IR 05000312/1987009

Insp Rept 50-312/87-09 on 870228-0417.No Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Maint,Surveillance,Esf Sys Walkdown,Facility Mods,Testprogram,Procurement,Warehousing & Bulletins

ML20214Q887
Person / Time
Site:	Rancho Seco
Issue date:	05/14/1987
From:	Dangelo A, Miller L, Myers C, Perez G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20214Q876	List: IR 05000312/1987009 ML20214Q873
References
50-312-87-09, 50-312-87-9, IEB-85-003, IEB-85-3, IEIN-86-003, IEIN-86-3, NUDOCS 8706050242
Download: ML20214Q887 (14)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

i Report No: 50-312/87-09 l Docket N !

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License No. DPR-54 Licensee: Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Facility Name: Rancho Seco Unit 1 Inspection at: Herald, California (Rancho Seco Site)

Inspection conductdd* Febr y2qth ugh April 17, 1987, inspectors:

A()). D'Angelo JMa lY~h Date Signed

~i RrpidentInspector ( 4- n b lb N j C. (K Ppn ,Resip hpedtpr Date Signed

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l G. I arc , Res nt ,pector Date Signed Approved By: [ 'l

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b/N7 L. (./ Miller, Chief, . actdr ProjectsSection II Date Signed Summary:

Inspection between February 28 and April 17._1987 (Report 50-312/87-09)

Areas Inspected: This routino inspection by the Resident Inspectors involved l the areas of operational safety verification, maintenance, surveillance, ESF system walkdown, facility modifications, test program, procurement.

l warehousing, welding, and followup of open items, bulletins and information

notices. During this inspection, Inspection Procedures,30702, 30703, 36100, ,

l 37701, 37703, 387010, 387020, 40700, 42700, 55050, 01726, 62703, 71707, 71710,

! 72701, 90712, 92700, 92701, 94702, and 94703 were use l l Results: No violations or deviations were identified.

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8706050242 870519 gDR ADOCK0500g2

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i DETAILS 1. Persons Contacted Licensee Personnel I

J. Ward, Deputy General Manager, Nuclear

. Bibb, Restart Implementation Manager '

G. Coward, Executive Assistant, Special Projects I 8. Day, Deputy Nuclear Plant Manager J

3 . McColligan Assistant to Nuclear Plant Manager R. Ashley, Licensing Manager l 0. Army, Nuclear Maintenance Manager i

W. Croley, Nuclear Plant Manager G. Cranston, Nuclear Engineering Manager  ;

W. Kemper, Nuclear Operations Manager [

J. Shetler, Implementation Manager [

T. Tucker, Nuclear Operations Superintendent  ;

M. Price, Nuclear Mechanical Maintenance Superintendent I L. Fossom, Deputy Implementation Manager  !

R. Colombo, Regulatory Compliance Superintendent i J. Field, Nuclear Technical Support Superintendent l S. Crunk. Incident Analysis Group Supervisor gF . Kellie, Radiation Protection Superintendent  !

S. Knight, Quality Assurance Manager C. Stephenson, Senior Regulatory Compliance Engineer J. Irwin, Supervisor, I&C Maintenance ,

C. Linkhart, Electrical Maintenance Superintendent (

R. Cherba, Quality Engineering Supervisor l T. Shewski, Quality Engineer Other licensee employees contacted included technicians, operators, l mechanics, security and offico personne I Management Analysis Company (MAC) Personnel Operational Safety Verification  !

The inspectors reviewed control room operations which included access control, staffing, observation of decay heat removal system alignment, }

and review of control room logs. Olscussions with the shift supervisors '

and operators indicated understanding by these personnel of the reasons for annunciator indications, abnormal plant conditions and maintenance l work in progress. The inspectors also verified, by observation of valve and switch position indications, that emergency systems were properly aligned for the cold shutdown condition of the facility during the Decay Heat Romoval "A" train outage. Walkdown of the systems and procedure review are further discussed in paragraph 2.d and The inspectors observed the control room operators in their conduct of operations during regular plant working hours, backshlft and turnover periods and found their demeanor to be appropriate.

