IR 05000397/1992032
| ML17289A954 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/16/1992 |
| From: | Royack M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17289A950 | List: |
| References | |
| 50-397-92-32, NUDOCS 9211020051 | |
| Download: ML17289A954 (18) | |
Text
U. S.
NUCLEAR REGULATORY COMMISSION Report Number:
Docket. Number:
License Numbers:
Licensee:
REGION V
50-397/92-32 50"397 NPF-21 Washington Public Power Supply System P., O.
Box 968 Richland, WA 99352 Facility Name:
Inspection Conducted:
Inspectors:
WNP-2 September
18, 1992 F.
Gee, Reactor Inspector P. Loeser, NRR/SICB E. Lee, NRR/SICB Approved by:
M. Royack, Acting Chief Engineering Section
/e /s N>-
Date Signed Inspection Summary:
Ins ection durin the riod of Se tember 14 throu h
1992 Re ort Number 50-397 92-32 Areas Ins ected:
The inspectors conducted an announced inspection of a modification of the recirculation pump trip logic and the licensee's preliminary assessment of plant specific parameters and their potential impacts on plant operation with respect to noncondensible gases in condensing chambers.
The inspectors used inspection procedures 37700, 37701, 92702, Temporary Instruction 2500/020 (25020),
"Inspection to Determine Compliance with ATWS Rule,
CFR 50.62," Revision 2, and Nuclear Regulatory Commission (NRC) Safety Evaluation Reports (SERs)
as guidance for this inspection.
Results:
General Conclusions and S ecific Findin s:
The review of documents and the walkdown indicated that the licensee implemented the modification of the recirculation pump trip logic.
The installed recirculation pump trip logic for the mitigation of anticipated transients without scram (ATWS) appeared 9211020051 921016 PDR ADOCN 05000397 A
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to be adequate.
The review of selected isometric drawings of the condensing chamber connections-to the reactor vessel liquid level sensing nozzles N14 A, B, C, and D indicated that the installations appeared to be in.
conformance with the General Electric guidelines for installing condensing chambers, as the licensee had preliminarily assessed.
The licensee stated that WNP-2 will comply with the resolution of the Boiling Water Reactor Owners'roup regarding condensing chambers.
The licensee had made adequate progress in updating the top tier drawings in the fuse walkdown program.
Si nificant Safet Matters:
None
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Summa of A arent Violations and Deviations:
None 0 en Items Summa The inspectors closed one follow-up ite Details 1.
Persons Contacted J. Baker, WNP-2 Plant Manager R. Barbee, Technical Staff I6C Supervisor J.
Bass, Quality Assurance Engineer D. Feldman, Assistant-Maintenance Manager C. Fies, Compliance Engineer S. Ghbein, Electrical Engineer, Bechtel S. Kirkendall, Plant Support.,Engineering Supervisor R. Koenigs, Design Engineering Manager R. Mazurkiewicz, Chief, Operations Branch, BPA T. Meade, Acting Plant Technical Manager J. Parrish, Assistant Managing Director for Operations W. Shaeffer, Operations Manager G. Sorensen, Manager, Regulatory Programs All of the above personnel attended the exit meeting on September 18, 1992.
The inspectors also held discussions with other licensee personnel during the inspection.
2.
Introduction
. The ob)ectives of this inspection were to:
a.
b.
Verify the plant modification for the recirculation pump trip (RPT) design, which was committed to by the licensee, complied with the anticipated transients without scram (ATWS) rule, 10 CFR 50.62.
Determine the acceptability of removal of fuses in accomplishing steps in shutting down the reactor in ATWS emergency operating procedure (EOP) 5.1.2,
"Reactor Pressure Vessel Control ATWS."
c ~
Review the licensee's preliminary assessment, in response to Generic Letter Number 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(f)," of plant specific parameters and their potential impacts on plant operation with respect to noncondensible gases in condensing chambers.
d.
Evaluate a design modification for the replacement of safety-related pressure transmitters.
3.
ATWS RPT Modification Back round In a safety evaluation report (SER), dated December 6,
1988, of the licensee's ATWS design, the staff found that the licensee's ATWS recirculation pump trip (RPT) initiation logic
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design was not in full conformance with the ATWS Rule,
CFR 50.62.
The trip signal was initiated by either a one-out-of-two reactor vessel low water level signal or a one-out-of-two reactor high pressure signal.
In this design, there were no provisions to test the logic relays that energize the breaker trip coil to trip the recirculation pump.
Additionally, in accordance with the ATWS rule implementation guidance, the frequency of inadvertent actuation and challenges to other safety systems was not minimized.
In response to this 1988 safety evaluation, the licensee committed to:
a ~
b.
Redesign the RPT logic.
Revise the technical specifications to reflect the hardware changes.
4.
Review of ATWS RPT S stem Modification 25020 37701 The inspectors reviewed documents and performed walkdown of the ATWS/RPT system to verify the modification discussed above was implemented as committed to by the licensee.
a.
