IR 05000206/1989014

Insp Repts 50-206/89-14,50-361/89-14 & 50-362/89-14 on 890430-0617.No Violations Noted.Major Areas Inspected: Operations Program,Including Radiological Protection, Security & Evaluation of Plant Trips & Events

ML20245G707
Person / Time
Site:	San Onofre
Issue date:	07/24/1989
From:	Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20245G706	List: IR 05000206/1989014 ML20245G703
References
50-206-89-14, 50-361-89-14, 50-362-89-14, NUDOCS 8908160172
Download: ML20245G707 (16)

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION V

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Report No C'6/89-14, 50-361/89-14, 50-362/89-14 -

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50-206, 50-361, 50-362

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Docket No l

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License No OPR-13, NPF-10, NPF-15 -

Licensee: Southern California Edison Company <

P. O. Box 800, 2244 Walnut Grove Avenue 4 Rosemead, California 92770 <

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3 Facility Hame: San Onofre Units 1, 2 and 3 >

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Inspection at: San Onofre, San Clemente, California Inspection conducted: April 30, 1989 through June 17, 1989 Inspectors: F. R. Huey, Senior Resident Inspector, Units 1, 2 and 3

'J. E. Tatum, Resident Inspector A. L. Hon, Resident Inspector Approved By: 7 P. inson, Chief Date Signed React r Projects Section 3 j Inspection Summary Inspection on April 30 through June 17, 1989 (Report Nos. 50-206/89-14, 50-361/89-14, and 50-362/89-14)

Areas Inspected: Routine resident inspection of Units 1, 2 and 3 Operations Program including the following areas: operational safety verification, radiological protection, security, evaluation of plant trips and events, monthly surveillance activities, monthly maintenance activities, refueling activities, independent inspection, licensee event reports review, and followup of previously identified items. Inspection procedures 30703, 35502, 37700, 37828, 61705, 61726, 62703, 71707, 71710, 82301, 90712, 92700, 92701 and 93702 were utilize Safety Issues Management System (SIMS) Items: None 8908160172 PDR 89072k '

ADOCK 05000206 G PNU

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-2-Results:

General Conclusions and Specific Findings: A weakness in interfacing between operations and engineering occurred after a plant system was modified.. It resulted in nonconservative equipment operability determination (paragraph 3.a). . Inadequate engineering involvement in startup of newly installed equipment was identified. One example is an auxiliary feedwater initiation (paragraph 8) and the other is the nuclear instrumen-tation system (paragraph 8). Inspectors observed an inadequate safety evaluation in support of continued plant. operation with known deficiencies (paragraph 9). An inadequate root cause evaluation was performed for LER 206/89-15 (paragraph 10)

Significant Safety Matters: None Summary of Violations: None Open Items Summary:

During this report period, 8 new followup ' items were opened and 6 were '

closed; 6 were examined and left open. One item was examined and left unresolve .:

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. DETAILS y

'I..' Persons' Contacted

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. Southern California Edison' Company C. McCarthy, Vice President and Site Manager

<< , *H. Morgan, Station Manager-

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D.-Herbst, Quality Assurance Manager D.cStonecipher, Quality-Control- Manager

' *R'.Krieger, Operations Manager

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D. Shull,' Maintenance Manager C. Chiu, Assistant Technical Manager P.'Knapp, Health Physics Manage 'D. Peacor,-Emergency Preparedness Manager

' P.' Eller, . Security Manager

• J. Schramm 0perations Superintendent, Unit 1

• V. Fisher, Operations Superintendent, Units 2/3

• L.. Cash,. Maintenance Manager, Unit 1 R. Santosuosso, Maintenance. Manager, Units 2/3

• R. Plappert, Compliance Manager

• R.' Baker, Compliance Engineer .

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San Diego' Gas and Electric Company

• J. Winter, Site Representative

• Denotes those attending the exit meeting on June 16, 1989'.

The inspectors also contacted other licensee emplo'yees.during the course-of the inspection, including operations shift superintendents, control-room' supervisors, control room operators, QA and QC engineers,

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compliance engineers,-maintenance craftsmen, and health physics' ,

engineers and technician . Plant Status - .

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Unit 1 .