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i 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 practices, i Nuclear Service Raw Water Pump Trip i

On March 12, 1987, the licensee reported a trip of the "B" train nuclear service raw water (NSRW) pump resulting from a surveillance start of the "B" train emergency diesel generator (EDG). The inspectors reviewed the licensee's investigation of the NSRW pump trip and found that the licensee was unable to identify the cause of the trip or to reproduce the event during subsequent EDG starts. In discussions with cognizant licensee personnel, the inspectors reviewed the licensee's plan to establish the root cause of the -

,i event or to establish precautionary compensatory measures prior to restoring the system to operability. The licensen planned to add

! additional instrumentation to the B" NSRW pump in a continuing effort to effectively troubleshoot the problem. The inspectors found the licensee action to be adequate but noted that a formal maintesance troubleshooting procedure had still not been implemented at this time for equipment malfunctions which did not result from significant plant transients. Maintenance troubleshooting of ,

equipment malfunction will be reviewed further in the next l Inspection report under existing open item 86-30-0 j Reactor Protection System Channel Separation l r

On February 31, 1987, the licensee reported a deficiency in the l design of the reactor protection system (RPS) circuitry involving a lack of channel separation. The licensee had been notified by Babcock and Wilcox (D&W) (NSSS vendor) on March 4, 1987 of a potential for failure of two RPS flux channels resulting from the single failure of a common amplifier. The licensee verified the as-built configuration of the RPS circuitry per the drawings analyzed by B&W. Since their Technical Specifications allow continued power operation with a single RPS channel in bypass (untripped), the ifconsee determined that the consequence of a ,

single failure of the subject common amplifier could have resulted ;

in only one operable RPS channel while under channel bypass operation. A valid high flux signal on the one operable RPS flux (

channel would not have satisfied the 2 out of 3 RPS trip logic to !

generate a trip signal for the RPS high flux trip and the RPS l power-to pumps tri The licensee subsequently submitted a 10 CFR 21 report on l April 6, 198 <

The inspectors reviewed the licensee's actions in evaluating the ;

safety significance of the design deficiency, in discussions with !

cognizant Ilconsen technical representatives and management, the :

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i inspectors reviewed the lir W s action to analyze the

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consequences of lack of ch, ,. eparation in addition to their I corrective actions to char : - .ircuitry to establish proper separation. Licensee revi

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analysis has determined that a postulated control rod e# - :. accident at power and beginning of 1 core life appears to be *t e rust significant transient which could I

occur concurrent with a iuss of the flux trip function. Currently, the licensee is having an analysis done to determine consequences of i

this postulated failure and will revise their LER 87-23.

i This item will continue to be followed by the inspectors during

review of the licensee's corrective actions in a subsequent

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inspection report, and tracked with LER 87-2 c. Temporary Modification of the High Point Vent Piping l While establishing the valve line-up for fill and vent of the

reactor coolant system on February 28, 1987, operations personnel i encountered a procedural deficiency which did not properly account

{ for a temporary modification of the RCS high point vent piping. This temporary modification modified the system in that reactor coolant ;

isolation valves which were described by procedure were removed and

, the piping configuration changed without a procedure modification to i describe the change. Subsequent root cause investigation by the

licensee identified several failures which contributed to the l failure to control of the temporary modification. The RCS boundary i was maintained during the fill and vent activity by newly installed i valve !

