View source for 05000312/LER-1985-016, Corrected Pages of Rev 1 to LER 85-016-01

{{Adams
| number = ML20207M774
| issue date = 12/31/1986
| title = Corrected Pages of Rev 1 to LER 85-016-01
| author name = Ward J
| author affiliation = SACRAMENTO MUNICIPAL UTILITY DISTRICT
| addressee name = 
| addressee affiliation = 
| docket = 05000312
| license number = 
| contact person = 
| document report number = LER-85-016-ERR, LER-85-16-ERR, NUDOCS 8701130281
| package number = ML20207M752
| document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
| page count = 3
}}
{{LER
| Title = Corrected Pages of Rev 1 to LER 85-016-01
| Plant = 
| Reporting criterion = 
| Power level = 
| Mode = 
| Docket = 05000312
| LER year = 2085
| LER number = 16
| LER revision = 
| Event date = 
| Report date = 
| ENS = 
| abstract = 
}}

=text=
{{#Wiki_filter:d.)                                                                                                   1 feltC Perm 3BSA                                                                            U.S. NUCLEAA RETULATORY COMMIS$ ION LICENSEE EVENT REPORT (LER) TEXT CONTINUATION                           Apenovso oms No aisomo.
EXPIRES: 8/31/88 PACILITV NAast H)                               DOCKET NUMBER (2)
LER NUMBER (6)                     PAGE (3)
RANCHO SEC0 NUCLEAR GENERATING STATION                               vraa "M.7,7      ' O *32                  -
UNIT NO. 1 serr_
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                    .           ancre     wnn On August 8 and August 14, 1985, while in cold shutdown, the Decay Heat Removal System (DHS) suction block valve (HV-20002) automatically closed on a high Reactor Coolant System (RCS) pressure signal, thus resulting in a temporary loss of the DHS system capability. In both cases, DHS flow was re-established in eleven minutes or less, and no noticeable increases in the incore temperatures were detected.
HV-20002 is a motor operated valve located in the decay heat removal drop line which is closed during plant operation and must be opened to initiate decay heat removal operation. HV-20002 is interlocked with Core Flooding Tank isolation valve HV-26514 and with RCS pressure transmitter PT-21099 to help provide decay heat removal pump suction header over-pressure protection in conjunction with the DHS relief valves. The DHS design pressure is 300 psig. RCS transmitter PT-21099, located off Reactor Coolant System (RCS) "B" hot leg, provides a RCS pressure signal. A cable exists between the RCS transmitter PT-21099 and SFAS cabinet H4SAB2. DHS utilizes components within cabinet H4SAB2 such as the power supply unit, buffers, and the test modules.
These components do not affect / actuate the SFAS system. Automatic closure of HV-20002 occurs when the RCS pressure exceeds 255 psig. The RCS pressure recorded by operations personnel at the time of the events was approximately 230 psig. PT-21099 was replaced and calibrated during the Cycle 7 Refueling Outage and a successful maintenance test was performed following the events to ensure the proper operability of the decay heat valve interlock and associated instrumentation. An on-going Root Cause Investigation notes that several transmitters have served in this capacity. Upgrades have been done with respect to the Environmental Qualification Program. The more recently used Rosemont transmitters appear to be more sensitive to this voltage spiking than previous transmitters used.
There were three similar trips of the decay heat removal system, each following spurious closures of HV-20002. Alternate decay heat removal capability was available via the RCP-D/0TSG, which was operating concurrently with the DHS, during the December 29th incident. The second incident, at 1618 hours on December 30, 1985, followed the shutdown of RCP 1 The incident was mitigated by resetting the valve closure circuit, reopening the drop line valves, opening the MOV breaker, and restarting the DHS "A" at 1624 hours the same day. The third incident on December 31 was mitigated by restarting RCP-D/0TSG cooling, and re-establishing DHS "A" operation; both within 30 minutes of the trip. Due to the presence of alternate decay heat removal capability (RCP/0TSG), these three trips were determined not reportable in accordance with 10 CFR Part 50.73.
l               8701130281 861231
: l.              PDR     ADOCK 05000312 S                     PDR pg.oau 3                                                                                                .u s oeo iese.om4 saa4ss

