IR 05000312/1988040

Insp Rept 50-312/88-40 on 881208-890113.No Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Health Physics & Security Observations,Esf Sys Walkdown,Maint,Qa & Surveillance & Testing

ML20235N042
Person / Time
Site:	Rancho Seco
Issue date:	02/06/1989
From:	Dangelo A, Miller L, Myers C, Qualls P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20235N041	List: IR 05000312/1988040 ML20235N037
References
50-312-88-40, CAL, NUDOCS 8902280550
Download: ML20235N042 (11)

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U.:-S. NUCLEAR REGULATORY. COMMISSION v

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+ Report No:, 50-312/88-40 l Docket:No. 50-312

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' License No. DPR-54 *

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' Licensee: . Sacramento Municipal Utility. District P. 0. 80x 15830 Sacramento, California 95813

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Facility'Name:. Rancho Seco Unit l'

o n -y 0 Y ?I nspection'at': Herald, California (Rancho Seco Site)

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' (.'LIhspection conducted: . December 8,.1988 through January 13, 1989.,

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IInspectorsi j ~, 3

.bD' Angelo,1 A

b5Sehior[wResident Inspector 2 Q - s9 Date Signed a ~

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N C. J. Myers, Resident Inspector Date Signed Q* -

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(t~ i P. M. , Qualls,' Resident Inspecto . Date Signe"d

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Approvedly: - ,4 # WO

' C~T. . Miller, Chief < Date Signed

5 Reactor ProjectsSection II Summarg: ,

Inspection -betkeen December 8,1988 and January 13,1989(Report 50-312/88-40)

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Areas' Inspected: This routine inspection by the Resident Inspectors involved the areas of operational safety verification, health physics and securit observations, engineered safety features system walkdown, maintenance

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surveillance and testing, quality assurance and followup items. During this inspection, Inspection Procedures 71707, 71710, 71714, 61726, 72700, 62703,

. 93702, 92701,.62702, 61706, 73051 and 30703 were use 'Results: An observed weakness during this period was the operation-'of auxiliary steam system valve ASC-009 as a compensatory measure without procedures to control the activity, and Operations management decision to operate for approximately. three hours with'the Auxiliary Steam System in a degraded stat No violations or deviations were note >

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p DETAILS

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' 1.- Persons Contacted

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Licensee' Personnel i e

J. F. Firlit, Chief Er.ecutive Officer, Nuclear

• D. Keuter, Asuistant General Manager (AGM), Nuclear Power

. Production

• J. Shetler, AGM, Plant Support Services

• B. Croley, AGM, Technical Services

• J. Vinquist, Director, Nuclear Quality

• D. Brock, Manager, Nuclear Maintenance W. Kemper, Manager, Nuclear Operations T. Redican, Manager, Materials Macagement

• S. Crunk, Manager, Nuclear Licensing

• R. Baim, Manager, Nuclear Cost Control and Plant Services P. Turner, Manager, Plant Performance

• M. Bua, Manager, Radiation Protection L. Fossum, Manager, Scheduling and Outage Management D. Ferguson, Manager, Nuclear Engineering L. Houghtby, Manager, Nuclear Security J. Delezenski, Supervisor, Incident Investigation Review Group (IIRG), Licensing G. Coleman, Quality Engineering Supervisor

• G. Legner, Licensing Engineer J. Robertson, Licensing Engineer S. Carmichael, Maintenance Engineer B. Wilson, ICS System Engineer Other licensee employees contacted included technicians, operators, mechanics, security, and office personne * Attended the Exit Meeting on January 13, 198 . Operational Status of Rancho Seco The plant started this inspection period operating at 92% reactor powe On December 9, 1988 the plant tripped due to main feedwater pump control problems resulting in high Reactor Coolant System (RCS) pressure. The plant restarted on December 12, 1988. When reactor power was about 12%,

problems in the Auxiliary Steam System resulted in the operators manually tripping the reactor. Due to failure to secure auxiliary steam to the 5th point feedwater heater, the plant cooled 13*F below its normal post reactor trip temperature. The "B" steam generator was allowed to empty as required by the emergency procedure. This event is discussed in Paragraph 7. The plant was shut down to conduct corrective actions until January 6, 1989. On January 6, the plant was restarted and brought to about 10% power when vibration induced problems with the Main Steam / Auxiliary Steam pressure reducing valve caused the operators to shut down the plant. 'The plant made modifications necessary to correct