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The inspectors reviewed the licensee's followup of this event and j found the licensee's actions to be adequate. In discussions with i

licensee representatives, the inslectors expressed concern that the j multiple failures indicated a lac ( of attention to detail in the i control of temporary modifications to the plant which require l changes to Operations procedures. Adequacy of Operations procedures

] which were affected by the major modifications in this outage is an

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area which will be reviewed further in a subsequent inspectio l d. Alternato Means of Decay lleat Removal

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In order to establish alternate means of decay heat removal in support of a scheduled upcoming dual train decay heat removal system (0115) outage, the licensee performed two special tests to determine .

the capability of the reactor coolant system (RCS) and the steam I

) generators to remove the current shutdown decay heat. Use of the i

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two steam generators for decay heat renoval is permitted under the Technical Specifications in IIeu of the Decay lleat Removal syste )

j The inspectors reviewed the special test procedures. STP-1011 and i STP-1074 and the safety analysis prepared for the tests under

10 CFR 60.59. The inspectors nbserved the pre test briefings and

! portions of the test to verify that thu test was being conducted under appropriate controlled conditions. The inspectors ruviewed  ;

the test data to insure that the data met appropriate acceptance

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- 4 criteria for terminating the test after demonstrating the functional capability. The purpose for running the tests were to verify the adequacy of the planned heat removal path (use of the steam

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generators) and the temperature control of the RCS. The methods I used had already been determined to be adequate by analysi:..

No violations or deviations were identifie . Monthly Surveillance Observation and ESF System Walkdown Technical Specification (TS) required surveillance tests and special j

tests were observed and reviewed to ascertain that they were conducted in accordance with these requirement '

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 the affected components were accomplished; test results confonned with TS

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and procedure requirements and were reviewed by personnel other than the

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individual directing the test; the reactor operator, technician or i engineer performing the test recorded the data, and the data was in

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agreement with observations made by the inspector, and that any

deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne Fortions of the following tests were observed during this period:

STP-1011 Determination of Decay Heat Load STP-1074 Demonstration of Alternate Decay Heat Removal Methods STP-1036 Maximum Differential Pressure Test of Three-Way Valve HV-23004 1 SP-210.21 Special Frequency Auxiliary Feedwater Pump - Surveillance Procedure and Inservice Test i Auxiliary Feedwater Functional Testing i

j in order to insure the ability of the auxiliary feedwater system

(AFW) to supply feedwater to the steam generators in support of i special testing of alternate decay heat removal methods (STP-1011 1 and STP-1074), the licensee performed functional tests of the AFW l pumps and control valves. Functional testing was completed

! successfully. THe inspector walked down selected safety related sections of the auxiliary feedwater system to verify proper system alignments for the tests. No alignment discrepancies were identifie No violations or deviations were identifie . Mnnthly HaIntenance Observatinn a

Maintenance activitics for the systems and compnnents listed below were observed and reviewed to ascertain that they were conducted in accordance i

with approved procedures, regulatory guides, industry codes or standards, l and the lechnical Specifications.

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The following items were considered during this review: The limiting conditions for operation were met while components or systems were removed from service; approvah were obtained prior to initiating the 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 were implemente Motor Operated Valve Refurbishment Program (IEB 85-03)

During this report period, the inspectors observed the progress and findings resulting from the ongoing licensee program to refurbish the plant's motor operated valves (M0V). The licensee had issued two licensee event reports (LER 87-06 and LER 87-06 Rev. 1)

detailing the scope of equipment problems which had been identified during the refurbishment program. The following problems were reported in the LERs: Over-thrust conditions Brake application i Undersized power cables to the operators Lack of staking of stem nuts Valve internals damage Unqualified operator grease l Pickup / dropout voltages out of specification

Due to the expanding nature of the problems identified during the initial progress of the refurbishment program, considerable rework

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and reinspection of MOVs has delayed the scheduled progress of the program. Through discussions with licensee representatives, the inspector found that the licensee appeared to be experiencing organizational problems in controlling the scope of the program while planning and scheduling for completion prior to restart.

! The inspector observed a sample of the work activity conducted under the reorganized refurbishment program involving verification of as-built components and component nameplate data acquisition. The hi inspector the programexpressed managemen several concerns These concerns included noted during(a)s observation to lack of Quality Control (QC) involvement, (b) lack of work request control of the activity,and(c)minimalproceduralguidance. The inspector questioned whether the work activity was appropriately controlled and documented to preclude the necessity for future repetition. The licensee acknowledged the inspector's, concerns regarding work packagin (The work observed was in process and had not been completed at the end of this inspection).