Design Basis Docume'nts The inspectors reviewed electrical wiring diagrams to verify the implementation of the modification in the ATWS/RPT system.
The inspectors verified that the modified system used coincident logic to trip both recirculation pumps and.
both recirculation pump low frequency motor-generator
,(LFMG) sets during an ATWS event.
Channel A low vessel water level or high vessel pressure coincident with Channel C low vessel water level or high vessel pressure was designed to trip the recirculation pump breaker A and LFMG-A.
Similarly, Channel B coincident with Channel D
was designed to trip the recirculation pump breaker B and LFMG-B.
This one-out-of-two-taken-twice logic should prevent inadvertent actuation and provide testability of the system.
b.
Testability of the ATWS/RPT System The inspectors verified the testability of the ATWS/RPT logic by reviewing the surveillance procedures.
The inspectors reviewed the most recent two sets (eight procedures)
of surveillance data for the ATWS actuation, the alternate rod insertion (ARI) and the RPT systems.
The procedures reviewed for surveillance data were:
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Plant Surveillance Procedure Number 7.4.3.4.1.1,
"ATWS-RPT-ARI, Actuation Reactor Level 2 Channels A
and C - CFT/CC,"
Plant Surveillance Procedure Number 7.4.3.4.1.2,
-"ATWS-RPT-ARI Actuation on Reactor Vessel Pressure High (ABC) CFT/CCg" Plant Surveillance Procedure Number 7.4.3.4.1.3,
"ATWS-RPT-ARI Actuation on Reactor Vessel Pressure High (BtlD) CFT/CC," and Plant Surveillance Procedure Number 7.4.3.4.1.4,
"ATWS-RPT-ARI Actuation Reactor Level 2 Channels B
and D
CFT/CC."
The surveillance procedures showed that the modified logic enabled each channel to be tested up to the recirculation pump breaker trip coil.
In addition, the inspectors identified that Section 5,
"Materials, Tools, and Test Equipment," of the surveillance procedures listed above,, did not clearly define the test equipment to be used.
The examples were:
(1)
Where a multimeter was required, the specification was only for "2 Calibrated Digital Multi Meters (DMMs)" (Procedure 7.4.3.4.1.2)
or "Two Calibrated Digital Voltmeters (DVM)" (Procedure 7.4.3.4.1.1).
The procedures did not sufficiently identify the equipment to insure it was within the specifications required to test the instrumentation, or to insure the accuracy was sufficient to )ustify the setpoint calculations.
(2)
For "Pressure Source and High Pressure Tubing and'ittings,"
Procedure 7.4.3.4.1.2 did not define the pressure range.
(3)
Procedure 7.4.3.4.1.1 called for a "Torque Wrench (Allen Head)," but did not specify the torque range nor the size of the Allen head.
The inspectors concluded that this finding was similar to that of the testing team inspection (Inspection 50-397/92-25).
The testing team found that surveillance procedures used to verify the setpoints of electrical protection assemblies in the reactor protection system did not include the accuracy requirements and data for the measuring equipment.
In response, the licensee was.
preparing a corrective action plan to address these
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finding I'
c.
Technical Specifications The inspector verified the subsequent changes, resulting from the modification, in Technical Specifications 3.3.4.1 and Table 3.3.4.1-1.
The changes were:
(1)
Provision that pump trip logic channels determined to be inoperable be placed in the tripped condition.
(2)
Provision that the remaining channels are verified operable within one hour when the number of operable channels per trip funct'ion is less than the minimum allowed.
(3)
Reduction from eight to six hours allowed for power.
operation with less than the minimum number of operable channels.
(4)
Revision of the minimum number of ATWS/RPT channels from one to two per trip function required,to be operable.
The inspectors concluded that the technical specification changes incorporated the hardware changes o'f the modified RPT system.
d.
Failure Trend Data The inspector reviewed the Equipment History Report, and it showed no failure trends.
The inspectors did not note any prolonged outages of the ATWS equipment resulting from hardware failure.
e.
The trending on the failure data of ATWS equipment appeared to be adequate in monitoring equipment performance.
The inspectors noted that the method of graphing the surveillance data allowed easy visual trending of the trip and reset points.
Walkdown On September 16, 1992, the inspectors performed a
walkdown of the ATWS equipment.
The inspectors did not observe any equipment deficiencies.
Conclusions-The 'inspectors concluded from the review of documents and the walkdown that:
o The implemented modification of the ATWS/RPT system
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appeared to be adequate.
Surveillance procedures for the'ATWS systems appeared to be adequate; however, certain procedural steps did not clearly define the test equipment to be used.
The licensee's ATWS equipment did not appear to'ave any prolonged outages resulting from hardware failure.
5.
ATWS Emer enc 0 eratin Procedures 37700 During a previous walkdown of the main control room, the inspector observed that fuses identified in the performance of emergency operating procedures had their locations and terminal points labeled.