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The unit entered Mode ~ 2 following the Cycle X refueling outage on. May

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16. During control rod withdrawal for the approach to criticality,e electrical interference on the new nucleer instrumentation system (NIS)

caused spiking on the intermediate range channels. The unit was returned to Mode 3 to troubleshoot and repair the noise proble Following completion of. low power physics testing, the unit was synchronized to the grid on May 25. On May 26, the unit was returned to Mode 5 to repair auxiliary feedwater automatic actuation circuitry and to' repair reactor coolant leakage from the B reactor coolant pump seal and the pressurizer safety valves. The unit remained shut down through the end of the inspection perio _ - _ _ - _-_

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Unit 2-

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, The unit 4 had 89 days of continuous power operation until May 12, when

,; primary to secondary leakage on steam generator E088 was observed to increase to approximately 700.gpd. The unit was placed in Mode.5 to y repair'the steam generator tube leaks and. returned to power operation on

. June 7.

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Unit 3-

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L The. unit operated throughout the inspection period,' having accumulated

[j 61 days of continuous power operation at the end of the perio *

I Operational Safety Verification (71707)-

p The inspectors performed several plant tours and verified the operabi- ,

, lity of selected emergency systems, reviewed the tag out log and , ,

verified proper return to service of affected components. Particular

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attention was given:to housekeeping, examination for potential ^ fire

,"; x 4 hazards, fluid leaks, excessive vibration, and verification = that  !

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, maintenance requests had been initiated for equipment.in.need of

,- maintenance...The inspectors also observed selected activities by  ;

h licensee radiological protection and security personnel to confirm y~

' proper implementation of and conformance with facility policies and "~

g procedures in these area Nonconservative Equipment Operability Determination #

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, . During the recent Unit I refueling outage (Cycle X), the steam

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generator level instrumentation associated with the auxiliary feedwater (AFW) system was modified. During plant startup from the L .

outage (May 25), the inspector noted that plant operators did not take timely action to involve cognizant engineering or technical

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l personnel following observed malfunction of the new steam generator

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level instruments. In particular, although plant operators had -

l observed improper instrument performance following Mode 2 entry on May '21, they did not properly monitor instrument performance or I effectively involve appropriate engineering personnel in evaluation  ;

of the observed deficiencies or properly assess the potential

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impact of instrument malfunction on spurious safety system

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.j actuation,or potential water hanrner concerns (if actual steam ]

generator level was less than 26% NR level). As a result, {

l warranted corrective actions were not initiated until after the j occurrence of a spurious AFW actuation on May 25.

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This incident was also indicative of inadequate operations super-

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visory involvement in timely correction of deficiencies affecting recently modified equipment. During the exit meeting, the inspec-tor noted that this event appeared to be similar to the failure of plant operators to properly respond to low Freon levels in Unit 2/3 emergency chiller units, which was previously addressed in NRC i inspection report 50-261/88-35, dated May 5, 198 l

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C The~ station Operations Manager agreed that additional actions and closer attention by plant operators following the observed

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instrument malfunctions were warranted and he committed to implement corrective actions to preclude recurrence of this

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problem. This item is closed (206/89-14-01). Inadequate Post-Modification Training of P1 ant Operators

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During the recent Unit I refueling outage (Cycle X), the; unit's nuclearinstrumentation(NI)systemwasreplaced. During plant startup from the outage (May 16), it appeared to the inspector that .

control room operations personnel were not as knowledgeable about

.the operation of the new NI system as would have been expected. In particular, on-shift control room supervision did not appear to fully understand the interaction of power range channel bypass, functions with reactor' trip functions, or whether more than one power range channel was capable of being bypassed simultaneousl The inspector raised this concern with the station Operations Manager, who agreed to reassess the adequacy of post modification training of plant operators. This item remains open, pending completion of the licensee's review (206/89-14-02). Lack nf Timely Action to Address Technical Specification Problems The inspector raised a concern with respect to two instances in which SCE failed to take timely action to properly address and correct observed technical specification problems. In particular:  :

(1) On May 5, SCE informed the inspector of a problem involving the failure to properly test the Unit I reactor protective system (RPS) Channel B shunt trip signal to the reactor trip breakers. However, SCE did not initiate action to correct the inadequacy in the unit's Technical Specifications, which contributed to this deficiency, until prompted by the inspecto (2) On May 17, SCE informed the inspector of a problem involving improper adjusting ring settings on Unit 2/3 steam safety valves, which resulted in failure to meet the valve relief capacity requirements of the Technical Specifications'. 3 Although SCE had been aware of this problem for several weeks, j they did not properly address the implications of valve {'

inoperability in accordance with the Technical Specification In particular, although the licensee considered that the provisions contained in the Technical Specifications for valve relief capacity were incorrect, SCE did not initiate action to correct the deficiency until prompted by the inspecto )

During the exit meeting, the Station Manager acknowledged that SCE )

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should have taken more prompt action to correct the Technical  !