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the vibration problems and restarted on January 12, 1989. At the

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conclusion of the inspection period, the licensee had the plant at about 33% reactor power with the intent to slovly increase power back to the 92% power platea . Operational Safety Verification (71707, 71714)

The inspectors reviewed control room operations which included access control, staffing, observation of system alignments, procedural adherence, and log keeping. Discussions with the shift supervisors and operators indicated an understanding by these personnel of the reasons for annunciator indications, abnormal plant conditions and maintenance work in progress. The inspectors also verified, by observation of valve and switch position indications, that emergency systems were properly aligned as required by technical specifications for plant condition The inspectors observed the licensee's implementation of cold weather ;

precautions during periods of ambient temperature below 34 F. Operations '

actions prescribed in licensee procedure C.33, Freeze Protection, included operation of various pumps to establish circulation in outdoor piping systems and energizing the outside heat tracing circuitry. These actions appe* red adequat Reactor startup on January 6,1989 was observed 5y the inspecto The inspector also verified the licensee's Estirated Critical Position (ECP) calculation used during the startup. The operators were knowledgeable about the procedure and ECP. They showed good procedural adherence. Criticality was observed and declared at the proper time and well within the ECP limits. The licensee's nuclear engineer and senior management were available in the control room during startup to provide assistance, if neede Steam plant startup was observed on January 12, 1989. The operators adhered to plant operating procedures. Operator attentiveness and formality were maintained. Steam plant startup was conducted without inciden Tours of the auxiliary, reactor, and turbine buildings, including exterior areas, were made to assess equipment conditions and plant conditions. Also, the tours were made to assess the effectiveness of radiological controls and adherence to regulatory requirements. The inspectors also observed plant housekeeping and cleanliness, looked for potential fire and safety hazards, and observed security and safeguards practice >

During this inspection period, no significant housekeeping discrepancies were identified by the inspector. This was an improvement from the previous perio During work activities, it appeared that the health physics managers were conducting plant tours and monitoring work in progress. They appeared aware of significant work which occurred during this perio The inspector's Radiation Work Permit (RWP) review revealed that the RWPs !

did include: job description, radiation levels, contamination, airborne

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V radioactivity.(if expected), respiratory equipment, protective clothing, dosimetry, special equipment, RWP expiration, health physics (HP)-

coverage, and signatures.. The RWP radiation and contamination surveys

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The inspectors observed that personnel in the controlled areas were wearing the proper dosimetry and personnel exiting the controlled areas E', were using the monitors properly. Labeling of containers appeared'

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(The.jn5pNetors= walked 'down portions of the protected and' vital area

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. boundaries 'to_ ensure that they were intact and that security personnel-

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were properly posted where known deficiencies existed. The inspectors

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, also observed protected area access control, personnel screening, badge j

, issuing and maintenance on access control equipment. Access control was j

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observe Personnel entering with packages were properly searched and {

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access control was in accordance with licensee procedures. The- i inspectors' observed no obstructions in the isolation zone which could

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- conceal:a person or~ interfere with the detection / assessment syste Protected area illumination appeared adequat No violations or deviations were identifie . ESF System Walkdown (71710)

- During the inspection period the inspectors walked down the High Pressure Injection (HPI) System outside the Reactor Buildin The inspectors concluded that:

All observed hangers and supports were properly made up and aligne Housekeeping was adequat )

No excessive packing leakage was observed on valve '

Major system components were properly labeled, lubricated and cooled. No excessive leakage was apparen Instrumentation appeared to be properly installe No out of. calibration. gauges were identifie Flow path components appeared to be in the correct positio Required support systems were availabl Proper breaker and switch positions were verifie ;

No violations or deviations were identified.

