IR 05000206/1985035
| ML13323B070 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/02/1985 |
| From: | Ivey K, Jim Melfi, Thomas Young NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML13323B068 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.D.2, TASK-2.E.1.1, TASK-2.F.2, TASK-2.K.3.01, TASK-TM 50-206-85-35, 50-361-85-34, 50-362-85-33, GL-82-28, GL-83-28, GL-85-6, NUDOCS 8512230044 | |
| Download: ML13323B070 (8) | |
Text
U. S. NUCLEAR REGULATORY COMMI'SSION
REGION V
Report No /85-35, 50-361/85-34, 50-362/85-33 Docket No, 50-361 and 50-362 License No DPR-13, NPF-10 and NPF-15 Licensee:
Southern California Edison Company P. 0. Box 800, 2244 Walnut Grove Avenue Rosemead, California 91770 Facility Name:
San Onofre Nuclear Generating Station Units 1, 2 and 3 Inspection at:
San Clemente, California Inspection conducted:
November 12-15, 1985 Inspectors:
vey, Jr., Reactor Inspecto Da e S gned a. Melfi, Reactor Ins tor Dae Signed Approved By:
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/2 KungJr.,
Chief, Engineering.Section-Date Signed Summary:
Inspection during the period November 12-15, 1985 (Report Nos. 50-206/85-35 50-361/85-34 and 50-362/85-33)
Areas Inspected: A routine unannounced inspection of the implementation of selected TMI Action Items, followup of inspector identified items, and followup of Generic Letters. The inspection involved 53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> by two NRC inspectors on Module Nos. 30703, 92701B, 25565, 92703, and 25401 Results:
No violations or deviations were identifie PDR ADOCK 05000206, G
DETAILS Personnel Contacted F. Briggs, NSSS Electrical Supervisor S. Foglio, NSSS Engineer J. Redmon, Instrumentation and Control (I&C) Electrical Engineer C. Brandt, Quality Assurance Engineer
,M. Freedman, Compliance Engineer B. Douglas, Compliance Engineer
- C. Kergis, Lead Compliance Engineer
- R. Santosuosso, Assistant Maintenance Manager
- T. Mackey, Jr., Supervisor, Compliance
- D. Peacor, Station Emergency Preparedness Manager
- K. Slagle, Manager, Administration
- A. Hammons, Quality Assurance
- B. McGee, Supervisor, Site Procedures
- J. Grosshart, Quality Assurance Engineer
- T. Herring, NSSS Engineer
- M. Short, Unit 1 Project Manager
- P. Stewart, NRC Resident Inspector J. Walderhaug, Computer Engineer R. Jervey, Quality Assurance Engineer
- Denotes those individuals attending the exit meeting, November 15, 198 The inspectors also held discussions with other licensee personnel during the inspectio.
TMI Action Plan Items Unit 1 (1) Item II.B.2.3 (Closed) "Environmental Qualification of Equipment" Environmental qualification of electric equipment important to safety was inspected and documented in NRC Inspection Report No. 50-206/85-30. The inspection was to verify conformance with 10 CFR 50.49 which includes radiation qualification of equipment and therefore included the requirements of this TMI item. This item is close No violations or deviations were identifie (2) Item II.F.2.3.B (Open) "Level Instrumentation for Detection of Inadequate Core Cooling" In association with this.NUREG -0737 item Generic Letter 82-28
"Inadequate Core Cooling.(ICC) Instrumentation System" requested that licensees submit a reactor; coolait inventory system design and evaluate the current ICC instrumentation at their plants. The licensee has committed to submit an evaluation of the current ICC instrumentation capabilities and
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a justification for relief from the requirement to install a reactor vessel level measurement system by March 1, 198 No,violations or deviations were identifie (3) Item II.K.3.1.B (Closed.) "Automatic PORV Isolation/
Test Signal" In a letter dated September 13, 1983 (Crutchfield to Dietch),
the NRC Division of Licensing concluded that the requirements of this item for Unit 1 were met by the existing PORV, safety valve, and reactor high-pressure trip setpoints and that an automatic-PORV isolation system is not required. Therefore, this item is close No violations or deviations were identifie Unit Item II.E.1.1 (Closed) Auxiliary Feedwater System Reliability Evaluatio The inspectors verified that:
(1) The Technical Specifications limit operation with one Auxiliary Feedwater Train out-of-service to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with a subsequen action time limit of 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (Tech. Spec. 3.7.1.2)
(2) The Technical Specifications require verification of position of all locked-open manual valves during monthly inspections of the.system.(Tech. Spec. 4.7.1.2.1.a.2 and 4.7.1.2.1.a.3). The procedure for this verification is also available (Procedure S023-3-3.16 "Auxiliary Feedwater System Monthly Tests").
