IR 05000344/1987037

Insp Rept 50-344/87-37 on 870906-1017.No Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Maint,Surveillance,Low Power Physics Testing & Event Followup

ML20236S813
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	11/06/1987
From:	Rebecca Barr, Mendonca M, Suh G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20236S806	List: IR 05000344/1987037 ML20236S804
References
50-344-87-37, NUDOCS 8711300109
Download: ML20236S813 (10)

Text

,

..

..

U.S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No.: 50-344/87-37 Docket No.: 50-344 License No.: NPF-1 Licensee: Portland General Electric Company 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name: Trojan Inspection at: Rainier, Oregon Inspection conducted: September 6 - October 17, 1987 Inspectors: %%7= _

./m <//9/P7 R. C. Barr v Date Signed Senior Resident Inspector

% h N/e /E D G. Y. Suh / Date Signed Resident Inspector j Approved By: %

~

1. M &7 M. M. Mendonca, Chief Date Signed Reactor Projects Section 1

Summary:

Inspection on September 6 - October 17, 1987 (Report 50-344/87-37)

Areas Inspected: Routine inspection of operational safety verification, ,

maintenance, surveillance, low power physics testing, and event follow u Inspection procedures 30703, 61700, 61710, 61726, 62703, 71707, 71709, 71710, 71881, 72700, 73051, 82301, 92700, 92701, and 93702 were used as guidance during the conduct of the inspectio Results:

No violations or deviations were identified.

8711300109 871109 PDR ADOCK 05000344 G PDR I

i l - - - - - - - - - - - - - - - - - - - - - - -

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

.

DETAILS Persons Contacted D. W. Cockfield, Vice President, Nuclear C. A. Olmstead, Plant General Manager

• R. P. Schmitt, Manager, Operations and Maintenance

• D. W. Swan, Manager, Technical Services J. K. Aldersebaes, Manager, Plant Modifications J. D. Reid, Manager, Plant Services A. N. Roller, Manager of Nuclear Plant Engineering R. L. Russell, Operations Supervisor R. H.'Budzeck, Assistant Operations Supervisor D. L. Bennett, Maintenance Supervisor R. A. Reinart, Instrument and Control Supervisor T. O. Meek, Radiation Protection Supervisor R. W. Ritschard, Security Supervisor C. H. Brown, Operations Branch Manager, Quality Assurance D. L. Nordstrom, Nuclear Engineer, Nuclear Safety and Regulation The inspectors also interviewed and talked with other licensee employees during the course of the inspection. These included shift supervisors, reactor and auxiliary operators, maintenance personnel, plant technicians and engineers, and quality assurance personne * Denotes those attending the exit intervie . Plant Status On September 5, 1987 the Trojan facility was at 61% reactor power, limited by main condenser backpressure since onlj one of two of the circulating water pumps was operable. On September 7, 1987 the North Main Feedpump (MFP) was shutdown due to excessive vibration and was later determined to have damaged 5th and 8th stage blading. The reactor was shutdown on September 28, 1987, due to exceeding the design base limits for the Emergency Core Cooling System (ECCS) external leakage when the Boron Injection Tank (BIT) outlet relief valve (PSV-8852) was found leaking in excess of 1560 cc/hr. On October 2, 1987, following the repair of the BIT relief valve the reactor was returned to 65% power. On October 8, 1987 the North MFP was returned to service and power increased to 75%. On October 14, 1987 the annual ' Radiological Emergency Preparedness Exercise' was conducted. On October 17, 1987 the facility was at 91% power with the repair of the circulating water pump and 100%

power operation projected for October 25, 198 . Operational Safety Verification During this inspection period, the inspectors observed and examined activities to verify the operational safety of the licensee's facilit The observations and examinations of those activities were conducted on a daily, weekly or biweekly basis.

- _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ _ __ _ _

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

.

~

On a daily basis, the inspectors observed control room activities to verify the licensee's adherence'to limiting conditions for operation as !

prescribed in the facility Technical Specification Logs, .