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. . Monthly' Surveillance Observation (61726)

Technical Specification (TS) required surveillance tests were observed and reviewed to ascertain that they were conducted in accordance with Technical Specification requirentent The following surveillance activities were observed:

- SP.56B Variable Diesel Generator (G-886B) Synchronization Surveillance Test

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SP.20 Monthly Turbine Driven Auxiliary feedwater Pump (P-318) Inservice Test

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SP.1198 Manual Operation of the Auxiliary Steam Reducing Valve, PV-36014A The following items were considered during this review: testing was in accordance with adequate procedures; test instrumentation was calibrated; limiting conditions for operation were met; removal and restoration of i the affected components were accomplished; test results conformed with TS '

and procedure requirements and were reviewed by personnel other than the individual directing the test; the reactor operator, technician er engineer performing the test recorded the data and the data was in agreement with observations made by the inspector, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne No violations or deviations were identifie . Monthly Maintenance Observation (62703) Maintenance Activities Maintenance activities for the systems and components listed below were observed and reviewed to ascertain that they were conducted in

- accordance with approved procedures, regulatory guides, industry codes or standards, and the Technical Specification .

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Main feedwater pump maintenance on both pumps which included the replacement of the Lovejoy control modules for the feedwater pump control syste Routine maintenance on the "A" HPI pump, which included verification of oil level in the pump speed controller, I nionitoring vibration levels, seal temperatures, and air filter

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Troubleshooting of the Main Steam / Auxiliary Steam reducing )

, valve, PV-36014A, following the valve's failure on December 12, i 1988. The troubleshooting had identified a loose adjustment cap within the pneumatic controller of the valve. In addition,

maintenance personnel had also identified a cleaning and lubrication task which was specified by the manufacturer of the l

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t pneumatic controller but had not been specified in the licensee's preventive maintenance program. The licensee began a review of other pneumatic controller, preventive maintenance schedules, with _the intention to add the previously omitted task if appropriat Modifications to the Main Steam / Auxiliary Steam reducing valve (

made as a. result of the December 12, 1988 valve failure. These I modifications included the replacement and relocation of the ' !

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pneumatic control system for the PV-36014A valve and also additional pipe / valve supports to ieduce valve vibratio The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to

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initiating _th'e work; activities were accomplished using approved

,_ procedures'and were inspected as applicable; functional testing or ;

calibration was performed prior to returning components or systems to service; activities were accomplished by qualified. personnel; radiological controls were implemented; and fire prevention controls

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were' implemente Temporary Modific'ations The insoector. reviewed the licensee's control of a temporary installation on one of-the main steam safety valves (MSSV),

PSV-20544. The licensee had identified that the valve had previously lifted below its setpoint during operational transient To preclude future inadvertent lifts, a gag was installed until the valve could be replaced during the next scheduled cold shutdown outag ' The inspector found that the licensee directed the installation of the gag in the disposition of PDQ-88-1459 (Potential Deviation from Quality), Although expected to exist for only a relatively short time, the alteration of the plant configuration was not controlled as a temporary modification, but rather as a permanent plant change involving a 10 CFR 50.59 review and a change to the plant drawing The inspector reviewed the 50.59 evaluation for the modificatio The 50.59 evaluation identified that the change in configuration required a safety analysis to determine if the modification involved an unreviewed safety question. The safety analysis determined that i no unreviewed safety question existed. Therefore, the modification was made pursuant to 50.59 without NRC approva The inspector noted, however that the modification was not reported '

to the Nuclear Licensing department for inclusion in the " Operating Plant Status Report" which informs the NRC of plant modifications as required by Nuclear Engineering Procedure. However, a Quality Assurance surveillance (88-5-591) had also identified a similar I problem concerning temporary procedures changes not being reported I as required by 10 CFR 50.59(b)(2) and Technical Specification 6. l J

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The licensee had'taken action on the Quality Assurance finding by

,. iss'uing aJstop work order as discussed below, c , !The' inspector concluded that the licensee's actions in response to

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the QA audit finding appeared adequate to resolve this issu 'c .. Stop Work Order ,

4, . A weakness in the licensee's control of temporary modifications was discussed during the previous inspection report. Stop Work Order No.88-003 had been issued by QA during the previous inspection period to stop the use of PDQs (Potential Deviation from Quality) to implement temporary modifications. The inspector reviewed the