(3) The licensee has' performed tests that verify that adequate procedures exist for refilling the.steam generator, preventing feedwater. hammer..(Feeding Integrity Test #34A-201-01, performed 1982)
(4) Procedures are available to-plant operators for transferring to and from alternate water sources. Procedure S023-9-5,
"Condensate Storage, and Transft -System" provides, procedures to fill condensate storage tank T-121'from condensate"storage tank T-120,.makeup demineralizer,'-and fire o
'tection sste wate (5) The licensee has procedures available e.o.ensure that after an AFWS outage' due to maintenance or periodic test,.the AFWS valves are properly aligned and a second operator verifies,the alignment. (Procedure S02-6-13 "Work Authorizations'
pp.6.13 system.restoratiod)
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.(6). Technical Specifications require!a flow test of the AFWS to verify.normal flow 'path from the primaiy water 'source to the steam generators following.acold shutdown (Tech, Spe III).
4.7.1.2.2). The procedure for this is S023-3-3.16.2 "Auxiliary Feedwater Flow Test", which is performed prior to entering Mode 2 after a cold shutdow T) Redundant AFW primary water level indication and low level alarms are available in the control roo (8) The licensee has performed 48-hour endurance tests on all AFW pumps. NRC acceptance of the pump test results are documented in a letter dated October 27, 1984 (Knighton to Baskin). Item I.D.2 Safety-Parameter Display System (1) Unit 1 (Open)
-The licensee currently utilizes a technical data transmission system that allows'persons 'in the TSC and EOF 'to receive-plant status informationduring an emergency condition. This syste has not been evaluated'to determihe.its adequay Iin fulfilling the SPDS function and': there is no uSPDSdisplay in the control roo The licensee plans to complete development of the SPDSdesign criteria when the Human.Engineering Discrepancies (IKDs) are resolved in the' Control Roi Design Revie (CRDR).
The present schedule for development of th PDS criteria is by October 10, 1986 with submittal of the criteria to the NRC by January.9, 198 This it em is open' for Unit. (2)
Unit '2 (Closed)
The SPDS. function is.performed by the Critical Factors Monitoring System (CFMS) and the SPDS. Included in these systems is the Q-SPDS (for seismically qualified parameters).
This item was open previously, (IR 50-361/85-12)'with the only item necessary for closure being installation and implementation of the Heated Junction Thermocouple (HJTC)
system. Completion of the HJTC system and input to the SPDS has been verified in previous NRC inspections (IRs 50-361/85-13 and 85-20).' This item is closed for Unit (3)
Unit 3 (Open)
The SPDS function is performed by the critical function monitoring system (CFMS) and the.qualified safety parameter display system (QSPDS) which is seismically qualifie The CFMS and QSPDS inputs and associated software will.be implemented by the end of the first refueling.outage (January 1, 1986).
The HJTC system will be verified and validated operational at that tim (Ref. letter October 28, 1982, Baskin to Knighton).
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Therefore, the licensee is committed to completion of an SPDS system by January 1, 198 This item is open for Unit No violations or deviations were identifie.