m-LICENSEE EVENT REPORT (LER) TEXT CONTINUATION                              i.PeRovEo oMe mo. uso-oio4 EXPtRES: 8/31/08 FAC1UTY NAast (1)                             DOCKET NUMBER (2)                 LER NUM8ER (6)                       FAGE (3)
RANCHO SEC0 NUCLEAR GENERATING STATION                                yp.     .
se=y        .
q.=
UNIT NO. 1 o 15 l o l 0 l 0 l 3l 1 l2 81 5 0 l 1l 6      -
0l1       0l 3 OF       Ql4 wira-. - . = -.            ee a-wnn The spurious decay heat isolation signal was traced to a shielded instrument cable improperly routed (cable number 1R1S04B6A) through Channel B power trays and conduit to a penetration, as documented in NCR S-5263, Revision 3. This instrument cable was originally routed at plant startup. The voltage spike on the PT-21099 instrument circuit appeared, in a test, when the power circuit to Motor Operated Valve HV-20002 was energized. The spikes caused the decay heat "B" interlock to actuate. The existing instrument cable was abandoned in place. A new instrument cable was routed in instrumentation cable raceways since this event. Testing of the new instrument loop will be completed by December 1,1986, once the Safety Features Panel involved is available.
Further investigation of the power cable tray involved revealed that 47 additional power cables were also routed with the HV-20002 power cable.
Any one of these power cables could, and most likely did, cause a voltage spike similar to that found during testing of HV-20002. The problem was corrected by moving the PT-21099 circuit out of the power circuits raceway. The District also checked the routing of Pressure Transmitter PT-21092 circuitry which provides the high pressure isolation signal for HV-20001, and found it in a similar configuration to PT-21099.
This routing will be corrected prior to plant restart.
The NEPM construction specifications and corresponding inspection documents were issued subsequent to the occurrence of these problems so that adherence to these procedures should prevent recurrence.
* The CRTS (Cable Routing Tracking System) is a computer program and set of administrative controls that has tracked electrical cable configuration since its installation in July 1980. The cable in question was installed prior to CRTS' use. The CRTS program has the ability to discern intermixing among Quality Class 1 Channels A, B, C, and D cables and prevent it, as well to preclude the installation of instrument cables within power cable raceways for Quality Class 2 and 3 cables.
The CRTS staff is composed of a full time qualified engineer who acts as an administrator; an engineering aid; and a data entry operator. Four additional engineers were added to the CRTS staff in order to execute a comprehensive program at Rancho Seco to rid the system data base of its discrepancies and deviations. The comprehensive program is being undertaken with continuing overview by the Rancho Seco Quality Department.
FonM 346*
* U S GPO 1986 0624 638,465

r MAC Per. 3BBA                                                                            U S. NUCLEIA EEIULATORY COMMISSION
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  ,                            LICENSEE EVENT REPORT (LER) TEXT CONTINUATION                     muovEO ous No. 3 so-oto.
EXPIRES: 8/31/8B FACILify NAasE (1)                                 DOCKET NUMBER (2)
LER NUM8ER 16)                     PAGE(3)
RANCHO SEC0 NUCLEAR GENERATING STATION                               vpa      55=m           -
3*,7; UNIT NO. 1 um ,aner o l5 lo l0 l0 l 3l1l2 8l5 -
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  .   .a           .,.w.v.            .ms wnn The comprehensive CRTS correction program includes a rigorous statistical survey of plant electrical equipment, raceways, terminations, and cables in order to ensure that the CRTS computer model and its reports accurately reflect the plant's electrical configuration. The correction program schedule has eight parts. The first six items are to be completed prior to plant Restart. The remaining items are to be complete before the conclusion of the Cycle 8 Refueling Outage.
: 1. Improve Efficiency of CRTS Program Operations.
: 2. Resolve Identified CRTS Problems.
: 3. Perform Engineering Analyses of Problems.
: 4. Perform Walkdowns of the CRTS Problems that Require Field Verification.
: 5. Complete the CRTS Data Base Verification after Data Base Update.
: 6. Document and Revise Where Needed the Existing Administrative Control Mechanism by Which Engineers, Designers, and Constructors Use CRTS.
: 7. Establish Annual Software Verification Drogram for CRTS with Rancho Seco Quality Department.
: 8. Establish Improved CRTS Electrical Configuration (NEP 4109) Configuration Control Procedures.
During the above program those cable configurations found to be in an unacceptable arrangement with respect to Nuclear Safety will be reconfigured as necessary to achieve conformance prior to plant Restart.
These actions will enhance the Cable Raceway Tracking System and rid it of its discrepancies and deviations, making it the accurate and precise engineering and construction tool it was designed to be. CRTS, thereby, will continue to prevent the installation of instrument cables into control cable or power cable raceways. Furthermore, should any other original plant CRTS occurrences of similar nature exist, these will have been detected and corrected.
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