IR 05000312/1988015

Insp Rept 50-312/88-15 on 880503-31.No Violations or Deviations Noted.Major Areas Inspected:Operational Safety Verification,Esf Sys Walkdown,Maint,Surveillance & Testing, Review of Control Rod Worth & Followup Items

ML20150D886
Person / Time
Site:	Rancho Seco
Issue date:	06/23/1988
From:	Dangelo A, Jim Melfi, Miller L, Myers C, Qualls P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:
Shared Package
ML20150D884	List: IR 05000312/1988015 ML20150D880
References
TASK-2.E.1.1, TASK-2.E.1.2, TASK-2.K.2.08, TASK-TM 50-312-88-15, GL-82-33, GL-88-07, GL-88-7, NUDOCS 8807140214
Download: ML20150D886 (15)

Text

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U. S. NUCLEAR REGULATORY COMMISSION

REGION V

Report No: 50-312/88-15 i Docket N License No. OPR-54 Licensee: Sacramento Municipal Utility District ,

P. O. Box 15830 Sacramento, California 95813 Facility Name: Rancho Seco Unit 1 Inspection at: Herald, California (Rancl.o Seco Site)

Inspection conduc May rp gh ay 31, 1988 Inspectors: A/ i b"2 $- kb AVJ D'Ange o M nio Resident Inspector Date Signed

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t) 6 2.3 4 3 WJ den Inspector Date Signed

/ ers, / ,

'?' ) (n - L 3 - 8V PMM.Qualls,IQpidentInspector Date Signed J

1 Ifi, Regiona T p4fctor

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6Ane Dat'e Signed Approved By: 1 [

(l./ F. Miller, C b@ *

, Reactor ProjectsSection II Date Signed Summary:

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Inspection between May 3 and May 31, 1988 (Report 50-312/88-15)

4 Areas Inspected: U is routine inspection by the Resident Inspactors and in

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part by a Regiona- .nspector, involved the areas of operational safety verification, engir.eered safety features system walkdown, maintenance, surveillance and testing, review of control rod worth and moderator temperature coefficient calculations, and followup items. During this inspection, Inspection Procedures 71707, 71709, 71881, 71710, 62703, 61726, 72700, 92702, 92700, 92701, 93702, 61710, 61708, 62702, 25565, 30702, and 30703 were use Results: No general conclusions regarding the adequacy, strengths and weakness of the areas inspected were identified. No major programmatic weaknesses or breakdowns were observed during this inspection perio PDR ADOCK 050 2

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DETAILS 1. Persons Contacted Licensee Personnel G. C. Andognini, Chief Executive Officer, Nuclear J. Fir 11t, Assistant General Menager (AGM), Nuclear Power Production

• J. Vinquist, Director, Nuclear Quality

• B. Croley, AGM, Technical and Administrative Services 80. Keuter, Director, Nuclear Operations and Maintenance D. Brock, Acting Nuclear Maintenance Manager G. Cranston, Nuclear Engineering Manager W. Kemper, Nuclear Operations Manager

"J. Shetler, Director, System Review and Test Program T. Tucker, Nuclear Operations Superintendent P. Kagel, Nuclear Mechanical Maintenance Superintendent L. Fossom, Manager, Scheduling and Outage Management J. Field, Plant Support Group Supervisor L. Conklin, Manager of Management Controls

• S. Crunk, Manager, Nuclear Licensing T. Fetterman, Electrical Engineering Manager'  ;

J. Irwin, Supervisor, I&C Maintenance Q. Coleman, Quality Engineering Supervisor

, T. Shewski, Quality Engineer i V. Lewis, Supervising Civil Engineer j J. Robertsoa, Licensing Engineer J

P. Bosakowski, Supervisor of Licensing, Technical Support J. Delezinsky, Supervisor, NRR Coordination, Licensing i

• G. Legner, Licensing Engineer

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S. Rutter, Supervisor, Incident Investigation Review Group J 3 (IIRG), Licensing

) D. Schuman, Incident Analyst, Lic3nsing/IIRG  ;

H. J. Sefick, Jr. , Nuclear Security Manager 1

• C, Wallen, Licensing Engineer Other ifcensee employees contacted included technicians, operators, {

mechanics, security and office personne '

Attended the Exit Meeting on May 31, 198 !