NRC Open Items and Followup Items /85-16-01 (Closed) "Trending of Reactor Trip Breaker (RTB)
Information" Previous Inspection The licensee's response to Generic Letter 83-28 "Generic Implications of ATWS Events at the Salem Nuclear Power Plant" stated that, "the RTB undervoltage device and shunt trip device independent actuation, trip time and triptorque are all measured, and trending of these parameters is'evaliiated and documented". The inspectors were not satisfied that their review of the trending documentation was complete and left it as an open item for future NRC inspectio This Inspection The licensee's NSSS Electrical Department has implemented a computerized trending prog'ramfok RTB parameters that is backed up by hand-drawn graphs. Each time a maintenance or surveillance activity is performed on the RTBs, parameter-information is input to the program for trending.'.Parameters trended by this program include undervoltage trip time, shunt trip time, and trip torqu The inspectors concluded that.this program satisfies the requirements of GL 83-28 and theiefore, this item is close No violations or deviations were identifie /85-16-02 (Closed) "Vendor Recommendations for RTBs" Previous Inspection The licensee's response to GL 83-28 stated that, "All vendor-recommended RTB modifications have been reviewed and implemented as appropriate."
The inspectors verified that vendor
- recommendations on.RTBs were reviewed with the exception of GE Service Advisory Letter (SAL) No. 175-9.21 "AK-25 RTBs Shunt Trip Paddle". Verification of licensee action on SAL 175-9.21 was made an open item for a future NRC inspectio This Inspection GE SAL 175-9.21 concerned improper heat treating of RTB shunt trip paddles in production from February 1983 through April 198 Improper heat treating could result in surface cracking on the paddle. The inspectors discussed the licensee's procedures for review of RTB vendor information with the cognizant engineer and verified that the licensee had reviewed this item. The licensee determined that the shunt trip devices in use at the plant were not
included in the production group in question ahd performed a visual inspection of the shunt trippaddles. The paddles.,displayed no signs of.surface cracking' The licensee'also included a step in procedure No. S023-1-9.27 "Breaker GE Ad 2-25 Annual.Routine Maintenance" to require visual inspection of the shunt trip paddles for cracking. This item is'close No violations or deviations were identified., /85-22-02 (Closed) "Review of Surveillance 4.8.1.1.2.D.6 and FSAR" Previous Inspection In comparing the technical specification 'surveillance requirements to the licensee's procedures and the FSAR, the inspectors'noted that due to the wording of two surveillance requirements, the technical specifications may not be consistent with the plant design. The inspectors left this item open for future NRC inspectio This Inspection The inspectors reviewed the potential inconsistencies and held discussions with licensee personnel to obtain the following conclusions to the two potential inconsistencie (1) Problem:
Surveillance requirement 4.8.1.1.2.d.6 requires that an a simulated loss of the emergency diesel generator with offsite power not available, the loads are shed from the emergency busses. The plant' design for this scenario has the high pressure safety injection (HPSI) pumps remaining connected to the bu Findings and Conclusions:
TS'4.8.1.1.2.d.6 requires that "...the loads are shed from the emergency buses and that subsequent loading of the diesel generator (DG) is in accordance with design requirements".
The licensee stated that this.. statement means the loads are shed from the emergency buses in accordance with design requirements just as the subsequent loading of the DG is in accordance with design requirements.' Furthermore, NRC Generic Letter 83-30 recognized that the surveillance requirement served no purpose, and suggested submitting a TS change request to delete the requirement. Accordingly, SCE submitted Proposed TS Change NPF-10/15-91 to NRR on January 15, 1984. To date, the proposed change has not been approve (2) Problem: Surveillance requirement 4.8.1.1.2.d.7,.b requires that on a simulated loss of offsite power in conjunction with an ESF test' signal, the...DG starts and enerizes the...buses within 10 second As described in the FSAR (Section 8.3.1.1.4.6), the diesels are designed to start and
- load within 10 seconds of a loss of voltage signal (LOVS),
- however, on a loss of power to the emergency buses, an approximately one second delay occurs before the generation of a LOVS. The surveillance requirements are not clear concerning whether the one second delay is considered in the testin Findings and Conclusion:
TS 4.8.1.1.2.d.7.b was written to test the capability of a DG to start and energize emergency buses and does not test the LOVS time delay. However, the LOVS circuitry is response time tested per TS 4.3.2.3 and Table 3.3-5 in accordance with the degraded bus voltage trip setting curve (TS Figure 3.3-1).