'

instrumentation, recorder traces, and other operational records were examined to obtain information on plant conditions, trends, and~

compliance with regulations. On occasions when a shift turnover was in progress, the turnover of information on plant status was observed to determine that all pertinent information was relayed to the oncoming shift personne Each week the inspectors toured the accessible areas of the facility to observe the following items: General plant.and equipment condition Maintenance requests and repair Fire hazards and fire fighting equipmen Ignition sources and flammable material control, Conduct of activities in accordance with the licensee's administrative controls and approved procedure Interiors of electrical and control panel Implementation of the licensee's physical security pla Radiation protection controls, Plant housekeeping and cleanlines Radioactive waste system Proper storage of compressed gas bottle The inspectors examined the licensee's equipment clearance control weekly with respect to removal of equipment from service to determine that the licensee complied with technical specification limiting conditions for operation. Active clearances were spot-checked to ensure j that their issuance was consistent with plant status and maintenance

'

' evolutions. Logs of jumpers, bypasses, caution and test tags were examined by the inspector Each week the inspectors conversed with operators in the control room, and with other plant personnel. The discussions centered on pertinent topics relating to general plant conditions, procedures, security, training and other topics aligned with the work activities involve The inspectors examined the licensee's nonconformance reports (NCR) to confirm that deficiencies were identified and tracked by the syste Identified nonconformances were being tracked and folloe d to the completion of corrective actio Routine inspections of the licensee's physical security program were performed in the areas of access control, organization and staffing, and detection and assessment system The inspectors observed the access control measures used at the entrance to the protected area, verified the integrity of portions of the protected area barrier and vital area barriers, and observed in several instances the implementation of compensatory measures upon breach of vital area barrier Portions of the isolation zone were verified to be free of obstructions, and the functioning of the central and secondary alarm stations, including the use of CCTV monitors, was observe On a sampling basis, the inspectors

.. . .

_ _ _ _ _ - _ _ _ _ _ _ _ _ __ _ _ _

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

'

' *

} !

verified that the required minimum number of armed guards and an individual authorized to direct security activities were on site, i The inspectors conducted routine inspections of selected activities of the licensee's radiological protection program. A sampling of radiation j work permits (RWP) were reviewed for completeness and adequacy of J information. The use of two RWF's relating to work performed inside the bioshield in containment were specifically reviewed. The R..P's were readily available for the workers' review, and the inspectors confirmed that the specified controls and requirements of the RWP's were implemented at the work site. During the course of other inspection activities and periodic tours of plant areas, the inspectors verified ,

proper use of personnel monitoring equipment, observed individusis leaving the radiation controlled area and signing out on appropriate RWP's, and observed the posting of radiation areas and contaminated areas. Posted radiation levels at several locations within the fuel and auxiliary buildings were verified by the inspectors using both NRC and licensee portable survey meters. The involvement of health physics supervisors and engineers and their awareness of significant plant activities was assessed through conversations and review of RWP sign-in record The inspectors verified the operability of selected engineered safety feature This was done by direct visual verification of the correct position of valves, availability of power, cooling water supply, system integrity and general condition of equipment, as applicable. Systems verified operable during this inspection period included the Safety Injection System and the Auxiliary Feedwater Syste No violations or deviations were identifie . Maintenance The inspectors observed portions of the work to repair a leaking steam generator drain line manual valve. The valve, SG-008, located on a one inch line upstream of the steam generator blowdown tank had leakage at the body to bonnet flange. The temporary repair required fabrication of a clamp to fit around the flange and injection of sealing compound between the clamp and flange. On the initial attempt, the sealant (furmanite) was injected between the clamp and flange. Upon increasing system temperature and pressure, leakage was detected at the flange bolt holes.

,

'

The subsequent repair, observed by the inspectors, was to drill through the clamp and into the body to bonnet flange to the outer radius of the flange gasket and seal with sealing compound. The injection of the sealing compound was performed by a vendor representative. The inspectors reviewed the drawing for the internal geometry of the valve, and examined a replacement valv The inspectors concluded that the small amount of sealing compound injected into the flange gasket area would not be expected to migrate into the flow line or interfere with valve motio The inspectors reviewed the applicable Maintenance Request 87-6023 and Nonconformance Report 87-38 A quality control check for leakage at normal operating temperature and c. essure was specified on MR

____-___-____-______-___-_-__-____A

_ - _____ - __ .

- - - .

. ..

'

.