2. Operational Safety Verification (71707 -71709, 71881)

The inspectors reviewed control room operations which included access control, staffing, and review of control room logs. Discussions with the

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shift supervisors and operators indicated understanding by these personnel of the reasons for annunciator indications, abnormal plant conditions and maintenance wnrk in progress. The inspectors also

, verified, by observation of valve and switch position indications, that

, emergency systems were properly aligned as required by technical l specifications for the plant conditions.

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' During this inspection period, the licensee operated the reactor at 30%

power until May 4, 1988, then shutdown the reactor for a short maintenance outage and returned to critical reactor operations in mid-May. The plant.was than operated at power levels up to 40% reactor power for the remainder of May. The inspectors verified the technical specification requirements for operations in this mod Tours of the auxiliary, reactor, and turbine buildings, including exterior areas, were made to assess equipment conditions and plant conditions. Also, tours were made to assess the effectiveness of i'adiological controls and adherence to regulatory req 91rements. The inspectors also observed plant housekeeping and cleanhness, looked for potential fire and safety hazards,. and observed security and safeguards practice The inspectors walked'dewn 20% of the protected area boundary and 70% of the vital area boundaries to ensure that they were complete and-that guar were properly posted where known deficiencies existe Jectcrs also observed protected area access control, personnel

. :ng, badge issuing and maintenance on access control equipmen No violations or deviations were identifie . ESF System Walkdown (71710)

During plant operations with the reactor critical, the inspectors walked down and verified the operability of the emergency diesel generator system, the auxiliary feed system and the high pressure infection syste While the reactor coolant system temperature was above 280 , the inspectors: 1) verified the capability to remove heat by use of two steam generators, 2) verified that one atmospheric dump valve (ADV) per .

steam generator was operable, 3) verified that the licensee had the  !

minimum 250,000 gal' .s of water in the condensate storage tank,  !

4) walked down the - isolations, 5) verified the operability of th steam safety valves, 6) walked down the backup instrument air bottle supply systems for the ADVs and auxiliary feedwater (AFW) system valves, i and (7) walked down the duclear Service Flectrical Building Essential i Heating Ventilating and A4: Conditioning Syste No violations or deviations were identifie . Monthly Maintenance Observation (62703)

Maintenance activities for the systems and components listed below were observed and reviewed to ascertain that they were conducted in accordance l with approved procedures, regulatory guides, industry codes or standards, l and the Technical Specification The following items were considered during this review: the limiting

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conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing or calibration was performed

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s 3 prior to returning compon:nts or systems to service; activities were .,

accomplished by qualified personnel; radiological controls were l implemented; and fire prevention controls were implemente ). The inspectors observed maintenance on the A2.Transamerica DeLaval Industries (TOI) diesel which included piping support modifications, snubber adjustments, bolt replacements, addition of air pressure gauges on the backup air bottles, and vibration testin ) Maintenance activities on the auxiliary boilers were observed. ,The licensee was making improvements to upgrade the performance of the boiler ) Maintenance activities on the control system of the "B" mair feedwater pump were observe ) Maintenance activities were observed in reoairing steam leaks outside of the reactor buildin ) Maintenance activiti2s on the integrated control system (ICS) were observe No violations or deviations were identifie . Monthly Surveillance Observation and Review of System Review and Test Program (SRTP) Testing (Manual Chapters 61726, 72700).

System testing under the Systems Review and Test Program (SRTP) was observed and reviewed to ascertain that the testing was conducted in accordance with the requirements of the approved special test procedure (STP). The STP test requirements to demonstrate system functionality were contained in the System Status Reports (SSR). Inspections and

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observations sere conducted during the performance of the.STPs and

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maintenance activities to ascertain that they were ccaducted properl The following testing activities were observed and reviewed during this report period: STP.660 - ICS Tuning i

Various activities were conducted to properly tune the integrated c.ontrol system (ICS) for proper responses to plant behavior. The  !

licersee conducted a turbine trip on May 4, 1988, to get data for the ICS tuning STP. With the reactor at 28% power, the licensee tripped the main turbine. About 30 seconds later the reactor tripped on high pressure. Indications of e leak in the area of the

"A" letdown cooler were observed. The licensee isolated the letdown cooler and stopped the leak. Paragraph 7.A contains additionci details on this event, LT?.1154 - Remote Shutdown Panel Test This STP required that the licensee take local control of the plant from the remote shutdown panel and demunstratt the ability to i

