IR 05000334/1988025
| ML20195E586 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/31/1988 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20195E579 | List: |
| References | |
| 50-334-88-25, 50-412-88-19, GL-88-05, GL-88-5, IEIN-83-11, IEIN-84-83, IEIN-85-074, IEIN-85-74, IEIN-86-037, IEIN-86-37, NUDOCS 8811080310 | |
| Download: ML20195E586 (25) | |
Text
'
m
'
.
N.: .
-
gh
$X :
'
, .
1 U. S. NUCLEAR REGULATORY COMMISSION
'
Region I Report Nos.: 50-334/88-25 License Nos.: DPR-66 50-412/88-19 NPF-73 ,
'
Licensee: Duquesne Light Company '
One Oxford Center [
301 Grant Street !
i Pittsburgh, PA 15279 l i
Facility name: Beaver Valley Power Station, Units 1 and 2 !
Location: Shippingport, Pennsylvania i :
- Dates: September 1 - 30, 1988 '
! 1
'
Inspectors: J. E. Beall, Senior Resident Inspector !
S. M. Pindale, Resident Inspector P.[.Wilon,RactorEngineer,RegionI i
'
Approved by: /3 -
4450 /0/ N i Lowell E. Tripp,IChief Date Reactor Projects Section No. 3A
'
Division of Reactor Projects l
<
Inspection Summary: Combined Inspection Report Nos. 50-334/88-25 and !
50-412/88-19 for September 1 - 30, 198 *
l
- Areas Inspected
- Routine inspections by the resioent inspectors of licensee
'
actions on previous inspection findings, plant operations, security, radiolog-l ical controls, plant housekeeping and fire protection, maintenance, surveil-
- lance testing, implementation of boric acid leak detection procedures, emerg-
- ency diesel generator silencer supports, safety system functional evaluation l findings, process for temporary changes to plant equipment and storage battery i adequacy audit.
! Results: No violations were identifie Two unresolved items were opened i regarding (1) the development of long term corrective actions to resolve EDG
'
silencer support thermal growth concerns (Section 9) and (2) weaknesses in the i proces; for temporary changes to plant equipment (Section 11). The identifica-
.
tion of two findings by the licensee during a Unit 1 Quench Spray Safety System Functional Evaluation demonstrate a strong commitment towards ensuring system
,' operability and reliability (Section 10). Weaknesses were identified concern-
<
ing the surveillance testing program with respect to coordination and proced-
'
ural compitance (Section 6). Two previously open NRC items were closed.
i 05$ 00$88fy PNV
r
. .
.
TABLE OF CONTENTS Page Persons Contacted . . . . . . . . . . . . . . . . . . . . . . 1 Summary of Facility Activities ............... 1 Followup on Outstanding Items (92701) . . . . . . . . . . . . 1 Plant Operations ...................... 2 4.1 General (71707, 71710, 40700). . . . . . . . . . . . . . 2 4.2 Operations (71707, 93702) . . . . . . . . . . . . . . . . 3 4.3 Plant Security / Physical Protection (71881) ...... 6 4.4 Radiological Controls (71709) ............. 7 4.5 Plant Housekeeping and Fire Protection (71707) .... 7 Maintenance (62703) ... ......... ........ 7 Surveillance Testing (61726). . . . . . . . . . . . . . . . . 8 Implementation of Boric Acid Leak Detection Procedures (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Overpower Delta Temperature Actuations (71707, 62703) . . . . 10 8.1 OPDT Spikes Due to Excess Conservatism and Thermal Noise. . . . . . . . . . . . . . . . . . . . . . . . . 10 8.2 OPOT Spike that Results in Reactor Trips . . . . . . . . 11 Damaged EDG Silencer Supports (71707) . . . . . . . . . . . . 13 1 Safety System Functional Evaluation Findings (71707). . . . . 14 10.1 Inadequate Surveillance of QS System Flow Switches . . . 14 10.2 Licen see Identified Safety Related Cable Separation Violation. . . . . . . . . . . . . . . . . . . . . . , 15 1 Process For Temporary Changes To Plant Equipment (37702). . . 16 1 Storage Battery Adequacy Audit (71707) . . . . . . . . . . . 18 1 Unresolved Items. . . . . . . . . . . . . . . . . . . . . . . 19 14. Meeting s (30703, 30900) . . . . . . . . . . . . . . . . . . . 19 i
- _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _- . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _
. .
l
.
DETAILS
, Persons Contacted
During the report period, interviews and discussions were conducted with !
members of licensee management and staff as necessary to support inspec- [
tion activitie '
2. Summary of Facility Activities At the beginning of the inspection period, both Unit 1 and Unit 2 were !
operating 4t or near full power. Manual. load reductions to 50% power were !
initiated each weekend'during September on Unit 1 in a continuing effort i to extend core life in order to effect a later 7R refueling outage start !
dat Unit 2 maintained reactor power at about 98% power until i September 12 in an effort to prevent OPDT turbine runback spikes from ;
occurring (Section 8.1). Full power operation resumed on . September 1 !