' licensee's resolution in response to the stop work order. The inspector found that the licensee had issued a change to the PDQ RSAP-1308~" Potential Deviation from Quality," Revision 2, adding a design checklist to improve the control of modifications made using the PDQ procedure. The Stop Work Order had also been revised to allow the use of the revised PDQ procedure to implement configuration changes. The revised Stop Work Order continued in effect during the report period pending review of previous PDQ disposition .In discussions with Quality Department representatives, the inspector found that the revisions to the PDQ procedure and the stop work order were considered to be interim resolutions to the proble Engineering was to develop a minor modification procedure to adequately control minor configuration changes without having to l involve the more cumbersome DCP (Design Change Package) controls developed for major modification The progress in the licensee's resolution of the stop work order will continue to be addressed during the next inspection perio No violations or deviations were identifie . Onsite Followup of Events (93702) j

- On December 9, 1988, the plant tripped due to high pressure in the reactor coolant system resulting from main feedwater control problem !

When the operators routinely transferred rod control from manual to automatic, an unexpected main feedwater flow oscillation was initiated which resulted in an underfeed to the Once Through Steam Generators (OTSG). The licensee determined the root cause of the transient to be drift in the output signal of the megawatt calibrating integral of the Integrated Control System (ICS). The drift introduced an erroneous feedwater demand signal when control was transferred to automati Operator action to manually reduce feedwater flow in response to the transient resulted in an underfeed to the OTSG The inspector reviewed the licensee's root cause determination and corrective actions implemented prior to restart. The inspector found the licensee's troubleshooting to be thorough and well supported by the supplemental diagnostic instrumentation used to monitor the control

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system performance. The plant restarted on December 12, 1988. When

, reactor power was about 12%, low steam pressure in the Auxiliary Steam System resulted in the initiation of the Emergency Feedwater Initiation and Control (EFIC). System, followed by the' operators tripping the reactor. Following the trip, the plant cooled dcwn to 527*F due to continued ' supply of auxiliary steam to the 4th point heater. The operators allowed the 'B' steam generator to empty. The plant was shut down to plan and conduct corrective actions which continued until January 6,198 The reactor was tripped manually by the onerators per procedure when the steam generator level fell to the Emergency Feedwater Initiation and Control (EFIC) System setpoint and EFIC initiated, starting auxiliary feedwater flow. Steam generator level fell due to a main feedwater pump transient caused by low steam pressure to the feedwater pum The low auxiliary steam pressure was a result of personnel error and equipment failure. Early on the morning of December 12, 1988, the Auxiliary Steam System was receiving its supply from the Main Steam System through a pressure reducing valve (PV-36014A). The controller for this valve malfunctioned. The licensee elected to isolate this valve and provide steam from the Main Steam System via the manual bypass valve around the regulatory valve. This activity was controlled by an existing procedur Later on that morning, a second control valve (PV-30702) failed partially open. . This regulator provided reduced pressure steam from the Auxiliary

~ Steam System'to the main feedwater pumps when the main turbine generator load was less than approximately 30% of full power. The second regulator did not have an~ installed bypass around the regulator ond no procedure addressed: compensatory measures for this regulator. The consequence of the second regulator sticking partially open was that the downstream safety valve began liftin The licensee's response to this second failure was to _begin manually throttling:the first valve upstream of the PV-30702 regulator, to lower

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downstream pressure to reseat the lifting safety valve. Approximately three hours later the reactor was manually tripped by the operators due

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to an.EFIC start on the "B"' steam generator low level. The low level was apparently caused by operators throttling control of steam pressur In response to this transient, the licensee modified the steam supply for main feedwater heating following a reactor trip, and provided remote valve isolation from the control room of the auxiliary steam to the /

feedwater heater. The licensee stated that this modification would -

prevent-the' post trip excessive steam load on the steam generators and j maintain the reactor coolant system temperatures within the expected I rang The NRC expressed concern about the events leading up to the reactor trip ,

and issued a Confirmatory Action Letter (CAL) to the licensee confirming I that the licensee would maintain the reactor subcritical until the licensee conducted an investigation of personnel actions and procedures used during the even !