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conduct a controlled plant couldown from outside of the control room from an initial hot shutdown condition. The licensee performed and satisfactorily completed'this test'un May 4, 198 No discrepancies-were noted, The monthly surveillance run of the A2 emergency diesel generator (EDG) was observed. .The licensee encountered a number of vibration problems resulting from high vibration (about 30 mils) on the turbocharge Licensee evaluations' determined that the vibrations were caused by a snubber support attached to the side of the engine near the turbocharger which would loosen during extended engin operatio The snubber support was retightened, but the high vibration was still present. The analysis also determined that the frequency of the turbocharger vibration was about 31 Hz. .This low frequency indicated that the vibration was Seing induced _from an outside source, probably from the engine itsel The diesel and turbocharger vendor representatives both stated that operation with vibration of this magnitude and frequency was acceptable, and that if a turbocharger bearing had been the source of the vibrrtion then the frequency would be about 300 H In addit *on,'when the diesel was operated at 2000 kw (the maximum expected load is about 1700 kw), the vibrations were within acceptable limits. Based on these considerations, the licensee determined that the diesel was operable with the 30 mil vibration at 30 Hz on the turbocharger. The licensee is continuing to monitor and analyze diesel vibrations during each diesel run to preclude excessive vibrations from occurring. The inspector concluded the licensee's resolution was acceptabl No violations or deviations were identifie . Followup on Previous Inspection Items (92702, 92700, 92701) Enforcement Item 87-44-01 NCR Closure (Closed)

By 'etter dated April 20, 1988, the licensee responded to the Notice of Violation from Inspection Report 87-44. The corrective actions specified by the licensee were reviewed and the inspectors concluded that the corrective actions should be adequate to prevent recurrence of the violatio The letter denied tnat the third nonconformance Report (NCR) was improperly ci sed and stated that the pneumatic test performed was properly reviewed by the correct personnel. The inspector found this to be true. It appeared that the review of the work request was accomplished prior to performing the work. The NCR disposition which specified a hydrotest requirement was, however, never revise This aspect of the violation is, therefore, withdrawn. This item (87-44-01) is close ,

, 5- Licensee Event Report (LER) 86-26-L1, Bistable Error in SFAS Circuitry (Closed)

The licensee identified that their safety features actuation system (SFAS) circuitry did not account for errors.in the bistable tri The actions documented by the licensee appeared adequate to correct the problem and to prevent recurrence. This' item is close Special Reports 88-03 and 88-07 Inoperable Fire Suppression Systems (0 pen)

These Special Reports 88-03 and 88-07 dealt with inoperable fire protection systems. .In these reports, the licensee reported the inoperability and the in place compensatory measures for the auxiliary feedwater puno area and the Nuclear Service Electrical Building fire suppression system The licensee's technical specifications also require reporting the projected plans to restore the system to operability and a proposed schedule. As of this inspection, the licensee had still not completed.the plans to restore these systems to an operable statu The licensee committed to update these reports when the plans were complet These items remain ope Follow-up Item 88-12-02, "B" Letdown Cooler Leak Documentation i (Closedl i

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The licensee was unable to find documentation for the1"B" Letdown Cooler leak which has apparently existed since prior to the December 1986 event and shutdown. The. licensee intends to continue operation with the "B" Letdown Cooler isolated until the next refueling j outage, when the unit is in cold shutdown, and to perform lesk l determinations at that time. This item is close l

No violations or deviations were identifie , Deviation 87-14-02, Use of Silicone Sealant in the-Essential Air Filtration System (Closed)

Inspection Report 50 312/87-14 identified a deviation from Regulatory Guide 1.52, which the licensee is committed to in the Rancho Seco Facility Safety Analysis Report (FSAR). This Regulatory Guide states in part that, "Tha use of silicone sealants o' any other temporary patching materials on filters, housing, motnting frames or ducts should not be allowed". The inspector identified in a walkdown of the air system that the licensee was applying silicona sealant on the syste The use of silicone sealant on the Heating, Ventilation, and Air Conditioning (HVAC) system helps maintain the control room pressure boundary. The concern with the use of temporary patching materials such as silicon sealant is that these materials might wear off with time, increasing the possibility of leaks from the syste . _ _ _ _ , ... . . . . ., , _ . - . . . __

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. 6 The licensee initially responded by addressing the Environmental ,