On September 20, an automatic reactor trip occurred on one channel on Unit i 2 due to a large OPDT reactor trip spike that occurred when another OPDT *
reactor trip channel bistable was already in a tripped condition due to i surveillance testing (Section 8.2). The reactor was made critical later i that night and full power operation resumed on September 21, and continued j until September 30, when a manual load reduction to. 75% power was initi- i-ated in order to inspect main condenser water boxe f
!
3. Followup on Outstanding Items 7 r
The NRC Outstanding Items (0I) List was reviewed with cognizant licensee i personnel. Items selected by the inspector were subsequently reviewed i through discussions with licensee personnel, documentation reviews and field inspection to determine whether licensee actions specified in the OIs had been satisfactorily completed. The overall status of previously :
identified inspection findings was reviewed, and planned / completed licen- :
see actions were discussed for the items reported below: l
.
3.1 (Closed) Violation (50-334/87-07-04): Failure to report an engi- l neered safety feature (ESF) system actuation to the NRC within the 10 i CFR 50.72 prescribed time. Since issuance of the violation, the !
licensee provided separate correspondence to the NRC (letter dated !
January 6, 1988), identifying a concern that the 10 CFR 50.72 report- i ing requirements are overly conservative. The licensee's documented i position in that letter was that ESF actuations should only be l reported if the actuating circuit was a designated ESF actuating i circui For the interim, licensee personnel have been instructed to '
continue to conservatively report all ESF actuation By letter I dated June 27, 1988, the NRC responded to the licensee's letter which l l
!
i !
'
t I t i
f
.- - - _ - -
~. .
.
specified that, since the current regulations (10 CFR 50.72) do not differentiate between partial or full actuations or whether the situ-ation was caused by an ESF circuit, such ESF actuations should continue to be reported under the current regulations. The licensee acknowledged the NRC letter, and is continuing its current ESF reporting practic The inspector reviewed the licensee's current -
ESF reporting program and found no deficiencies. This program and i its implementation will continue to be monitored during routine i resident inspections. This item is close !
!
3.2 (Closed) Unresolved Item (50-334/87-02-02): Review procurement con- !
trol program to assure that Category I components returned to the l warehouse are properly stored and maintained. All materials that t were returned to the warehouse / storeroom w0re maintained in level A i storage (highest storage level; controlled temperature, humidity), :
'
however, a large amount of materials had accumulated and were not restocked. The licensee subsequently identified all Category I ,
equipment and the associated special storage and maintenance require- [
ments. As of August 1,1988, the effort had been completed and the ;
large backlog of materials for storage had been eliminate The '
licensee also revised procedures to ensure that returned Category I .
material is handled on a routine basis, thereby minimizing the time between material return and reshelving. The inspector reviewed the affected procedures and toured the material return areas. No defici- l encies were identifie This item is close . Plant Operations I 4.1 General ..
Inspection tours of the following accessible plant areas were con-ducted during both day and night shif ts with respect to Technical Specification (TS) compliance, housekeeping and cleanliness, ' ire protection, radiation control, physical security / plant protection and operational / maintenance administrative control Control Room -- Safeguard Areas
-- Auxiliary Building -- Service Building
-- Switchgear Area -- Diesel Generator Buildings
-- Access Control Points -- Containment Penetration Areas
-- Protected Area fence Line -- Yard Area
-- Turbine Building -- Intake Structure
_ _ _ _ _
_
. .
.
4. ESF Walkdown The operability of selected engineered safety features sys-tems was verified by performing detailed walkdowns of the accessible portions of the system The inspectors con-firmed that system components were in the required align-ments, instrumentation was valved-in with appropriate cali-bration dates, as-built prints reflected the as-installed systems and the overall conditions observed were satisfac-to r The systems inspected during this period include the Emergency Olesel Generator, Safety Injection and Recircula-tion Spray systems. No concerns were identifie . Onsite Safety Committee The inspector attended a Unit 2 Onsite Safety Committee (OSC) meeting on September Technical Specification 6.5.1 member attendance requirements were met. The agenda included procedure, incident report and design change pack-age reviews. In general, member participation for safety issues was adequat The inspector will continue to monitor the adequacy and ef fectiveness of OSC meetings and discussions during future inspection No significant concerns were identifie .2 Operations During the course of the inspection, discussions were conducted with operators concerning knowledge of recent changes to procedures, facility configuration and plant condition During plant tours, logs and records were reviewed to determine if entries were properly made, and that equipment status / deficiencies were identified and com-municated. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, unit off-normal and draft incident report The inspector verified adherence to approved procedures for ongoing activities observed. Shift turnovers were witnessed and staffing requirements confirmed. Inspector com-ments or questions resulting from these reviews were resolved by licensee personnel. In addition, inspections were conducted during backshif ts and weekends on 9/3, 3:00 pm - 9:30 pm; 9/17, 8:00 am -
4:00 pm; 9/29, 3:00 am - 7:00 a . Inadvertent Start of Steam Driven AFW Pump Oue to Valve failure On September 23, with Unit 1 operating at full power, the steam driven auxiliary feedwater (AFW) pump inadvertently started due to a failed diaphragm in one of the two paral-1el steam admission valves (105A). Plant operators deter-mined that the pump was running af ter receiving a control
.