Specifications inadequacies. However, he stated that SCE had recognized the deficiencies and had established tasks to submit ,

appropriate Technical Specifications changes. The Station Manager j

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stated that more aggressive action would be taken in the future,-

particularly for problems similar to item (2) above. . This item is closed (361/89-14-01). Improper Sense of-Equipment'" Ownership" by Plant Operators During discussions with Unit 3 plant operators on May 14, the inspector-noted that control room operations personnel did not ;

appear to have a good feel for deficiencies associated with the Qualified Safety Parameter Display System (QSPDS) or the potential impact of observed deficiencies on OSPDS operability. Plant operators indicated that they were not involved in the surveillance or evaluation and tracking of OSPDS operability, and stated'that this function was the responsibility of plant computer technicians. 1 The inspector reviewed this concern with the station Operations

., Manager, who agreed with the inspector's concern and stated that he would take action to increase plant operator involvement in.the surveillance, evaluation and tracking of QSPDS operabilit This item remains open, pending completion of licensee action (361/89-14-02).

No violations or deviations were noted during the inspectio . Evaluation of Plant Trips and Events (93702) s Spurious Reactor Trip During Control Rod Exercise (Unit 1)

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On May 4, while conducting control rod drop test, an automatic reactor trip occurred as a result of a spurious'high indicated level and startup rate on both intermediate range (IR) instrument channels 1203 and 120 The licensee conducted troubleshooting to identify the source of

• the problem. The licensee concluded that the problem was the result of electromagnetic interference (EMI) from ~two different

'1 , sources. The first involved noise generated by the rod position analog to pulse stepping motor. The second involved the alarm relay for rod position deviation circuitry. The licensee was able y to duplicate the trip condition by actuating these components. To prevent recurrence, the licensee installed additional noise suppression devices on the appropriate components, Unit Shutdown Due to Excessive Primary to Secondary Leakage (Unit 2)  !

On May 13, while the unit was at 100% power, plant operators )

observed steam generator E088 blowdown radiation monitor increase j by one decade. The Shift Superintendent directed power reduction !!

to 80% and sampled steam generator blowdown chemistry. The sample indicated an approximate 540 gallon per day (gpd) primary to secondary leakage. A second sample was analyzed at 2:25 a.m. and

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indicated approximately.700 gpd leakage;.the licensee then. decided-

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z to shut'down the unit, anticipating that letkage was likely to

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exceed the. Technical Specification.LCO.of 720 gpd. -The unit was

. placed in Mode' 3 at 6:36 a.m. and the' leakage us stabilized at-

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- approximately .81 gpd without- further anomalie The licensee cooled down the.' unit to. Mode'5 'on May 14 and in.'tiated 1 draining.to mid-loop-in' order to identify and repair the leaking tube. The licensee'found three welded tube plugs which were leak '

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ing due to defective welding'. The three plugs were part of the 78'

~ welded plugs installed by Combustion. Engineering.(CE) in~198 ..

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(The licensee's written analysis in support of continuing operation-

' with.the remaining welded plugs ,in place is, discussed further in

paragraph 9.a)...The unit was returned to service on June 7, 19891

'following a 25' day outag ,

No' violations or deviations.were noted during the inspectio . . MonthlySurveillanceActivities'(61726).

Diaring this'; report period, the inspectors observed or conducted

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inspection of the following surveillance activities:

- ObservationofRoutineSurveillanceActivities(Unit 1) c~

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S01-12.9-11 Miscellaneous Surve111ances . 1: a M089050971- Weekly Battery Surveillance Inspectiori - A y

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f - Observation of Routine Surveillance Activities (Unit 2) ,

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S023-3-2.18.1 Atmospheric Dump Valve Operations

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- A S023-11-1,1.3 Plant Protection System 31 Day Surveillanc ,

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E Observation of Routine Surveillance Activities (Unit 3)

s 5023-3-3.34 Turbine Overspeed Protection Weekly Surveillance l2

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,6i ' Monthly Maintenance Activities (62703) ,

x- During this report period, the inspectors observed or conducted ,

i inspection of the following maintenance activities: Observation of Routine' Maintenance Activities (Unit 1)

M089050971 Weekly Battery Surveillance Inspection .]