This.curve clearly specifies a-trip setpoint of one second when the bus reaches zero (0) volts as described in the FSA The inspectors discussed these findings with the inspector who initiated the item and concluded that this item is close No violations or deviations were identifie Generic Letter 85-06 (Closed) "QA Guidance for ATWS Equipment That Is Not Safety-Related" On June 1, 1984, the NRC approved publication of a final rule, 10 CFR 50.62, regarding the reduction of risk from anticipated transients without scram (ATWS) events for light-water cooled nuclear plant This generic letter was issued to provide explicit QA guidance for non-safety related equipment encompassed by the ATWS rule. Section 50.62(d) of the rule required that each licensee develop and.submit a proposed schedule for meeting the requirements Sof the rule within 180 days after issuance of the QA guidance (issued April 16, 1985).
The licensee responded to this item with separate letters for Unit 1 and.Units 2 and 3 on October 15, 1985.. Their responses were as follows:
Unit 1:
Provisions for automatic initiation of auxiliary feedwater were installed as part of the TMI upgrades. The equipment is primarily safety-related and portions of the system that utilize control grade equipment will be upgraded to safety-related during an upcoming refueling outage. Th:-licensee plans to evaluate the methods proposed by the'.Westinghouse.Owners Groupfor independent
'and diverse initiation of-a turbine trip and implement the design which is best suited for the unit. T licensee has scheduled implementation of the turbine trip design prior to return to service from the Cycle 10 refueling outage, which-corresponds to the second refueling outage following publication of the rul Units 2 and 3:
The licensee's approach -was-developed through participation in the Combustion Engineering"Owners' Group. The licensee plans to install a Diverse Scram System (DSS).ridependent from the existing reactortrip system, (RTS),from the most accessible
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sensor output circuit to interruption of power to the control rod The licensee stated that with the completion of the DSS, the existing turbine trip function,.and the existing diversity between the RTS'and the emergency feedwater actuation system, the'
requirements of the ATWS rule will be met. The licensee has scheduled' the implementation of the.DSS for the first.refueling outages which begin after June 1, 1987 (Cycle 4 refueling outage for Units.2 and 3).
No violations or deviations were identifie NREG/CR-3791 "Closeout of IE Bulletin 79-09" (Units 2 and.3)
IE Bulletin 79-09, "Failures of GE Type AK-2 Circuit Breakers in Safety-Related Systems", was issued-for response and specific actions by all licensees and construction permit holders on April 17,_1979. The bulletin was closed for Unit 2 in Inspection Report 50-361/81-07 and for Unit 3 in IR 50-362/81-0 Additional problems with GE Type AK-2 breakers and Reactor Trip Breakers in general resulted in the issuance of other Bulletins and Information Notices, and.Generic Letter 83-28. NUREG/CR-3791 suggested followup inspection to verify that the latest GE Service Advisory Letter (SAL) recommendations pertinent to GE AK-2 circuit breakers have been incorporated into the licensee's maintenance procedure The inspectors verified that the licensee has incorporated GE SAL recommendations into their maintenance procedures (see paragraph b).
No violations or deviations were identifie.
Exit Meeting On November 15, 1985, an exit meeting was held with the licensee personnel identified in paragraph 1.,The inspectors summarized the scope of the inspection and findings as described in the report.