. .
I
.
room steam generator level deviation alarm and observing about a five percent increase in all three SG water level The pump remained running while troubleshooting activities were in progress since 105A could not be closed due to the failed diaphragm. The pump was subsequently shutdown by closing an upstream steam isolation valve (MOV-105) and was declared inoperable at 8:54 The valve repair (replaced diaphragm) was completed by 2:30 a.m. on September 2 See Section 4.2.2 for a discussion of the post maintenance testing activities associated with this repai The AW system start logic is such that if a start demand signal is present for the steam driven pump and a predeter-mined pump discharge pressure is not achieved in ten seconds, then the motor driven A W pump associated with the opened steam admission valve will automatically start. The system configuration consists of one common steam isolation valve (MOV-105) which branches into two parallel steam admission valves (105A and 1058). An open signal from the control room benchboard or an automatic start signal to the pump constitute start demand signals. Neither of the above start demand signals were present, however, 105A was actually open due to the diaphragm failure and the steam driven pump was running. Therefore, the 3A motor driven AN pump did not automatically start when upstream steam isolation valve MOV-105 was closed and the steam d iven A W pump discharge pressure and flow were stoppe The licensee reviewed the maintenance work request (MWR)
database to determine when the diaphragms for valves 105A and 105B were last replaced, however, no such MWRs were foun iherefore, the licensee initiated an effort to replace the diaphrtgm for 1058 as a preventive measur The failure of the diaphragm was suspected to be due to aging while in an environment of high temperatures. These ,
valves are located in the lower level of the main steam
'
valve area where temperatures are at about 100 degrees during normal plant operatio The inspector questioned whether these or similar valves are included in preventive maintenance (PM) activille The licensee stated that several valvos that are either physically located in areas where adverse environmental '
conditions or had otherwise failed in service periodically are included in the PM program. The licensee further stated that an investigation for AN system valves 105A and 105B will be initiated to determine if periodic diaphragm replacement is appropriat The inspector will review the results of the licensee's investigation, as well as a review of other similar valves not included in the PM program, during a subsequent inspectio _
- - - - - - - - - - - . - _ _ _ - - _ _ _ _ _ - - - - _ _ _ _ _ _ _ _
-. .
.
5'
4.2.2 Automatic Start of Motor Driven AN pump Due to Personnel Error During the performance of the steam driven AN pump opera-tions surveillance test (OST) to verify proper maintenance on 105A,'an automatic start of the motor driven AFV pump occurred on September 24 while the plant was at about 52%
power (load reduced to extend core life). During the OST, a higher than nnrmal pump discharge pressure was observed by plant operators, as well as an unexpected full closure of the A N pump recirculation valve. The steam driven pump was manually started by opening steam admission valve 105A at the start of the OST. Dua to the abnormal response of the steam driven pump and the prompt action -required by such a response, an operator immediately shutdown the oper-ating pump by closing the steam isolation velve (MOV-105)
in the common line upstream of the two steam admission valves. As discussed in Section 4.2.1, closing the steam isolation valve and leaving 105A and its associated control switch in the cpen position, produces a start demand signal. Since the pump was actually shutdown, pump dis-charge pressure could not achieve its required value and the 3A motor driven AW pump automatically started as per design. AW pump 3A was later shutdown and returned to normal system alignmen The licensee attributed the primary cause for this event to be cognitive personnel error in that the licensed opera-tor failed to realize AW system automatic start logi Additional causes include (1) tasks not covered by proced-ures (OST did not address rapid pump shutdown), and (2) operators are trained on the simulator to shutdown the steam driven pump using MOV-105 during accident conditica MOV-105 is used during accident conditions because an auto-matic AW system actuation is likely to have occurred, and 105A and 105B will immediately reopen automatically if an automatic signal had been generated due to the signal lock-in feature. This event was discussed with plant operators, with emphasis on the AN system automatic start logi Additionally, the licensee plans to enhance operator train-ing (both classroom and simulator) in this are The inspector will review the training program improvements and its effectiveness during a subsequent routine inspectio .