During general work activity observations, the inspector also noted

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MOVs in various stages of refurbishment without adequate control of the interim condition of the equipment. Specifically, SFV-29107 was observed to have the motor removed from the operator without any protective covering over the exposed operator internals. Clear unmarked plastic bags of unidentified parts were found laying in the

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work area. No tagging was visible on the partially disassembled operator to identify any work in progress. The inspector identified these deficiencies to program management for resolution and expressed concern as to the general housekeeping being maintained during refurbishment work activities. The licensee responded to the concern and improved the housekeeping practices use The licensee initiated the maximum differential pressure testing portion of their MOV refurbishment program under STP-1036. The inspector observed portions of the testing and found the activity to be adequately controlled. However, the results of this first maximum differential pressure test again indicated an overthrust condition existed even though the valve had been refurbishe Engineering continued to resolve this problem through the valve manufacture These concerns were brought to the attention of licensee management by the inspector. Additional followup of the continuing M0V program will be addressed in future inspection report b. Environmental Qualification of Limitorque Motor Valve Operator Wiring (Information Notice 86-03).

The licensee's program in response to IN 86-03 was initially reported in Inspection Report 86-30. Replacement of unqualified motor operator wiring had been incorporated into the licensee's ongoing MOV refurbishment program in response to IE Bulletin 85-0 Initial results of the original 30 motor operators in the refurbishment program identified 21 instances of unqualified PVC insulated wiring in the operators as supplied by Limitorque. The licensee subsequently expanded the inspection effort to encompass all motor operators in the plant (approximately 173).

ThelicenseehadconductedaLimitorqueupgradeprogram(MOD 011)

during the 1983 outage which included replacement of internal wiring with qualified nuclear grade SIS wiring on approximately 70 MOV The subject MOVs, located in the reactor building and auxiliary building, had been identified by the licensee as subject to the environmental qualification (EQ) requirements of 10 CFR 50.4 In April, 1985, the licensee identified an additional 21 M0Vs located in potentially harsh environments in the tank farm and turbine building and established EQ requirements for the equipment. In response to IEN-86-03 dated January 14, 1986, the licensee walked down a sample of the EQ MOVs in March, 198 Initial physical inspection in April,1986 of the as-found internal wiring of 7 of the 21 MOVs found all 7 to contain unidentified jumper wiring. The licensee contacted Limitorque to identify the wiring supplied by the manufacturer and was informed that polyvinyl chloride insulated wiring had been supplied in non-containment applications during the 1970-71 time period. The licensee also sampled 5 of the MOD 011 valves to verify the use of qualified Rockbestos SIS or Brand Rex t

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wirin The licensee initiated ECN 0699 to implement EQ upgrading of these valves prior to restar The inspector sampled the refurbishment activity of 18 of the approximately 90 EQ MOV operators to verify the identification of the operator control wiring by physical inspection and review of licensee walkdown data. The sample included valves located in the reactor building, turbine building, auxiliary building and tank i

farm. The inspector observed that actual installed wiring was identified by the licensee based on identification markings or similarity in appearance to other identified wirin The inspector reviewed the documentation and analysis established by the licensee for the as-found PVC wiring. The inspector found that the licensee had concluded that the PVC wiring was not reportable to the NRC under 10 CFR 50.72 or 10 CFR 50.73. The licensee evaluation was based on a NUGEQ (Nuclear Utility Group on Equipment Qualification) position memorandum dated May 28, 1986, on PVC wiring and physical properties of PVC material determined through type testing. The inspector found that the documentation appeared

adequate to establish EQ qualification based on conservative j preliminary thermal hydraulic analysi The inspector found that the scope of the licensee's actions relative to IEN 86-03 appeared to be appropriate. Information