Qualification (EQ) requirements of the silicone sealant and updating '

the FSAR to allow the use of silicore sealant. The licensee was I informed that the NRC concern was erosion of the materia The HVAC system was inspected.by the NRC and noted in NUREG-1286, l Supplement 1, section 4.2. At the conclusion of the inspection, the licensee was still' conducting tests and has committed to submit an 18 month report to the NRC staff.regarding the long-term adequacy of the silicone sealan The licensee's corrective action for this deviation oppeared adequate. This item is close f. Followup on Unresolved Item 88-05-03 Removal of Spare Limit Switch Assemblies from Environmentally Qualified Valve Operators (Closed) (62702)

This issue involved the removal of spare two-rotor limit switch i sub-assemblies from four-rotor assemblies in environmentally '

qualified Limitorque valve operators. The licensee identified on January 15,.1988 the potential adverse impact of the qualification '

of the opera ors in this condition in Corective Action Request (CAR)

88-0 Subsequently, Nuclear Engineering reviewed this configuration under Nonconformance Report (NCR) 7782 dated February 17, 1988 and determined that the condition was acceptable with the spare limit switch rotor sub-assembly remove However, at the next available once all affected operators were to have the spare limit switch sub-asse.?bly reinstalled to return t:1e operator to the originally qualified configuratio During this inspection period, the inspector reviewed the modification package which initially removed some of the spare limit switch sub-assemblies as part of Engineering Change Notice (ECN)

A-3651. The inspector found that this ECN was installed in July,

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1983 as part of the 1icensee's equipment upgrace program to meet the i environmental qualification requirements of_IE Bulletin 79-01 ilowever, the inspector found that there was no analysis in the design basis report for the-ECN to establish an adequate basis for environmental qualification of the resulting limit switch configuratio The inspector reviewed the analysis and partial test data developed by the licensee under NLR-7782. The licensee determined that the specific equipment was qualifiable for the application in questio The inspector found no evidence that the licensee clearly knew or should have known of the deficiency based on prior notifications of similar problems. Based on his review, the inspector determined that the qualification deficiency was not considered sufficiently si,nificant t for enforcement action under the provisions of Generic Letter 88-07, dated April 7,198 This unresolved item is close .. .

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. 7 Onsite Follow-up of Events at Operating Power Reactors (93702)

l Unusual Event; Reactor Trip and Letdown Cooler Relief Valve Lift I On May 4, 1988, with the. reactor at 28%. power,.the licensee was conducting testing in accordance with special test procedure (STP)

660, Integrated Control System (ICS) Tuning. The tist required a- ,

turbine trip from about 25% reactor power. The ICS was expected to l reduce reactor power to 15% following'the turbine trip and control '

power using the Turbine Bypass Valves (TBVs). The turbine was tripped from 28% reactor power to initiate the tes About 30 seconds later the reactor tripped on high pressur At 9.08 the licens had indications of a primary coolant leak. Operations 1 personnel de ermined that the leak appeared to be through the relief l valve on the "A" letdown cooler, which was in service at the time, j Operators isolated the "A" cooler and placed another cooler in 1 service. .The leakage stopped at about 9:40 a.m. when the "A" cooler l was isolated. The leakage was about 1000 gallons released to'the l containme:nt sump. The licensee declared and terminated an Unusual ,

Event at 9:47 a.m. The licensee conducted a post-trip review to l determine the cause of the trip and to confirm the source and cause of leakag The licensee's post-trip review determined that the trip was caused by the ICS response differences between control rod insertion and the feedwater flow reduction rates. The feedwater flow was reduced too rapidly by the ICS, resulting in the overpressure transien Computer modeling by Babcock and Wilcox confirmed this using-the data collected during the tri The coolant leak occurred due to the lifting of' the "A" letdown cooler relief valve approximately 100 psi lower than its setpoin The licensee determined that the valve had been adjusted cold rather than ho The vendor verbally recommended that the valve should be adjusted hot, and that setting the valve cold could account for the ,

premature lift of the valv The licensee's procedure called for i setting the valve per the process standards which did not address  !

the temperature at which to set the valve. The valve failed to-reseat due to damage to the seat and disc when the valve actuation occurre The licensee replaced the relief valve with an identical valve, and returned the cooler to operation. An evaluation by the  !