'

l l 87-6023 and subsequently performed as observed by the inspector. As i stated on NCR 87-387, the repair with sealing compound was temporary and

'

the valve will be replaced during the next outage. The inspectors noted that radiological controls were appropriate, and the requirements of Radiation Work Permit 87-89 were properly implemente q i

No violations or deviations were identifie j l S. Surveillance The inspectors conducted a review of the licensee's surveillance procedure used in the heat balance calibration of the power range nuclear instrumentation channels. Periodic Operating Test POT-22-1, " Nuclear Instrumentation System Heat Balance Calibration", was verified to be in conformance with technical specification requirements. The inspectors reviewed the equations used in P0T-22-1 and the analyses in calculation TNP-85-09 Revision 0 and Revision 1 which presented the bases for various factors and equations used in P0T-22-1. TNP-85-09 results were found to be consistent with POT-22-1. The inspectors determined that TNP-85-09 used conservative assumptions and verified on a selected basis the accuracy of the calculation The inspectors reviewed a sampling of previously completed POT-22-1 data sheets to confirm that surveillance test documentation was reviewed and that test discrepancies were addressed and resolved. In addition, the inspectors observed the performance of POT-22-1, reviewed the resulting test data for accuracy and completeness, and performed an independent calculation to verify the licensee's results. The calibration of feedwater and steam generator flow meters used in the procedure was verified to be curren No significant discrepancies were identified in the review of POT-22- In addition, a review of P0T-29-1, " Core Quadrant Power Tilt Ratio Verification," was performed to verify that technical specification j requirements were being me The inspectors observed the conduct of the '

surveillance test, reviewed the test data for accuracy and completeness, and performed an independent calculation of the test results dat Test personnel referred to the test procedure during the conduct of the test and followed the instructions set forth in POT-28- In %e review of POT-28-1, the inspectors identified several apparent wenknesses in the surveillance procedure and discussed them with licensee r' representatives:

The inspectors' review did not identify positive controls which assured that the quadrant power tilt ratio (QPTR) was determined to be within the limit within the required time period prior to exceeding 50% of rated thermal power. This was required by Technical Specification (T.S.) 4.0.4 as applied to T.S. 4.2.4 for QPT POT-28-1 was only applicable for measurements of QPTR at power levels above 50% of rated thermal power and thus did not address the T.S. 4.0.4 requirement to perform the surveillance requirement within the stated surveillance interval prior to entering the specified applicability condition. Below 50% power, QPTR was measured using the movable incore detection syste POT-28-1 specified that before a determination could be made that

_ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ -

______ _ _ _ _ __-

.

.

the QPTR exceeded T.S. limits, the QPTR must be measured in different ways, including operator hand calculations which used

' power range detector voltages, computer calculations which also used power range detector voltages, hand calculations which used power range detector currents, and a flux mapping procedure which used the incore movable detectors. The surveillance procedure did not specify a time frame for making the various calculations and making a determination that QPTR exceeded its limi Another procedural weakness in POT-28-1 was the procedural step which specified the actions to be taken once-QPTR was determined to exceed the limit. The procedural step was incomplete and in some respects inconsistent with the T.S. required actions. However the inspectors would not expect this to be a problem, since operations personnel know T.S. requirements and take appropriate action Plant management has committed to correct this proble The inspectors reviewed POT-28-1 data sheets completed to date for Cycle 1 In three instances, the steps specified in POT-28-1 for making a QPTR determination were apparently not followed per procedure, Hand and computer calculations using power range detector voltages indicated, or ,

should have indicated, an excessive QPTR value; however, calculations '

using detector currents were apparently not performed contrary to i POT-28-1 requirements. (The nuclear engineer indicated that calculations for this function may have been done). Instead, QPTR was concluded to be within limits based on discussions with the cognizant engineer. (QPTR was measured at various power levels, including 50%, 75%, and 85%, using the incore movable detector system and determined to be well within the T.S. limi These measurements, however, were not performed as explicit parts of the POT-28-1 surveillance tests.) No procedural deviations to POT-28-1 were apparently processe In the review of other recent surveillance tests, approximately eight hours passed between hand and computer calculations which indicated an excessive -QPTR value and a calculation based on detector currents which confirmed that the QPTR was within limits. This compared with a two hour action statement in T.S. 3.2.4 once the QPTR is determined to exceed the limi The inspectors will follow the above concerns as Unresolved Item 50-344/87-37-01 in accordance with the licensee's procedure pending verification of calculation performanc i ECCS External Leakage Greater Than Design Allowable On September 27, 1987 while at 65% power, an operator on a routine ,

facility tour noted the baron injection tank (8IT) outlet relief valve I (PSV-8852) to be leaking at approximately 13,560 cc/ min (3.5 gal /hr). On September 28,1987 at 2:40 p.m. , the licensee recognized the 1580 cc/hr '.

design basis leak rate for ECCS external leakage was being exceeded. The Resident NRC Inspector was notified at 3:15 p.m. The report also noted the discharge of PSV-8852 was directed to the auxiliary building drain system vice the pressurizer relief tank as stated in the Final Safety Analysis Report (FSAR). At approximately 8:00 p.m. after discussions with i

__ .

.. ______

-__-_ __

.