. .
.
The licensee suspected a fault in the recirculation valve flow switch to account for the valve closure and the high discharge pressure. Subsequent troubleshooting activities, however, did not identify any deficiencies and the flow switch and recirculation valve were found to function pro- "
perly. After exercising the recirculation valve and flow -
switch, the steam driven cump OST was again performed to verify operability, however, additional operational prob- ;
1 ems were noted after the pump was manually started, including higher than expected pump RPM. Fo11cwup investi- r gation concluded that the pump governor was not operating !
properl The governor was subsequently replaced on September 2 The pump was then tested satisfactoril returned to service and declared operable at 4:00 p.m. The i apparently failed governor is suspected to be the cause for ;
the recirculation valve closure, and was sent to the vendor for testing and root cause analysi The inspector will review the results of the vendor analysis and the licen-see's associated corrective and/or preventive measures '
during a future inspection, i 4.3 Plant Security / Physical Protection Implementation of the Physical Security Plan was observed in various plant areas with regard to the following:
--
Protected Area and Vital Area barriers were well maintained and not compromised;
--
Isolation zones were clear; !
i
--
Personnel and vehicles entering and packages being delivered to ,
- he Protected Area were properly searched and access control was
. i in accordance with approved licensee procedures;
,
--
Persons granted access to the site were badged to indicate [
>
whether they have unescorted access or escorted authorization; ;
--
Security access controls to Vital Areas were being maintained i j and that persons in Vital Areas were properly authorize [
l --
Security posts were adequately staffed and equipped, security f
! personnel were alert and knowledgeable regarding position !
requirements, and that written procedures wera available; and !
- :
,
--
Adequate illumination was maintaine l No deficiencies were identifie !
! !
t I
i r
.
!
'
l
_ _ _ _ _ _ _ _ - _
.
.
.
4.4 Radiological Controls Posting and control of radiation and high radiation areas were inspected. Radiation Work Permit compliance and use of personnel monitoring devices were checked. Conditions of step-off pads, dis-posal of protective clothing, radiation control job coverage, area monitor operability and calibration (portable and permanent) and per-sonnel frisking were observed on a sampling basis. No concerns were identifie .5 Plant Housekeeping and Fire Protection Plant housekeeping conditions, including general cleanliness condi-tions and control and storage of flammable material and other poten-tial safety hazards, were observed in various areas during plant tours. Maintenance of fire barriers, fire barrier penetrations, and verification of posted fire watches in these areas were also observed. The inspector conducted detailed walkdowns of the accessi-ble areas of both Unit 1 and Unit 2. Overall, housekeeping was found to be adequate for both unit Individual deficiencies were identi-fied to the licensee for resolutio . Maintenance The inspector reviewed selected maintenance activities to assure that:
--
the activity did not violate Technical Specification Limiting Condi-tions for Operation and that redundant components were operable;
--
required approvals and releases had been obtained prior to commencing work;
--
procedures used for the task were adequate and work was within the skills of the trade;
--
activities were accomplished by qualified personnel;
--
where necessary, radiological and fire preventive controls were ade-quate and implemented;
--
QC hold points were established where required, and observed;
--
equipment was properly tested and returned to servic ** .