Notice 86-03 is closed. (IN 86-03: CLOSED) Makeup Pump Discharge Valve Replacement (SIM-003)

On April 4, 1987, the inspector observed portions of the maintenance 4 activity and post maintenance testing of valve SIM-003 which was cut out and replaced due to excessive leakage. The inspector found the work to be adequately controlled and performed in accordance with procedural requirement The inspector sampled the post maintenance test results to verify that they met the test acceptance criteria ,

and that the test equipment was proper and calibrate l No violations or deviations were identifie j 5. System Review and Test Program (SRTP)

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The areas of inspection included reviews of newly issued Special Test Procedures (STP) Test Outlines, Administrative Procedures (AP), STP performance and completed STP result The following newly issued STPs were reviewed for scope and adequacy.

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Procedures reviewed were found to incorporate test scope and functionality requirements for the system as defined in the System Status Report documen STP-665 EFIC Cabinet Functional Test

, STP-771 Pressurizer Spray Valve Interlock Functional Test '

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STP-779A CR/TSC Ess. Air Flow Transmitter Data Collection Train "A" STP-781 Main Steam System Acoustic Monitor Hammer Test Rev 1 STP-786 RPS Instrument Ground Verification STP-7870 SFAS Digital Channel "A" Module Removal Int'erlock Verification STP-787E SFAS Digital Channel "B" Module Removal Interlock Verification STP-963 Aux. Feedwater System Heat Tracin STP-1009A New Diesel Generator G100A/GEA2 Engine Integrated Rev 1 Phase 2 Testing STP-1011 Determination of Decay Heat Load Rev 2 Rev 3 STP-1024 Fill NSCW from AFW STP-1031A NSRW Component Flow Verification Loop "A" STP-10318 NSRW Component Flow Verification Loop "B" STP-1035 Max. Diff. Press. Test of M/V Tk. Isolation Valve SFV-23508

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STP-1036 Maximum Differential Pressure Test of Three-Way Valve HV-23004 ,

I STP-1044 Aux. Boiler Interlock and Load Test '

STP-1045 Letdown Valves Interlock Test I STP-1046 Feed and Bleed (Boration/Deboration) Permissive j Interlock I

i STP-1047 HV26105 Differential Pressure Stroke Test STP-1048 HV26106 Differential Pressure Stroke Test STP-1051 SFV-25003 Differential Pressure Stroke Test STP-1052 SFV-25004 Differential Pressure Stroke Test STP-1055B Auxiliary Steam Pressure Reducing Station Transfer Test i

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STP-1059 Operational Verification of Refrigeration System Rev 1 STP-1061 Operational Verification of the Refrigeration System for CR/TSC Ess. HVAC Train 8 STP-1074 Demonstration of Alternate Decay Heat Removal Methods STP-1055A Aux. Steam Pressure Reducing Station Checkout Test STP-1055B Aux. Steam Pressure Reducing Station Transfer Test ,

!' STP-1062A G.M. Diesel Generator (G-886A) Hydraulic Governor Load Limit Setup STP-1062B G.M. Diesel Generator (G-8868) Hydraulic Governor

! Load Limit Setup

! The inspector observed the performance of portions of STP-781, STP-665, STP-1059 and STP-1009A. The actual conduct of these procedures was being performed in accordance with applicable APs and was satisfactor No violations or deviations were identifie . Receipt, Storage, and Handling of Equipment and Materials The inspector selected three recent cable requisitions for Class 1 wor The inspector reviewed the receipt inspection reports for three types of wire. The receiving inspections appeared to be conducted in accordance with the appropriate controls, and nonconforming conditions were dispositioned in a timely manne The inspector also reviewed the storage of the three cable reels. They were found stored in accordance with the Receiving Inspection Data