'

the NRC Regional and Headquarters Management, the licensee shutdown the reactor upon recognizing that, based on existing accident analyses, ECCS external leakage greater than 1580 cc/hr would result in both trains of the control room ventilation being unable to perform their design function if, in the highly unlikely event, a design base loss of coolant accident were to occu Areas of concern were:

(a) management recognition and response to the safety significance of ECCS external leakage (based on the leakage having existed for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following the leak identification);

(b) implementation of NUREG-0737 III.D.1.1. (based on plant walkdowns that indicated other ECCS external system leakage existed but was not documented or tracked against the 1580 cc/hr criteria; and, (c) the disparity between the FSAR and the as-found design (based on the PSV-8852 relief discharge line routing to the auxiliary building drain system vice the presurizer relief f<nk).

A review of this event is being conducted by an independent licensee group. The review is expected to address the above concerns. As an interim measure, the licensee has briefed all operation and maintenance personnel on the significance of ECCS external leakage and the importance of identifying and repairing any ECCS external leakage. Additionally, the licensee is processing a design change to improve the recirculation filtration on the Emergency Control Room Ventilation System. This issue will be followed as Open Item 50-344/87-37-0 . Review of Physics Testing Results As discussed in Inspection Report 50-344/87-30, the licensee conducted low power physics tests per Periodic Engineering Test PET-13-1 and was in the process of reviewing the test results during the previous inspection perio The results of the inspectors' review of the low power physics test data are described below. Additional physics testing specified in PET-13-2, " Reload Cycle 10 No Load and at Power Tests" have not been completed and reviewed to date, since the plant has not achieved 100% of rated power due to equipment outages in the main feedwater and circulating water systems (see paragraph 2 for discussion). The inspectors will review PET-13-2 test data at a later dat The inspectors reviewed low power physics test (PET-13-1) results which included isothermal and moderator temperature coefficient determinations, control rod worth measurements, and zero power critical boron concentration determinations for various rod configurations. The ,

measured values were verified to meet applicable Technical Specification 1 limits and were fcund to be in good agreement with the estimated design values presented in the nuclear design report prepared by the fuel vendc In a review of technical specification (T.S.) surveillance requirements, j the inspectors identified and discussed with licensee representatives

!

.

_

_ . _ - ,

.

? r, 7 apparent. discrepancies with regard to surveillance requirement .10.1.2. This T.S. required that the rod drop time of each full length rod not fully inserted be verified to be less than or equal to L seconds within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the shutdown margin to less I than 1600 pcm. From discussions with the licensee engineers, the inspectors understood that this surveillance requirement was met by step 8. 6. 7 of PET-13- For Cycles 9 and 10 (the current cycle), step 8.6.7 called for manually tripping the reactor and measuring with'a stopwatch the time for the digital rod position indicating (DRPI) status lights of the withdrawn rods to reach the dashpot region. Step 8.6.7 specified that this measurement be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to diluting control bank C below 100 steps. The withdrawn rods were not taken to their fully withdrawn position prior te the manual reactor trip. The drop times were measured by three individuals each with stop watches looking at the drop time of the oost reactive rod, and the drop time of record was the slowest of the three. Prior to Cycle 9, step 8.6.7 called for a manual trip of the reactor, full withdrawal of the most reactive red to stop 228, followed by removal of the movable and stationary gripper fuses for  ;

the most reactive rod and measurement of the rod drop time through use of t a stopwatch and DRPI status light The inspectors' concerns were as follows:

(a) for cycles prior to Cycle 9, the manual reactor trip and measurement of the drop time of the most reactive rod may not have fully met the j requirements of T.S. 4.10.1.2 which called for the measurement of

{

rod drop times of each rod not fully inserted; (b) rod drop times have not been measured from the fully withdrawn i position, and stopwatches and CRPI status lights were used instead of the instrumentation specified in Periodic Instrument and Control Test PICT-16-1, " Hot Rod Drop Time Measurements";

(c) for Cycles 9 and 10, the rod drop time of the most reactive rod, vice the slowest rod, was measured and recorded; the licensee's rod insertion limit for control bank "C" at zero power is 118 steps, vice the 100 steps specified in step 8.6.7. (This apparent discrepancy was not noted in the review and approval process for PET-13-1, which was revised on an annual basis for recent relc;d cycles.)