i O
Maintenance activities reviewed included:
MWR 883357 Calibrate QS-P-1A Discharge Flow Switch MWR 883358 Calibrate QS-P-1B Discharge Flow Switch MWR 889404 Replace OPDT Lead / Lag Card No deficiencies were identifie . Surveillance Testing The inspectors witnessed / reviewed selected surveillance tests to determine whether properly approved procedures were in use, details were adequate, test instrumentation was properly calibrated and used, Technical Specifi-cations were satisfied, testing was performed by qualified personnel and test results satisfied acceptance criteria or were properly dispositione The following surveillance testing activities were reviewed:
OST 1.3 Weekly Motor Driven Fire Pump Operation Tes OST 2.1.12E Safeguards Protection System Train B Miscellaneous Go Tes OST 2. Nuclear Power Range Channel Functional Test OST 2.2 Motor Driven Auxiliary Feed Pump Tes MSP-6.3 Reactor Coolant Temperature Loop 2RCS-T412 Delta T - T-avg Protection Channel I Calibration During the performance of OST 2.24.3 the following discrepancies were noted: Acceptance criterion 1.C requires that the auxiliary feedwater (AFW)
pump be started from the control roo The AFW pump was remotely started from a test switch located outside the control room during the performance of OST 2.1.12E and then OST 2.24.3 was entered. After discussions with the licensee, the AFU pump was started from the control room to satisfy that acceptance criterion,
' The surveillance requires a 0 to 600 F or equivalent contact pyro-meter be used for measuring pump casing temperature. A minus 40 to 1999 F pyrometer was utilized. Several operators and other members of the operations staf f were asked what constituted an acceptable ,
equivalent test instrument, however, it was not apparent that they were provided with specific guidance on the definition of equiva-1ency. The test pyrometer that was used for this test appeared to be adequate for its intended purpos s s
9 The OST requires that constant communications be maintained between the control room and the AFW pump roam while the pump dischargs valve is closed during the test. Direct coraunication was not maintained during the majority of time the r mpa discharge valve was close Step 17, requires that, after the AFW pump is stopped, the differen-tial pressure across the pump discharge v11ve be equalized by momen-tarily opening an upstream drain valve (2FWE*334). The next step (step 18) requires that the drain valve be verified closed and its pipe cap reinstalle These two steps were not performed and the pump discharge valve was subsequently opened without equalizing the pressure. Step 18 was signed off as completed even though it was not performe Although not specifically covered during this inspection period, an ESF actuation occurred on October 3 during the performance of an AFW system OS Specifically, during OST 1.24.4, an automatic stsrt of motor driven AFW pump 3A occurre The root cause for this actuation appeared to be due to coordination and communication deficiencies. During an independent review of this event, the inspector determined that the required constant communication may not have been properly implemented. A 10 second time delay is allowed for operator response to close certain valves af ter the ,
steam driven AFW pump is secured before an automatic motor driven pump i start will occu This short time period required extensive operator i coordination and communication. This event will be described in more detail in the next routine resident inspection report, r
The apparently deficient coordination and communication associated with the performance of OST 1.24.4 on October 3, along with the discrepancies described above (associated with OST 2.24.3) indicate weaknesses in sur-veillance testing activitie These concerns were brought to the licen-see's attention, who acknowledged the inspector's comments. Specific corrective actions for the above will be reviewed during future routine inspection i 7. Implementation of Boric Acid leak Detection Procedures On March 17, 1988, the NRC issued Generic Letter (GL) 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Boundary Components in PWR Plants." The f letter requested that licensees implement procedures to address the cor-rosive effects of reactor coolant system leakage at less than Technical ;
Specification limits. By letter dated May 31, 1988, the licensee respond- i ed to the letter, stating that procedures were already in place as a re- l sult of previous NRC and industry correspondence, and would be enhanced to +
address the concerns identified in GL 88-05. The procedures are intended to emphasize personnel awareness, good housekeeping practices, e.ttention '
to all leaks, importance of proper reassembly of bolting closures, system walkdowns, and documentation and tracking of leaks and repairs. On the ;
basis of such commitment, by letter dated August 24, 1988, the NRC con-cluded that the licensee had fulfilled the requirements of the GL, while stipulating that the licensee should maintain auditable records of the program and results obtained from implementation of the program, i
I i
!
-__ .
,
.
The inspector reviewed the licensee's procedures and found that these pro-cedures have addressed the concerns expressed in GL 88-05. The procedures also met the licensee's commitment in its letter dated May 31, 1988. The licensee's records of the program and results obtained under these proced-ures will be the subject of future routine inspections. No concerns were identifie . Overpower Delta Temperature Actuations 8.1 OPDT Spikes Due to Excess Conservatism and Thermal Noise Since initial plant startup of Unit 2, the licensee has implemented a NSSS vendor (Westinghouse) ricommendation regarding setpoints of selected protection feature Included are the overpower delta tem-perature (0PDT) and overtemperature delta temperature (OTDT) auto-matic turbine runback and reactor trip signal The setpoints for the above four automatic reactor protection system (RPS) features were set and maintained at values that were three percent more con-servative than the Technical Specification required values per vendor recommendation, pending the collection and analysis of data to con-firm design predictions and calculations. Data was to be collected in a full power reactor trip during startup testing to confirm the response time of the reactor coolant system resistance thermocouple detectors (RTOs). However, all of the necessary data was not col-1ected during initial startup testin Therefore, the necessary channels were connected to strip chart recorders so that the data could be obtained upon an unplanned plant tri That data was sub-sequently obtained during the July 27, 1988 reactor trip (see NRC Inspection Report 50-412/88-18), and analyses were ongoing during this ir,spection period so that the three percent extra conservatism could be partially or totally eliminate For the past several months, the licensee was monitoring and evalua-ting the full power delta temperature (DT) values. The results of the study identified that the current values were conservative and the licensee subsequently calculated and confirmed more realistic full power DT values for each of the three reactor coolant system loops. Therefore, revised calibration values were used to recali-beate the channels using the new full power setpoints. On August 30, channel "A" was recalibrated. Shortly thereaf ter, momentary spikes on that channel were received in the control room such that a one out of three OPDT turbine runback signal was generated. A two out of three coincidence is required to initiate the runback. The licensee then suspended the channel recalibration evolutions before channels B or C were affected. On August 31, special operating instructions were provided to plant operators to maintain reactor power at approx-imately 98% to prevent OPOT turbine runback spikes f rom occurrin .