Emergency Operating Procedure Number 5.1.2,
"Reactor Pressure Vessel Control ATWS," Revision 8, indicated, in the reactor power logic steps, that removal of fuses from the control panels was necessary in emergency support procedures 5.5.10,
"Overriding ARI Logic," and 5.5.11,
"Alternate Control Rod Insertions."
After examining the reactor power logic steps, the inspectors concluded that removal of fuses in accomplishing a step in an ATWS emergency operating procedure= 5.1.2 was acceptable when all the steps in operating control board switches failed to accomplish the intended function.
6.
Condensin Chamber 37701 On August 19, 1992, the staff issued Generic Letter 92-04,
"Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to
CFR 50.54(f)."
The generic letter requested plant specific information and the associated corrective actions from the licensee with respect to the effects of noncondensible gases on system operation.
The licensee was to respond to the generic letter by September 27, 1992.
At the time of the inspection, the preliminary assessment of the licensee was that the installed condensing chambers were in accordance with the General Electric guidelines,.and that there was no deviation from the generic requirements.
The licensee stated that WNP-2 will adhere to the resolution of the Boiling Water Reactor Owners'roup regarding condensing chambers.
General Electric document 22A3039,
"Process Instrumentation,"
Revision 1, specified the general guidelines for differential pressure instruments, dead-ended instrument lines connected to process systems inside containment, and requirements for reactor water level instrumentation.
Additional guidelines were provided
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in General Electric Nuclear Services Information Letter SIL Number 470,
"Reactor Water Level Mismatches,"
dated September 16, 1988, and its Supplement 1, "Reactor Water-Level Measurements,"
dated April 20, 1989.
.The general gui'delines stated that, the steam leg and reference leg of the condensing chamber should have a descending slope.
Condensed steam maintained a constant height of water in the reference leg, while the excess condensate returned to the reactor pressure vessel through a descending steam leg piping.
A reference leg, with a descending slope, from the bottom of the condensing chamber to 'the sensing differential pressure transmitter eliminated any high point for the accumulation of noncondensible gases.
The inspector reviewed selected isometric drawings'or the geometry of condensing chamber connections to the reactor vessel liquid level sensing nozzles N14 A, B, C, and D and concluded that the geometry of the installations appeared to conform to General Electric guidelines for installing condensing chambers.
7.
Desi n Chan e and Modifications 37701 The inspector selected Plant Modification Record Number 89-0299-022,
"Replace HPCS-PT-4,"
and the related Maintenance Work Request AR8180 for review.
This modification record was typical of the many transmitter modification packages.
Modification 89-0299-022 replaced a safety-related high pressure core spray transmitter which exhibited pressure fluctuations in the output signal when the input system pressure was steady.
The replacement transmitter was of a different make and model with equivalent qualification.
The inspector reviewed this plant modification record and the maintenance work request for adequate work instructions for removal, installation, leak-testing, and recalibration.
The inspector concluded that the modification 89-0299-022 appeared to have been adequately implemented.
8.
Closed Follow-u Item 50-397 92-01-14 To Tier Drawin Control and U date 92702 During the electrical distribution system fu'nctional inspection
{EDSFI, Inspection Report 50-397/92-,01),
the team identified a backlog of discrepancies between the fuses installed in the facility and the fuses the top tier drawings indicated were installed.
At the time of the EDSFI inspection, there were fifteen pending requests for technical services
{RFTSs) which related to fuses and top tier drawings that required engineering analysis to determine the fuse type and.rating.
That inspection found that the licensee had not promptly resolved the following. RFTSs:
90-10-070, 90-10-111, and 91-08-030.
The RFTSs were to determine if the
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controlled top tier drawings required revision or if the incorrect fuses were installed.
The licensee committed to complete the fifteen RFTSs related to the top tier drawings by July 1, 1992.
The inspector found that the licensee did not complete the resolution of the fifteen RFTSs until the first part of September 1992.
Basic Design Changes (BDCs) 55-1040-0F, 55-1040-0G, and 55-1040-OC resolved RFTSs 90-10-070, 90-10-111, and 91-08-030 respectively.
The inspector reviewed these three BDCs and considered them to be adequate in resolving the RFTSs.
The inspector also randomly selected three other RFTSs, 90-05-192, 90-07-116, and 90-10-013, from the group of the fifteen for review.
The inspector concluded that the licensee appeared to resolve these three RFTSs adequately with BDCs 55-1040-0E, 55-1040-0C, and 55-1040-OF respectively.
The inspector concluded that the licensee had made adequate progress in updating the top tier drawings in the fuse walkdown program.
This item is closed.
9.
Exit.Meetin The inspectors conducted an exit meeting on September 18, 1992, with'embers of the licensee staff as indicated in Section 1.
During the exit meeting, the inspectors summarized the scope of the inspection activities and reviewed the inspection findings as described in this report.
The licensee acknowledged the concerns identified in the repor Iyl