I M089031639 - Repair of Valve SI-CCW-CV-737A Actuator d

M089060326 Packing Adjustment on SWC Pump G-13A

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b. Observation of Routine Mainten'ance Activities (Unit 2)

M089030060 Ultrasonic Test (UT) Inspection of 3rd Point Heater Piping M089011476 Machine High Pressure Safety Injection (HPSI) Pump MP017 Discharge Head Iw M089040656 Repair Failed High Level Transmitter 2/3LT5656 for l(a , Fire Water. Tank T103 -

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c. Observation of Routine Maintenance Activitids (Unit 3)

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t 4 M089042035 Repair of 011 Sightglass on CCW Pump MP024

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M089041187 Grease'SWC Pump MP307 Motor Bearings b ),

M089041576 Clean & Inspect Condensate Pump MP050 Motor ,.

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y d. Improper Attention to Control of Heavy Lifts In Containment 1 On May 30, motor assembly fasteners on Unit I reactor coolent pump

"B" were damaged and the reactor coolant loop was potentially over-stressed as a result of poor control of rigging operations. The inspector raised a concern that licensee management appeared to have underreacted to this event. In particular, the inspector noted that:

(1) Seven days following the event (when the inspector became aware of the event) root cause assessment of the event was not complete and licensee maintenance supervision and management I were not yet aware of the reasons for the event or of needed corrective actions (programmatic weaknesses, procedure compli-ance, adequacy of supervision or quality oversight, etc).

(2) As of seven days after the event, no briefings on the event had been conducted with site personnel responsible for rigging of heavy loads over safety related system (3) No interim limitations over continued heavy lifts were imple-mented, pending completion of the licensee's evaluation of the event. In fact, additional heavy lifts had been conducted in Unit I containment during the seven day perio The inspector discussed his concerns with the Station Manager, who l agreed and stated that he would expedite resolution of the above concern This item remains open, pending completion of licensee action (206/89-14-03).

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' Inadequate Procedure for Surveillance Testing of Reactor Protective hystem Trip Breakers As part of the review of Unit l' RPS trip breaker testing problems

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noted above, the inspector performed a detailed review of the licensee's surveillance test procedure. Although SCE had identi-fied the above noted problem, involving failure to properly test

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Channel "B" shunt trip features, the licensee's evaluation did not

' address-other deficiencies in the procedure. In particular, the inspector noted that the test procedure (S01-11-2.4.4) was not adequate in that it did not properly test undervoltage (UV) trip functions following removal of a mechanical defeat device, which was installed as part of the test. The licensee agreed and took action to correct the procedure and retest the UV trip functions prior to plant startup. This item is closed-(206/89-14-04).

No violations or deviations were noted during the inspectio . Engineered Safety Feature Walkdown (71710)

Unit 2 The inspector verified the valve alignment of the fire water syste Procedure 502323-7-1, " Fire Suppression Water System Operation" and selected P& ids were use ~No violations or deviations were noted during the inspectio . Plant Modification and Refueling Activities (37700, 37828, 82301)

Examples of Poor Engineering / Technical Work Previous NRC inspection and meeting reports have acknowledged that SCE appeared to have done a good job of diagnosing the root causes of deficiencies in their engineering and technical support organization However, the inspector observed that, during the inspection period, several significant problems involving the same root causes continued to

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recur. During the exit meeting, the inspector noted that a lack of timely implementation of effective corrective actions and inadequate management oversight appear to be primary contributors to these

, continuing problem The following examples were discussed: 4

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(a) Inadequate Engineering Design Work Associated with Modification to the Unit 1 Auxiliary Feedwater System During the recent Unit I refueling outage (Cycle X), the steam generator level instrumentation associated with the auxiliary feedwater (AFW) system was modified. The modification involved

% redesign of existing safety related narrow range level indication to perform as wide range level indication. The new design was

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intended to accomplish the dual function of an automatic initiation

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-Q* , $ '- signal for AFW and redundant indication of AFW flow to the steam

' generators. During' initial Mode.1_ operation', following.the modifi--

N cation, the redesigned instruments malfunctioned, resulting' in spurious initiation of AFW and the need to. return the plant.to Mode 5 to redesign the modification. This; event appeared to involve *

several aspects which indicate. inadequate engineering and technical work and poor: quality oversight:

1) The original design mo'dification effort was. inadequate in _that it did .not_ properly account for steam generator downcomer dynamic' flow effects in the vicinity of the relocated d/p cell instrument taps. -This was a~known effect that should have been considere )- The original design modification: did.not include appropriate nuclear, steam supply system _(NSSS), vendor input to the design,

in that the. vendor subsequently expressed concern about using a wide' range instrument for automatic AFW initiation' .