O e =
s O
.
This action appeared to be effective for the short term in that addi-tional spikes did not occur. While maintcining the reactor power restriction orders in effect, the licensee proceeded to recalibrate channels B and C to incorporate the new full power OT values. This was performed without further inciden Concurrent with the above efforts, the licensee and Westinghouse ton-tinued efforts to evaluate the recent data obtained durica the July 27 Unit 2 reactor trip. Westinghouse performed an interi] real-uation, which concluded that 1.5% of the extra RPS setpoint cct>erva-tism could be eliminated. Consequently, on September 10, the licen-see performed adjustments and calibrations of the OPDT and OTOT auto-matic turbine runback and reactor trip signal setpoints. This action allowed the licensee to relax the previously invoked 94% reactor power restrictio Licensee management then authorized full power operation on September 12, and full power was reached on September 13. No additional spikes were received due to the above
,
concerns. See section 8.2 for apparently unrelated subsequent OPOT spikes that occurred beginning on September 20. No deficiencies were identified concerning the above licensee activities. The inspector will review the final resolution of the RPS setpoint values during a future inspection, 8.2 OPDT Spike That Results in Reactor Trip On September 19, the licensee performed an 18-month surveillance test (channel calibration) on the OPDT Loop A circuit, which identified that the low limit portion of the lead / lag card was faulty, such that acceptable values could not be obtained upon simulating a decreasing average temperature signal. The licens e estatlished that the lead /
lag card (hereafter referred to as card A) had been in service for at least one year. Card A was then replaced with a spare card (Card 8)
that had been previously removed from operation and repaired. Card B was operationally calibrated and tested per the surveillance test, which was completed later on September 1 On September 20, while at full power, plant operators initiated oper-ational surveillance test (OST) No. 2.2.1, Nuclear Power Range Chan-nel Functional Test. While testing channel PRN43, plant operators manually tripped the loop C OPDT rod stop/ turbine runback and reactor trip signals by placing the associated bistables in the test
!
positio Shortly thereaf ter (about two minutes), a momentary 0PDT setpoint spike occurred on another channel (Loop A), and the two out of three reactor trip logic was satisfied. The setpoint spiked to the point that it momentarily was less than channel OT, thereby actuating l a reactor trip signal on that channe The unit automatically tripped at 3:06 a.m. and was subsequently staoilized in Mode 3 (Hot Standby). Plant response to tne transient was as expected and was
'
verified by the performance of a Unit 2 specific computerized trip response analysis. The licensee notified the NRC of this event in accordance with the reporting raquirements of 10 CFR 50.7 !
!
, _ .
.
_ _ _____________
s .
O
.
The plant was maintained in Mode 3 (Hot Standby) while technicians were troubleshooting OPDT channel A. Laad/ lag Card B remai. 'd in plare while full power signals were inserted into the loop A cir-cuitry to simulate normal operating conditions as portions of the loop were menitored ' ia strip chart recorder During the activ-ities, a spike v;as ob 'rved on the output of the card, but not on the input, indicating ar iarent fault within Card B. The licensee had since identified anu stred the fault on tard A, which previously could not be calibrati, iroperly. The card was satisfactorily bench tested and was used to replace Card 8 on September 20. Card A was calibrated after installation into the system and it was satisfac-torily operationally teste Plant startup then commenced. The reactor was critical at 7:50 p.m. on Septamber 20 and Mode 1 (Power Operation) was entered at 8:28 Full power was reached on September 2 The licensee instituted restrictions on the perform-
- ance of OSTs that would intentionally trip OPDT bistables so as to
! provent unwarted automatic actuations sheuld another channel experi-I ence a spik This was to be in effect while monitoring of the i affected channels could confirm the adequacy of the licensee's cor-l rective action The licensee continued to monitor loop A by connecting a strip chart recc-der to the OPDT circuitry in the associated protection cabine Addiuonally, plant operators selected loop A to be indicat.,d on the control room DT, OTOT and OPDT recorder (one of the thre'. loops can be monitored at any one time). At 10:30 p.m. on Septe".oer 21, both recordrers monitored a spike on the channel A OPDT sr cpoint signa Preliminary troubleshorting activities identified that the spike was the result of some type of problem with the recently replaced lead /
lag card (Card N. Technicians then monitored several points on the affected card to localize tne faul On the followin2 day (September 22), another similar spike occurred at 7:41 Upon review of the collected data, technicians were not yet able to iden-tify a specific componen+ as being the cause for the spike, although the data did confirm that the spike was caused by Card A. The loop remained instrumented (with decreased band monitored for the recorder) with the plant at full power, and no spikes occurred u.*itil September 25, Technicians were able to identify a failed input gain stage component on that card, which was subsequently determined to be the cause for the intermit'.ent spike Thtt card was then replaced with a new card received from warehouse (Card C), which was cali-brated and tested satisfactoril The Loop A OPDT circuitry was monitored for several days, ano no spikes were observed on the con-nected equipment, nor from control room indications or alarm )
_ - - -
~ .