! Reports (RIOR) of the cable. Markings were available on the cable reels

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that identified the wire type, quality certifications, purchase and receipt document The inspector reviewed a receiving inspection for ASTM A325 bolts, nuts, and wa3hers. The package appeared to contain all the pertinent information needed for the inspection. The certificate of conformance contained description and heat number for the bolts and included the tests performed. The inspector also verified that the supplier was on the QA approved vendors lis No violations or deviations were identifie . Followup Items 86-07-07 (Closed) Licensee to Reexamine Procedures to Assure Proper Control of Noncondensible Gases in an Emergency Inspection Report 86-07 identified that the licensee needed to review its procedures and to establish actions to be taken with regard to

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noncondensible gases collected in the control rod drive mechanism (CRDM)

following a loss of pressurizer leve Inspection Report 86-38 reviewed three licensee operating procedures A.74 " Control Rod Drive System", " Reactor Coolant System", and B.4 " Plant Shutdown and Cooldown". These 4 procedures addressed the venting of the CRDM following any event that...

" caused compensated pressurizer level to decrease below 10 inches indicated" or caused an analysis that shows the pressurizer did not empt The inspector, however, left the item open due to procedures not addressing a condition after an event which could cause the pressurizer to empty and then the plant remains in a hot shutdown condition.

The inspector discussed the_ licensee's emergency procedures with the Operations' Superintendent and concluded that in various situations after the pressurizer empties that there exists appropriate guidance for the operators to assess the condition of the CRDM prior.to exercising the control rods. This includes situations where an event occurs and the pressurizer empties and the plant remains in Hot Shutdown and then returns to criticalit Therefore, this item is considered close Closed, (86-07-07).

86-07-08 (Closed) Add Pressurizer Heater Current Status to IDADS 86-07-09 (Closed) Evaluate Preventative Maintenance of the Interlock Between Pressurizer Heater Current and Pressurizer Low Level During the December 26, 1985 event there was a potential-for the pressurizer heaters to be damaged during a nine minute transient when the heaters were uncovered. As stated in the 86-07 inspection report the inspector concluded that no damage occurred to the heaters. However, the inspector identified two open items: One that the licensee had planned to include the pressurizer heater current to IDADS and two, that the licensee did not provide any preventative maintenance for the low level interlock between the pressurizer heaters and pressurizer low leve The licensee submitted to the NRR in a letter dated November 12, 1986, an accelerated schedule for the implementation of Regulatory Guide 1.97. In this letter the licensee has upgraded the installation of the pressurizer

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heater status to be performed in the next refueling outage, Cycle 8. The inspector was concerned that in light of the November 1986 pressurizer a

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heater burnout the licensee might want to upgrade their schedul i However, it appears that the heater current status would not have aided i

in the prevention of the heater burnout incident, therefore this item is 1 considered close Closed (86-07-08). .

l The licensee has completed their evaluation for preventative maintenance (PM) on the pressurizer level heater interlock and has added this item to the preventative maintenance program. In addition, a work request has already been performed and completed which performed the PM functio Therefore, this item is considered close (86-07-09) Close (Closed) Weld Rod Control Procedure Needs To Be Reviewed After Next Revision The inspector found an incident where two welders had consolidated the

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same type weld rod into a heater can due to a lack of electrical outlets

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in a work area. At the end of the inspection the inspector had no concerns on the welding performed on this job, however he expressed that the administrative control of weld filler material was to ensure that each welder only used weld rods that.were issued specifically to him for a particular jo The licensee stated that the procedural intent would

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be clarified in the next revision of the weld control procedur The inspector found in Welding Manual procedure M.307 Rev.1:

"6.7.5.3... Welders / brazers are responsible for their filler material (s)

, and shall not transfer these materials to other individuals or portable rod warmers." Therefore, this item appears to be closed by the above procedure revision, 86-18-04 is Close . Unresolved Items 86-18-05 (OPEN) " Discrepancies With Licensee's Piping Classification and NRC Safety Evaluation" 86-18-06 (OPEN) " Current Piping Classification Does Not Identify Seismic Piping Past Class I Boundary" During followup of the June 23, 1985 event in which a crack in the reactor coolant system highpoint piping resulted in a non-isolable reactor coolant leak, the NRC began an assessment of the appropriate

