Manager, Health Physics R. Lovell, Manager / Plant Operations, Unit 1 8. Masse, Plant Manager, Unit 2 A. McIntyre, Director, Engineering Projects G. Parkey, Plant Manager, Unit 1 D. Rencurrel, Manager, Electrical /l&C Design Engineering F. Timmons, Manager, Nuclear Plant Protection T. Waddell, Manager, Maintenance, Unit 1 INSPECTION PROCEDURES USED IP 37551: Onsite Engineering-IP 61726: Surveillance Observations IP - 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support ~

IP 92700: Onsite Followup of Written Reports at Power Reactor Facilities IP 92901: Followup .- Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering ITEMS OPENED, CLOSED, AND DISCUSSED

'

Opened-498;499/97006 01 VIO Two examples of Inappropriate Use of Equipment Clearance Orders vice Using a Proper Procedure

.

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

. - _ _

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

,

j. y *

-2-

,

, 498/97006-02: ~NC Inadvertent Mode Change lirom Mode 4 to Mode 31 l While Placing the Residual Heat Removal System in -

'

' Service ' ,

498/97006-03 NCV Failure to Verify the Operability of Offsite Power'

' Sources when a Vital Battery was Removed from

. Service

"

- 4Lo,499/970.06 04. VIO Manual Valves in Certain Containment Penetrations not Surveilled in Accordance with Technica Specification 4.6.1. /97006 05 NCV Reactor Containment Building Airlock incorrectly Returned to Service while Inoperable .

'

499/97006-06 NCV Improperly Installed Jumper on a Main Steam Line

_

Pressure Solid State Protection System Circuit Card 498:499/97006-07 URI Failure to inspect Standby Diesel Generator Heat Exchangers for Biofouling in Accordance with Commitrnents 498:499/97006-08 VI Failure to Document a Written Unreviewed Safety

,

Question Determination upon Installation of the Advance Liquid Waste Processing System 498:499/97006-09 IFl Review the implications of Contaminated Tools being

, Released from the Site Closed 498/97006 02 NCV Inadvertent Mode Change From Mode 4 to Mode 3

, While Placing the Residual Heat Removal System in Service

,

498/96003-02 VIO Failure to Properly Evaluate the Effects of an Out of

.

-

Calibration Condition

'50-498/96-001 LER Failure to Verify the Operability of Offsite Power Sources When a Standby Diesel Generator was inoperable-498/97006-03 NCV Failure to Verify the Operability.of Offsite Power Sources when a Vital Battery was Removed from Service

__

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

,o '. '

3-498 499/9303136 IFl Senior Shif t Managers will Provide Continuous Management Representation and Presence during Selected Evolutions Throughout the Power Ascension Program 498;499/97005-02 URI Manual Valves in Certain Containment Penetrations not Surveilled in Accordance with technical Specification 4.6.1.1.a 498/90004-04 VIO Mid Loop Level Transmitters not Placed in Service 499/97001-01 VIO Reactor Coolant System Level Sight Glass not Properly Aligned during Mid loop Operations 50-499/97-006 LER Manual Reactor Trip initiated upor Malfunction of a Main Feedwater Regulating Valve 50-498/96-005 LER Reactor Containment Building Personnel Airlock Incorrectly Declared Operable 498/97006-05 NCV Reactor Containmant Building Airlock Incorrec'8y Returned to Service while Inoperable 50 499/96 003 LER improperly Installed Jumper on a Main Steam Line Pressure Lead / Lag Circuit Card 499/97006-06 NCV improperly Installed Jumper on a Main Steam Line Pressure Solid State Protection System Circuit Card 498:499/97002-01 VIO Repaat Violation of inadequate Containment inspection Regarding Debris not Removed that Could Affect Containment Sump Operability 50-499/97-003 LER Failure to Meet the Requirements of Technical Sr.ecification 4.5.2.c Requiring an Inspection of Containment for Loose Debris 498:499/93202-11 IFl Administrative Controls for Backup Pressurizer Heaters while Cold Overpressure Mitigation System was Out of Service 498;499/96004 03 URI Advr.nced Liquid Waste Processing System was not Described in tiie UFSAR 498:499/97006 04 URI Review ImpheatDns of not Providing Safety related Seals for Spent Fuel Pool Gates 7772

v t e Inspection Report - South Texas - 1997006
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1997 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 99(0) Security Operator Exams 301(0) Emergency Planning Other

IR 05000498/1997006

Text

REGION IV

Navigation menu

Search