licensee into the cause of the damage was in prcgres l l

No violations or deviations were identifie Reactor Coolant Pump (RCP) Seal Leakage The l'censee determined during the May 4 outage that a small leak i from the casing of the "C" RCP existe This small leak was of concern due to the boric acid crystal buildup, and subsequent corrosion of the carbon steel pump casing studs. The licensee measured the corrosion rate and obtained an analysis, BAW-1892P, from Babcock and Wilcox, which showed that safe plant operation z_______ __________-__________________ _ ________________ _ ______________ _ ________ _ _ _ __ _ . _ . _ _ _ _ _ _

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could continue ~until late fall, 1988 without pump degradation. The l licensee committed to inspect the RCP studs beture July 's,198 No_ violation = or deviations were identifie . Review of Control Rod Worth (61710)

The inspector reviewed the licensea's completed control rod worth-determination. The inspector verified that:

"* The procedure appeared to meet t' e licensee commitment The prerequisites and initial conditions had been complete The precautions and limitations appeared to have been followe *

Plant condition appeared to have been as specifie *

Selected calculations were checked and appeared correc *

The rod worth values obtained were within the licensee's acceptance criteri No violations or deviations were ' etifie . Review of Moderator Temperaturc ,afficients (61708)

The inspector reviewed the licensee's temperatere reactivity coefficient (MTC) measuremen The inspector verified that:

The measurement met the Technical Specification requirement The prerequisites had been complete *

The licensee appeared to have complied with the test precaution The plant conditions established appeared to be consistent with the licensee's analytical prnd'.:tion *

The HTC met the calculated technical specification requirement No violations or deviations were identifie . Follow-up on TMI Action Items (25565)

The areas inspected in this report were modifications to the licensee's Safety Parameter Display System (SPDS) and Auxiliary Feedwater (AFW)

syste The inspector reviewed previous inspection reports, NRC documents specifying inspection requirements, and the licensee documentation on these items. The status of these THI items are noted below.

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. (0 pen) Item I.D.2.2.3, "Plant Safety Parameter Display Co sole, Fully Implemented" This item requires the addition of a Safety Parameter Display System (SPDS) in the control room to display a minimum set of plant parameters for operating personnel. This provides a redundant system to help rapidly assess how the plant is parforming. The important plant functions that this system shall display include:

(1) reactivity control, (2) reactor core cooling and heat removal from the primary system, (3) reactor coolant system (RCS) integrity,

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(4) radioactivity control, and (5) containment conditions (Generic Letter 82-33, "Clarification of TMI Action Plan Requirements"). The critical functions of the SPDS (Regulatory Guide 1.97,

"Post-Accident Instrumentation") shall be provided during and following design basis events. The functional criteria of the SPDS was described in NUREG-0696, "Functional Criteria for Emergency R1sponse Facilities", and NUREG-0835, "Human Factors Acceptance Criteria for the Safety Parameter Display System".

The requirements for closure and previous NRC inspections of this TMI item were noted in Inspection Report 50-312/88-06. The following sub-items were noted in that report to be inspected for closur (1) Examination of the final turn over to operations of the SPD (2) A review of procedure and Te:hnical Specification changes due to the SPDS modificatio (3) Evaluation of the NRC audit of the upgraded SPDS performed September 29, 198u (See inspection report 50-312/86-39).

(4) Inclusion of certain Regulatory Guide 1.97 variables included  !

onto the SPDS as noted in Inspection Report 50-312/86-3 The 1 variables were coolant inventory, neutron flux, residual' heat removal (RHR) heat exchanger outlet temperature, quench tank '

temperature containment atmosphere temperature, component cooling water (CCW) temperature to engineered safety featur )

(ESF) system, and pressurizer heater statu (5) Resolution of certain issues of an NRR inspection at Rancho Seco on April 7, 1987, as noted in Inspection Report 50-312/87-14. These issues were software validation and ,

qualification of electrical isolator l Additional items necessary for completion of this item are:

(6) On the SPDS installation, ,

(b) Verify that equipment cnanges are properly approved and '

controlle (c) Verify that as-built drawings are changed to show equipment change The above items 1, 2, and 3 were not reviewed during this inspectio Item 4 concerns the inclusion'of certain RG 1.97 variables onto the SPD The inspector was informed that the fully qualified

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instruments would be installed during the next refueling outag i Item 4 will remain open pending installation of the qualified variable Item 5 concerns the issues of software validation and electrical isolator qualificatio These issues were also addressed in NUREG 1286, "Safety Evaluation Report Related to the Restart of Rancho '