e e i 13 l Followup monitoring of the loop A OPDT circuit continu1d in an effort to confirm that the lead / lag card was not being affected externall The restriction regarding the performance of other OSTs and M0' that require tripping OPDT bistables was lif ted, however, provis1L.is for
"soak times" following the performance of such tests or maintenance was required. Specifically, following entry into any OPDT loop for calibration purposes, other loop bistables were not to be manipulated (manually tripped) for any reason for a period of eight hours )f operation with no channel spike Additicnally, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of opera-tion with no channel spikes was required following loop corrective maintenance before another channel could be tested. These restric-tions were to provide assurance that reactor trips due to inadvertent channel spikes following channel entry are prevente Regarding Card B (card which caused 9/20 reactor trip), no component failures have been identified to date which could have accounted for the spik The licensee maintained the card on a test device for several days, simulated several environmental conditions, and mon-
'
itored the card, however, no spikes were observed. The licensee plans to perform additional testing of the card in an attempt to
"
explain the in-service spike that was experienced on September 2 For the interim, card B has been tagged and identified as one that has caused spikes, and will not be permitted to be reinstalled in the system. The licensee is also reviewing associated OSTr. to determine whether bistables are required to be tripped for specific evolution It appeared that, in some cases, tripping the OPDT bistables was an overly conservative step, ana the test could be performed without tripping them. The licensee has contacted the Wolf Creek site which has experienced similar difficulties. Based on subsequent evaluation of the above experience, the licensee has concluded that the trouble-j shooting and testing procedures as supplied by the vendor are not
- adequate (incomplete) for detecting failures in lead / lag cards that
,
can lead to spiking. In the interim and until long term improvements 4 can be made in troubleshooting procedures, the licensee will replace
- any such cards that develop problems witn new card The inspector will review the licensee's Card B testing activities and routine operation of the OPDT circuit during routine in pections.
,
9. Damaged Emergency Diesel Generator Silencer Supports During a walkdown of the Unit 2 Emergency Diesel Generators (EDGs), the inspector observed that the supports for the silen:ers of both diesels had been damaged. The silencers are large (29 feet long, 8 feet in diameter)
cylinders which are installed horizontally on tha second level of the EDG building. Each is installed on two wide (about eight feet) steel supports which are mounted on steel pads. The steel pads rest on concrete niers which are six inches tall, one foot wide and about eight feet ac e Each support structure (two per silencer) is anchored to the floor Sy four bolts (3/4 inches thick and two feet long) which go through b, ex in the steel pads and are embedded in the concret _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - - _ _
s .
o e
The support structure at each silencer inlet is fixed while the support at the outlet end is free to slide in the silencer's length-wiss direc-tion. The sliding design is necessary because each silencer thtrmally
"grows" about one inch in the lengthwise direction as it heats up to about 900 F during EDG operatio The support design does not, huwever, allow for growth in the silencer's width direction. Th support was installed well insulated such that the support temperature approached that of the silencer and resulted in thermal growth of about 1/2 inc The thermal growth caused the top of the bolts to move with the steel pad although embedded in the concrete. The bolt movement bent the bolts and broke the corners of the concrete piers. Silencer cooldown after the EDG was secured caused contraction of the steel pads wfiich bent tha bolts bac '.a c h period of EDG operattun, therefore, cycled the bolts and further damaged the concrete pier The inspector identified the damage to licensee management and questioned the integrity and ductility of the bolts following more than two years of EDG operatio The silencers are seismically supported so that an earth-quake cannot damage the EDG exhaust piping and render the EDG inoperable due to heat and other EQ problems. The licensee performed a UT inspection of the bolts and found that one bolt was ,otentially cracked near the pier-slab interface (about five inches from the head). The licensee's preliminary analysis indicated that the EOG silencer was still adequately supported, but committed to additional UT examinations following each future thermal cycle. The support insulation was als' removed in an at-tempt to reduce the support temperature rise and thermal growth during future EDb operatio These measures are considered to be interim in nature and the licensee indicated that the silencer supports would be repaired in a future outage, pending approval of a suitable design modifi-cation. This item is Unresolved (50-412/88-19-01).