3) Specified post-modification testing was not adequate to' detec and allow correction of the' design problem' prior to its resulting in a spurious plant transien (b) Inadequate Engineering Involvement in Startup of Modified Auxiliary Feedwater System During plant startup following AFW system modifications (May 25),

the inspector noted that cognizant engineering personnel were not adequately involved in the' evaluation of the performance of new AFW instrumentation. In particular, although the new instruments were observed to indicate improperly following entry into Mode 2 on May 21, cognizant engineers did not become properly involved in assess-ment of the problem until after a spurious AFW actuation occurred on May 25.

4 (c)-InadequateEngineeringInvolvementinStartupofNISystem ,

During the recent Unit I refueling outage (Cycle X), the unit's nuclear instrumentation (NI) system was' replaced. Considerable design effort to correct NI system noise problems was accomplished

.g during this outage. During plant startup from the outage,(May;16),

', + the inspector noted that SCE engineering and station personnel had 3 E not effectively coordinated or implemented a formal course o ,

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,, action for monitoring NI noise during the plant startup, in order 4 j' f to confirm that NI noise reduction efforts hed been' effective. In> ,

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,c i particular, appropriate cognizant engineering personnel'were not ,

a eu present in the control room to properly evaluate NI noise data. As ,s 7 Y 3 '

a result, there was considerable delay and confusion in arriving at I a decision on how to deal with observed NI noise effects.during the,

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'l The licensee acknowledged the inspector's concerns and stated that the t

. need for additional short term actions to improve engineering / technical '

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,i support performance were_being evaluated. This item remains unresolved, !

pending completion of licensee and NRC review (206/89-14-05). i No violations or deviations were noted during the inspectio . Independent' Inspection (35502, 61705) -- Inadequate Safety Evaluation in Support of Continued Plant Operation with Known Design Deficiencies j On May 23, the inspector reviewed SCE plans to return Unit 2 to ,

power operation following the repair of failed steam generator tube j plugs (approximately 700 gpd primary to secondary leak), which ha resulted in a May 13 plant shutdown. The leaking plugs involved 3 of a total of 39 CE welded plugs installed in each of the Unit 2 )

and Unit 3 steam generators. The cause of the leaks was determined i to be an inadequate weld process for welding the plugs during ]

initial installation in 1985. During this review, the inspector !

raised the concern that the licensee had not documented a safety i evaluation 'to substantiate return to operation without repairing .

other plugs with similar potential weld deficiencies. The licensee

.l agreed.that such a safety evaluation was needed and committed to

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. complete the evaluation prior to Unit 2 restar pI During the exit meeting, the. inspector notedLthat this deficiency

'y' was similar to an inadequate Unit 1 safety evaluation, involving *

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improper ~ restart of Unit 1 in July 1988 following the identifica-

l tion of several improperly rolled steam generator tube sleeve On June 9, the inspector reviewed a licensee safety evaluation associated with correction of Unit 2/3 auxiliary fctdwater check l valve noise problems. The inspector noted the following deficien-cies in the licensee's evaluation

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(1) The evaluation did not substantiate the basis for concluding that it was satisfactory to operate Unit 3 until the Cycle V e refueling outage (May 1990), when wear problems with the Unit 3 valves are expected to be worse than those recently cor-rected on Unit 2. In particular, the safety evaluation noted that approximately 50% of the valve disk pin bushings had worn away on the Unit 2 valve, whereas Unit 3 valves continue to demonstrate the same wear mechanism (as indicated by continued valve noise problems) and are planned to be operated in this condition for about twice the time period as the Unit 2 valv (2) The evaluation did not factor in valve vendor input relative to the decision to delay repair of the Unit 3 valve (3) The evaluation did not address the need for any periodic testing to confinn the continued operability of the Unit 3 valves during the interim period that repairs have been delaye As part of the inspector's review, he also noted that SCE is not performing a reverse flow check of the Unit 2/3 AFW check valves as

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<* part of the licensee IST program. Such a test appears-to be war-af ranted, considering the safety function of the valves, as defined in the FSAR. The licensee and NRR are evaluating this concer This item remains open, pending completion of the licensee's evalu-ation(361/89-14-03).