Seco Nuclear Generating Station, Unit 1, following the event of December 26, 1985," issued March 21, 198 NUREG 1286 noted that the licensee is undergoing a verification and validation process, and the NRC staff will conduct an onsite audit to evaluate these result Supplement 1 to this NUREG concluded that the licensee's program.to validate the software was adequate, and the isolators were qualifie NUREG 1286 also noted that the NRC will review the licensee's installation test report on the software validation at a later dat The inspector discussed the issues of software validation and electrical qualification with the. licensee. The inspector also examined the cabinets with the SPDS hardware. No i problems were identifie Item 5 remains open pending a review of l the installation test repor !

I Item 6 was reviewed by the inspector. The modification was made on l the SPOS system for full redundancy, Class IE power, seismic ;

qualification, and Regulatory Guide 1.97 Category 1 variable j indication (Engineering Change Notice (ECN) R-0952). No problems l were identified. The equipment changes were properly approved and l controlle The as-built drawings were changed to reflect the system. Item 6 is close This item will remain open pending closure of items 1, 2, 3, 4 and 5

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b. (Closed) TMI Item II.E.1.1.2, "Auxiliary Feedwater System Evaluation

- Long-Term System Modifications" This item requires licensees to evaluate their Auxiliary Feedwater (AFW) System and to upgrade the system where needed based on the evaluation. This evaluation was done to increase AFW reliabilit The previous inspection reports on this item were 50-312/82-06, 50-312/85-21 and 50-312/88-0 The last repcrt left closure of this item pending review of the following items of the Emergency

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Feedwater Initiation and Control (EFIC) syste (1) Installation of the EFIC syste (a) Verify that as-built drawings ane changed to show equipment change (b) Necessary procedure changes have been made and that training has been accomplishe (2) Review of the new ttchnical specifications on the AFW syste (3) Verify that preoperational testing is complet _ .

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(4) Review of the new operating procedure The first item concerned the installation of the EFIC syste The NRC staff performed a Safety Evaluation of the system, NUREG 1286; the EFIC modification was completed under Engineering. Change Notice (ECN) 5415, most of which was reviewed in the previous inspection repor The status of the drawing chaages was in the process of being updated at the time of the inspection. The licensee is keeping track of the Jrawing changes for th9 EFIC system, and provided a status to the inspecto Several of the drawings and installations were reviewed by the inspecto They reflected the system configuration and appeared to be properly approved and 1 controlled. The operators were noted to be trained in Inspection Report 50-312/88-06. Item 1 is close The technical specifications (TS) on the EFIC system were issued under TS amendment 93, issued January 5, 198 The new technical specifications were reviewed and no problems were identified. The licensee is currently implementing the new technical specification Item 2 is close The system hot functional tests under Special Test Procedure (STP)

1113 to verify system operability at temperature were completed March 18, 1988. These tests were reviewed during the previous inspection, with a question on whether the hot functional density compensation for the level in the steam generator is correctly done by the EFIC system. Upon further review, the density compensation for the level in the steam generator appeared to be adequately performed. This test on EFIC system was completed satisfactoril The system is functional at normal operating pressures and temperature Item 3 is close For item 4, some of the new operating procedures were reviewed in NUREG 1286, Supplement No. These procedures were also reviewed by the inspecto No problems were identified in the review of the operating procedures in this repor Item 4 is close Based on the closure of these four items, this item is close (Closed) Item II.E.1.2.1.B.2, "Auxiliary Feedwater System Initiation and Flow," AFW Safety Grade Initiation This item required the licensee to provide a reliable, automatic ,

initiation feature for the AFW system consistent with General Design )

Criteria (GDC) 2 l This item was last inspected in Inspection Report 50-312/88-06. The l previous inspection reports addressing this item were reviewed and the items remaining to be closed from all the inspection reports were summarize The remaining open items are from Temporary Instruction (TI) 2515/65 for the equipment inspection and require:

(1) Equipment Installation / Modification

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. Verify that as-built drawings are changed to show the equipment change Verify that the necessary precedure changes are made and that the necessary personnel training has been accomplishe (2) Operation _ Verify that preoperational testing is complete, Verify that equipment is operable, and that operational procedures are being use Items 1 and 2 on the equ'ipment installation and operation of the EFIC system were performed in conjunction with NUREG-0737 item II.E.1. The equipment changes were verified to be properly approved and controlle The equipment was determined to be operable at the time of the inspection. The preoperational testing was determined to be complete, including for hot functional test Items 1 and 2 are close This item is closed since items 1 and 2 are close (Closed) Itec II.E.1.2.2.C.2, "Auxiliary Feedwater System Initiation and Flow", AFW Safety Grade Flow Indication This item requires the licensee to provide a safety grade indication of the AFW flow to each steam generator in the control roo This is for the control room operators ascertain the actual performance of the AFW system when it is called to perform its intended functio The inspections on this item were summarized in Inspect' ion Report 50-312/88-0 A violation was written on this item (50-312/86-21-05) which stated that the flowrate indicators in the, control room were neither powered by a class IE power supply, nor l were built to class 1E requirement This was closed out in Inspection Report 50-312/86-33, _ noting that the flowrate indicators will be installed concurrently with the EFIC system. These were j verified by the inspector to be installed in the control roo The i work package installing the indicators was' reviewed in Inspection i Report 50-312/88-0 An issue that remained open from the previous inspection reports was I

from NUREG-0737, which stated the following:

(1) The AFW instrnment channels shall be powered from the emergency buses consister.t with satisfying the emergency power diversity requirements of the AFW system set forth in the Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10. The Branch Technicai Position 10-1 (from NUREr.0800, Standard Review Plan, Revision 2) states the design guidelines for the AFW pump drive and power supply diversity for PWRs. The guidelines are:

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. The AFW system should consist of at least 2 full-capacity Lindependent systems that include diverse power sources, Other powered components of the AFW system should use the concept of separate and multiple sources of motive energy, The piping arrangement, both intake and discharge, for each train should be designed to permit the pumps to supply feedwater to any combination of steam generator This arrangement should take into account pipe failure, active component failure, power supply failure, or :ontrol system failure that could prevent system functio The AFW system should be designed with suitable redundancy to offset the consequences of a single active failure, The system should be arranged to assure AFW to the' steam generators, assuming a postulated rupture of any high energy section of the system, assuming a concurrent single active failur Additional items identified in TI 2515/65 for the equipment inspection were (2) Equipment Installation / Modification Verify that as-built drawings are changed to show the equipment changes, Verify that the necessary procedure changes are made and that the necessary personnel training has been accomplishe (3) Operation Verify that preoperational testing is complet Verify that equipment is operable, and that operational procedures are being use Item 1 is concernea with the AFW instrumentation. The applicable portions of BTP 10-1 for the flow instrumentation is addressed by considering sub-items Ib, Id, and le only. Item lb was addressed by the inspector by verifying that the flow instrumentation channels are powered by separate emergency power buse The flow instrumentation in the control room panel.was also verified to be separated. These we.*e noted by the inspector during a walkdown of the system. This addresses the concerns of, item lb. For itemr Id, and le, the logic and failure modes were addressed as noted in NUREG 1286, Supplement 1, Section 3.1.3. In this supplement to NUREG 1286, the EFIC system met the redundancy, seperation, and power criteria for the flow indication. These criteria _sssesser' encompass items Id and le above. The inspector also discussed the logic and j failure modes with the licensee. No problems were identified from the discussions with the licensee with the EFIC logi Item 1 is 4 close I Items 2 and 3 on the equipment installation and operation of the-EFIC system were performed in conjunction with NilREG-0737 item II.E 1. The equipment changes were verified to be properly approved and controlle The equipment was determined to be operable at the time of the inspectio The preoperational testing

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was determined to be complete, including the hot functional test Items 2 and 3 are close This item is closed, (Closed) Item II.'K.2.8, "Orders on B&W Plants, Upgrade of the AFW System" This item was reviewed.in the previous Inspection. Reports 50-312/85-21 and 50-312/88-0 In these inspection reports, it was noted that'this item remained open pending closure of THI items II.E.1.1 and II.E. Since items II.E.1.1 and II.E.1.2 are closed, this item is close . Unresolved Items Unresolved items are matters about which more information ir required to determine whether they are acceptable or may involve violations or deviation . Exit Meeting (30702, 30703)

The inspector met with licensee representatives (noted in Paragraph 1) at various times during the report period and fo*mally on May 31, 1988. The scope and findings of the inspection activities described in this report '

were summarized at the meetin Licensee representatives acknowledged the inspector's findings and violation ident'fie l l

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