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piping classification for this system. This included a safety evaluation report (SER) which was documented in report 86-18 in which two unresolved

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items were issued. The licensee was requested to respond to the items, i

The licensee responded in a letter to the Regional Administrator on September 16, 198 In the letter the licensee discussed the history of their original design

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and constructio For the situation of the high point vent, the seismic analysis extends into the nitrogen supply system piping past valve

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RCS-Oll. The nitrogen supply system piping was structurally designed and analyzed to mes- Seismic Category 2 requirements up to a structural anchor, however, it was classified as Quality Class 3, Seismic Category Therefore, the licensee concluded that the classification of the reactor coolant vent piping and the nitrogen supply piping is in accordance with the required safety guide and the original licensing basis of the plant and that the design functionally provides the double barrier protection from the reactor coolant system and that any more recent requirements must be reviewed in light of 10 CFR 50.109, Backfitting Requirement i The NRC is currently resiewing the licensee's submitta Items 86-18-05 and 86-18-06 will remain OPE No violations or deviations were identifie . Management Meetings The inspector attended the SMUD Implementation Meeting, onsite,

, March 4 1987, and observed briefings by plant personnel on the progress

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of the restart program.

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On March 25, 1987, licensee representatives toured the plant and met with NRC management representatives from Region V and Headquarters and the

inspector on the subject of licensee readiness for restart.

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Region V management and the inspector met with licensee representatives ,

on March 31, 1987 at the NRC Region V office in Walnut Creek to discuss the progress of the restart program. The licensee presented background information on the stop work order issued by QA on welding due to inadequate control of weld filler material documentation.

No violations or deviations were identifie '

1 Regional Requests Pipe Wall Thinning In response to a Region V information request, the inspector reviewed the licensee's program for inspection of pipe wall thinning. The inspector determined that the licensee had recently upgraded the program previously established to inspect steam line piping to include feedwater piping. Also, licensee representatives -

stated that approximately 20 locations in the steam generator feedwater lines will be inspected under the enhanced program prior

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to restar Voluntary Entry into Technical Specification Action Statements In response to a Region V information request, the inspector reviewed the licensee's controls over voluntary entry into technical specification action statement Since Rancho Seco Technical Specifications do not specifically have action statements associated with limiting conditions for operation, the inspector reviewed the

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existing controls over voluntary bypassing of reactor protection system (RPS) channels. Operation with a single RPS channel bypassed

, was allowed by Technical Specification 3.5.1.3, reducing the system j logic from 2-out-of-4 to 2-out-of-3, for the purpose of on-line

testing or in the event of a channel failur Furthermore,

Administrative Procedure AP.1, " Responsibilities and Authorities",

Paragraph 3.10.2.6 required the Shift Supervisor to obtain permission from the Plant Superintendent prior to placing the one RPS channel in bypass while the reactor was critica The Shift Supervisor could solely authorize an RPS channel to be placed in bypass if the action was directed in an approved procedur Effect of Pipe Rupture on Fire Protection Equipment (Surry Event)

] During the feedwater line rupture event which occured at Surry, water from the rupture entered some of the electrical conduits for the fire protection carbon dioxide system causing the system to spuriously actuate.

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The inspectors discussed this problem with licensee personnel and determined that the licensee was aware of the occurence from review of the Surry event. The Operations Fire Protection Coordinator stated that an increased effort to seal breached conduits had been initiated to reduce the suseptibility of the fire protection system to water intrusion. No additional review of the adequacy of the existing system has been initiated or planned by the licensee at '

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this time.

i No violations or deviations were identified.

11. Exit Meeting

)- The inspectors met with licensee representatives at various times during the report period and formally on May 15, 1987. The scope and findings of the inspection activities described in this report were summarized at

, the meeting. Licensee representatives acknowledged the inspectors'

findings.

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1

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