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No violations or deviations were noted during the. inspectio e 10. Review of Licensee Event Reports (LERs) (90712, 92700); <

Through direct observations, discussion with licensee personnel, or review of the records, the following Licensee Event Reports were closed:

Unit 1 89-04 Diesel Generator Load Sequencer Single Failure 89-07 Reactor Coolant Pump Locked Rotor Single Failure 89-09 Delinquent Effluent Monthly Report 89-13 Incorrect Power-0perated Relief Valve (PORV) Setpoint and Improper Inservice Test 89-17 Spurious Reactor Trip Due to NI Noise 89-18 -Technical Specification 3.0.3 Entry to Test Pump Unit 3 89-03 Delinquent Fire Watch 89-04 Technical Specification 3.0.3 Entry to Test ADV 89-06 Reactor Trip Due to Loss of Control Element Drive Mechanism Control System (CEDMCS) Motor-Generator (MG)4 Set The following LERs were not closed based on the reviews conducted, for the reasons indicated:

i Unit 1 89-15 Improper Safety Injection Alignment During Cold Shutdown The report, which involved several violations of Unit 1 safety i injection system alignment requirements, failed to properly define the root cause of the several noted deficiencies or to address warranted corrective actions for those deficiencie The inspector reviewed his concerns with the station

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Operations Manager, who agreed and stated that he would take prompt corrective action, j i

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'1 No violations or deviations were identifie .

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< 1 Follow-Up of Previously Identified Items (92701) <k

' Operation During Reduced Reactor Coolant System (RCS) Invent'ory ,

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Condition (Mid-Loop) Generic Letter 88-17 and TI 2515/101:(Unit 2)

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In order to support the steam generator 2E-088 tube leakage repair ~

during May 13 - June 7, 1989, the Unit 2 RCS was drained to mid-

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loop during May 17 through 24, 1989. The licensee implemented the following short term " expeditious program" to address concerns relative to NRC Generic Letter No. 88-17, " Loss of Decay Heat Removal" before draining the RCS to mid-loop. The licensee committed to these actions in a letter dated January 5,'198 '

(1) Loss of Decay Heat Removal (DHR) Events and Training (GL item 1)

The licensee's Training Division developed and conducted classroom training for all the control room operations personnel and selected people from other departments. The training covered GL 87-12 and GL 80-17, relative to loss of DHR events, and lessons learned from specific events which occurred at Diablo Canyon, SONGS and other sites. The i inspector interviewed selected licensee staff and attended one

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of the training sessions and found it to be satisfactor (2) Containment Closure (GL item 2)

Before entering a reduced RCS inventory condition, the licen-see implemented procedures which would achieve containment closure before core uncovery in the event of a loss of DH The necessary administrative controls and action requirements were included in the following procedures:

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5023-3-1.8 (TCN 6-1) Draining the RCS

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S023-13-15(TCN1-5) Loss of Shutdown cooling These procedures defined containment closure responsibilities and required the use of quick disconnect fittings on cables and hoses going through the personnel, escape and equipment hatches. Maintenance walked down the system and certified that the containment openings could be closed within two hours upon operator notification. These procedures also required strict accountability of work orders that could affect con-tainment integrity. The inspector reviewed the list, walked down the containment, interviewed selected individuals, and ,

found the controls to have been effectively implemente l l

3) Reactor Vessel Temperature (GL item 3)

In order to monitor the core exit temperature data at mid ,

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loop, the licensee established controls to' maintain at least i two independent core exit thermocouple (TC) d'hannels in service continuously.

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" The-thermocouple channels are part of the' Qualified Safety.

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4 m_ Parameter Display System's-(QSPDS) heated junction thermocoup-4 les'(HJTC). The unheated TCs of the HJTC probes of different

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qualified core exit TCs are also available as: backup. Proce--

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dure S023-3-1.8 specifies that-the reactor core exit tempera-L ' ture be: recorded hourly while;the,RCS level is.below the top L,r of the hot leg. The frequency is. increased to every 15

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' avail abl e.- (The licensee.was planning to' add automatic high-

.f . temperature alarm capability during a future outage of

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3 sufficient duration.)