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. . 1 Licensee corrective action centered on two key areas concerning hardware and management decision making during the event. Both regulating valves were repaired and failure modes determined. The auxiliary steam regulator (PV-36014A) failure was attributed to vibration at the valve i, causing a pilot adjusting cap in the pneumatic controller to work loos I The second pressure regulator was failed due to a metal foreign object j jammed in the valve seat which the licensee believed was a pin from a  !

bottom guided globe valve (MSS-032) in the Auxiliary Steam Syste The licensee also identified that the decision to maintain the reactor at power with two failed regulating valves in series which had failed was incorrect. Also, the compensatory measure taken for the second regulator failure required the throttling of a valve, but no procedure was written for this special activit Licensee action appeared to be extensive and complete in the )

investigation of this event. Action taken by the licensee appeared to be  !

adequate to prevent reoccurrence. The Confirmatory Action Letter commitments were satisfied by the licensee on January 5,1989, and the NRC had no objection to planned restart of the reacto . Followup Items (92701, 92700)

Temporary Instructions TI 2515/98 (Closed), " Containment Temperature" This Temporary Instruction required the review of licensee containment temperature records, when operating at high power levels during the hottest portions of the summer. The purpose was to verify that the containment temperature did not routinely exceed the containment design temperatures as stated in the Updated Safety ,

Analysis Report (USAR) or the equipment qualification temperature for containment equipment. Data from 1978, 1979, and 1980 were reviewed for summer operation at full power. The. unit did not nave 4 high power operations in the summers of 1981-1987. Due to the restart program, power levels for the summer of 1988 were not above j about 65% reactor power, but this temperature data was also reviewe t l

The inspector concluded that the bulk average containment air j temperatures were less than the equipment qualification temperatures j of 150 F maximum for the reactor and steam generator compartments, )

and a maximum of 120 F for the containment general area. The only exceptions noted were for the containment dome air temperature which j was measured as high as 126 F for several days and the local reactor j cavity measurements as high as 153 F. Licensee calculations based on these local temperature measurements also indicated acceptable bulk average area temperatures. This item is close ;

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Unresolved Item ,

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E 3 y ' l88-10-01 (Closed),;"EDG Backup Air Supply Design Basis" L

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This. item was identified in a special operational team inspection

and ' concerned the design capacity of the Transamerica DeLaval (TDI)

3 , f , Emergency Diesel Generator backup air supply. The item was reviewed

by the team in' Paragraph 7 of Inspection Report 88-19 during a

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',% return visit. The tedm found the design acceptabl This item is ?

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.d Licensee Event Reports'(92700)

E '88-05-LOL(Closed),"LetdownReliefValveActuation"

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lThis LER.~c'oncerned a relief valve that stuck open during the performance of STP.1156 on March 15, 1988. Licensee corrective actions and inspector followup are documented in Inspection Report 88-05, paragraph 2.e. Licensee corrective actions taken have

, appeared adequate to prevent a recurrance. This' item is close i Special Reports 88-03-X0 (Closed),~" Inoperable Fire Suppression System" This report concerned the CO 2 fire suppression system in the Nuclear Services Electrical Building (NSEB).. The licensee had determined that a fire in the hallway of this building could inadvertently render both NSEB Essential HVAC trains inoperable due to the closure of the HVAC dampers stopping air flow into both switchgear rooms. Due to the undesirable system interaction, the licensee subsequently disarmed the C0, system and took the proper technical specification compensatory measures. 'The licensee' subsequently identified that the NSEB CO,, system was redundant in meeting the 10 CFR 50 Appendix R requir8ments because the walls.of the switchgear rooms were adequately rated as a fire barrier to meet Appendix R requirements. By letter dated September 7, 1988, the District documented'an intent to permanently disable the syste The' licensee is pursuing a technical specification change to delete .

this system. In the interim, the licensee has maintained the proper compensatory actions. This item is close j No violations or deviations were identifie . Management Meetings (30702)

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On January 3, 1989, a meeting was held with the licensee to discuss their actions taken as a result of the December 12, 1988 reactor trip. A j Confirmatory Action Letter was issued December 13, 1988 and resolved at i l the January 3,1989 meetin i l

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10. Unresolved Items

' An ' unresolved ~ item is a matter about which more in' formation is required to ascertain whether it is an acceptable item, a deviation, or a violatio . 11. Exit-Meeting (30703).

The inspector met with licensee representatives (noted in paragraph 1) at various times during the. report period and . formally. on January 13, 198 The scope and findings of the inspection activities-described in this report were summarized at the meeting.. Licensee representatives

acknowledged the inspector's findings at that' tim .

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