The inspectors will follow these concerns as Unresolved Item 30-344/87-37-0 . Subcooled Marain Monitor (SMM) Operability At 9:15 a.m. on October 10, 1987 Technical Specification (T.S) 3.3.3.9.,

Accident Monitoring Instrumentation, was invoked when both channels of the SMM were declared inoperable. On October 12,1987 at 1:30 p.m. and 2:20 p.m. SMH Channels A and B, respectively, were declared operabl After evaluating the $MM and other temperature monitoring systems, the licer.see concluded the SMM had not actually been inoperable but rather

>

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

- - - _ _ _ _ _ _ __

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

. _ _ _ _ _ _ - -

c.

another temperature monitoring system (the Fluke data logger) was inoperable due to an electronic. failur Both these systems were modified or added during the 1987 refueling outag Based on log review, discussion with operators, maintenance personnel and engineers, the inspectors concluded confusion on SMM operability occurred ,

because training on newly installed equipment was not detailed enough to I provide required knowledge on the newly added/ modified systems and the acceptance criteria of PET 7-5 was insufficient to enable users to correctly determine equipment operabilit i Although technical specification action statement requirements were satisfied, the question of SMM operability lasted for over 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> apparently due t i a) no direct plant management involvement with the operability concern; and b) the System Engineer not being involved with problem resolutio In the October 22, 1987 exit the licensee committed to further examination of this even This issue is identified as Open Item 50-344/87-37-0 !

9. Off Normal Event Ev_aluation j

<

On August 26, 1987 during a reactor startup, a turbine trip and feedwater isolation occurred at 9:10 p.m. due to hi-hi steam generator water leve On August 28, 1987 while placing a second main feedwater pump in  !

service, a reactor trip and feedwater isolation occurred at 8:31 a.m. due to hi-hi steam generator water level. Event Report 87-139 identified the cause of the turbine trip to be the result of incorrect setting of a feedwater regulating valve manual positioner. Event Report 87-140 identified the cause of the reactor trip to be the incorrect positioning of the main feedwater pump turbine speed controller by the Control Operator. Additionally, Event Report 87-140 identified sluggish operation of the "C" feedwater regulating valve as a contributing caus !

Because the two events were similar, the inspectors evaluated the event for common mode failure. The inspectors found no common mode failur However, they did conclude that the "C" feedwater regulating valve had ,

been identified to hunt in the automatic mode at 24% reactor power prior I to the turbine trip, Maintenance was performed on the valve via MR 87-5449 before the reactor scram of August 28, 1987. A functional retest j was not performed to evaluate the effectiveness of the repair, j Consequently, the sluggish operation of that valve contributed to the j subsequent reactor scra The inspectors concluded the effect of the sluggish operation of the "C" feedwater regulating valve was not identified in Event Report 87-139 because the event evaluation was not of sufficient detail to identify the sluggish valve operation as a contributor to the turbine tri Event

)

Report 87-140 included assessment of chart recorders and trended data to j l

- - - - -

.

n .

ascertain the sluggish valve operation was a contributing facto Event Report 87-139 did no The thoroughness and timaliness of evaluation of facility off normal events has been a subject of previous inspector concern and discussion with plant management. Several categories of event evaluation exis The category most frequently used is routine. A routine evaluation results in the organization responsible for causing the event as the organization evaluating the cause of the even Additionally, this organization normally has the greatest workload in returning the facility to power operatio As a result, the evaluations are not always timely and thoroug To improve evaluation thoroughness and timeliness plant management has committed to more frequent use of independent organizations to evaluate ;

future safety related off normal event The inspectors will closely l monitor the effectiveness of future event evaluation . Unresolved Item An unresolved item is a matter about which more information is required to ascertain whether it is an acceptable item, a deviation, or a violatio Unresolved items are documented in paragraphs 5 and . Exit Interview i

The inspectors met with the licensee representatives denoted in paragraph 1 on October 22, 1987, and summarized the scope and findings of the inspection activitie I i

l l

1