The inspector periodically checked the temperature readouts in the control room and reviewed the records and found them'to.be

/ . satisfactor ). RCS Water Level (GL. item 4)

In order to obtain reliable RCS water level data' during reduced inventory conditions, theflicensee established the following-measures: -

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Refueling Water Level Indication (RWLI) in the contro room and the corresponding local sight glass inside the

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containmen HJTClocatedabovetheactivefuelprovidestl1e' heat removal data of its surround medium (water or steam).

-This information is displayed on the QSPD .

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RCS Draindown Calculation Record tracks ~the total RCS

. inventory balance. It is update.d whenever water is added

, A; ' to or removed from the RCS. The various elevations of the'RCS have been determined for expected water inve.ntory

Y by analysis and thus can be used by the operators to jh track the water leve $ '

These in' dependent level measurements are correlated periodi-cally and at.various hold points during'the.draindown evolu-

, tion. The inspector periodically checked these measurements:

and the. operator's attention to them and found.them to b'e- 1 '

satisfactor '

5) RCS Perturbations (GL item 5)

The licensee established measures to control RCS perturbations l during reduced inventory operation. Procedure S023-3- !

identified the types of activities that'may become.a threat to RCS-inventory, RCS stability, system control and instruments-tion. These activities were to be delayed, if possibl Otherwise, management approval would be required for such activities and they were to be tracked by Attachment 20 RCS

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u- 13

y .

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" Perturbation List, to the procedure. The inspector-reviewed

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the procedure and found it to be satisfactor ,

6)- RCS Inventory Addition (GL item 6)

In order to provide two available means.of adding water.to the RCS upon a loss of DHR event, the licensee's' procedure A S023-3-1.8 establish controls to specify that:

+ .. *

g ,

"One HPSI pump shall_be operable and another. pump >> ,

f<

, availabl '

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The available pump may be either a HPSI p' ump or. a

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',

Containment Spray Pum % The Operable HPSI Pump shall be capable of injecting into 34 ~

two cold legs and one hot leg with associated. injection -

valves operable from the control room."

' *

> <

,L The inspector reviewed the daily equipment status and found

~

i them to be satisfactor <

7) Nozzle Dam Usage (GL item 7) ~

In order to prevent rapid RCS inventory loss due to pressurization caused by boiling in conjunction with hot leg blockage, the licensee established controls in procedure

'S023-3-1.8. This is accomplished by a specified operational sequence to open the pressurizer manway and hot leg manway before installing the nozzle dams: The removal sequence is the reversed order. The inspector walked down the pressurizer-and the hot leg, reviewed the licensee's daily equipment status, and found them to be satisfactor This inspection and the inspection do:umented in report 206/88-28 for Unit I closed NRC Inspection Manual Temporary Instruction 2515/101 for all three units, (Closed) Open Item (362/87-10-02), Technical Specification for ADVs As described in report 362/89-06, paragraph 3a, the licensee inclu-ded the atmospheric dump valves (ADVs) in the periodic surveillance program. The licensee was preparing a proposal to change the Technical Specifications to include the ADVs in the surveillance and operability requirements. This item is closed, (0 pen) Unresolved Item (361/88-25-01), Calorimetric Uncertainties The licensee presently limits recctor power to within the uncer-tainty level during each plant restart until the power level is confirmed by secondary calculations. The licensee is continuing the collection and evaluation of plant data in this regard. The licensee also plans to issue a revised LER on this problem. This item remains open pending completion of these action l

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'L . '

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d.- (C10 sed) Open Item (361/87-20-01), Steam Driven Auxiliary Feedwater

• Pump 2P-140 Speed Control Deficiency Tne licensee replaced the EG-M and ramp generator modules and adjusted the speed sensor. The maintenance procedure was revised to adjust the controller to a tighter tolerance ~during each surveillance. The speed 'cc atrollers for both units have been functioning properly since.th2 replacement.- lThis item is close ,

12. ExitMeeting(30703) -

On June 16, 1989 an exit meeting was conducted with the licensee y representatives identified in Paragraph 1. The inspectors summarized '

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the inspection scope and findings as described in the Results section of-4'

this repor , The licensee acknowledged the inspection findings and noted that appropriate corrective actions would be implemented where warrante The licensee did not identify as proprietary any of the'information

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provided to or reviewed by the inspectors during this inspectio .

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