10. Safety System Functional Evaluation Findings During this 1.ispection period, the licensee was performing a Safety System Functional Evaluation (SSFE) for the Unit 1 Quench Spray (QS) System, The SSFE is a broad-based technical audit involving significant licensee resources and is expected to provide enhanced assurance of system opcra-bility and reliabilit Two significant findings have been identified during this inspection period and are further discussed in thq following sections. The final resolution of these items will be reviewed during subsequent inspection .1 Inadequate Surveillance of QS System Flow Switches Quench Spray (QS) System flow switches FIS-QS101A and FIS-0S1018 monitor the discharge flow of the A and B QS pumps, respectively. A low flow alarm is provided in the control room, and the low flow signal (less than 8C0 gn) automatically trips the four chemical addition pumps. The SSFE identified that during testing of the QS
____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __. _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- .
.
.
15 >
,
System to conform with Technical Specification surveillance require-ments, a jumper is installed to simulate greater than 800 gpm QS System flow to allow the chemical addition pumps to run during the test. The function of the two flow switches was not verified during i the operations surveillance test (OST). Additionally, there were calibration procedures to calibrate the flow switches, nor were they identified in any calibration schedul The licensee subsequently l
determined that both switches were last calibrated in August, 1980, :
with no functional check of the annunciators or interlock function *
The licensee determined that the potential existed for a system oper- '
ability concern in that the failure of the flow switches to operate could possibly disable the chemical addition pumps upon receipt of a containment isolation (phase B) signal. The chemical addition pumps
_ are designed to provide metered sodium hydroxide into the QS pump
] suctien Sodium hydroxide is added to the system to improve removal of radioactive iodine from the containment atmosphere following an acciden The SSFE com11ttee documented the finding in a deficiency report dated September 2 Recommendatiers as documented in the report included (1) initiating a surveillance procedure to calibrate the affected flow switches, (2) functionally testing the interlocks, and (3) reviewing other surveillance procedures to determine if this is a programmatic proble J Upon notification of this event, maintenance work requests were initiated to complete a calibration and functional check for each of the flow switches and interlock Both were out of specification as found, however, the licensee evaluated the as found data and con-
<
cluded that the flow switches could have performed their intended i
safety function using the as found values. The flow switch interlock
,
functional tests were performed satisfactorily. FIS-QS101A was sat-
'
isfactorily calibrated to within specification, however, the reset
value for FIS-QS1018 could not be calibrated to a value within the l acceptance criteria. An evaluation was subsequently performed, which de: ermined that although the as-lef t reset value was acceotable, a more conservative value should be used. The licensee plans to recal-ibrate this instrument. The inspector will review this activity and l the results of the licensee's review of other potentially affected surveillance procedures during a future routine inspectio .2 Licensee Identified Safety Related Cable Separation Violation
, Two 05 system chemical addition tank level transmitters (LT-QS101A and LT-QS101B) provide control room annunciation for low and extreme-
'
low chemical addition tank leve They also provide an interlock function such that the chemical addition pumps and associated dis-charge valves are automatically stopped and closed, respectivel The instruments are designated in the UFSAR as Category 1 safety-a related equipment and are therefore required to be electrically
!
4 . .-
-
- ~
_
.
~
'
- e i
,, , _ _ _ . - - . _ - - - - - - . -
_ ________ _ _ __-__ __. _ - . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
~ .
o e
redundant and independen During the SSFE study, the licensee identified that the above criteria were not satisfied in that (1) the associated cables were not identified as safety-related and (2) the cables were routed in a common non-safety grade raceway with other non-Class 1E cables. This was an apparent violation of Specification BVS-3001, Endorsement Specification for Installation and Identifica-tion of Electrical Cables for Continuing Service Task The licensee reviewed this finding for its immediate impact on the plant and determined that immediate correction was appropriate. Spec-Ifically, the licensee electrically defeated the interlock function of LT-QS101A and LT-QS101B to preclude a common mode failure from prematurely isolating the chemical addition tank. Defeating the interlock would, however, prevent an automatic isolation when the extreme low tank level was reached during an accident. This could damage the chemical addition pumps if manual operator action was not taken, but by that time, the safety function of those pumps would have been fulfille A long term corrective action plan is being pursued by the license This problem is similar to the inadequate cable separation violation identified in NRC Combined Inspection Report Nos. 50-334/88-22 and 50-412/88-16. Identification and cor-rective actions for similar cable separation deficiencies is planned to continue until December 16, 198 Engineering reviews into this event are currently ongoing. The safety significance of this event and its resolution will be reviewed during a subsequent inspectio . Process for Temporary Changes To Plant Equipment Federal regulations authorize licensees to make changes in the facility and procedures un'iess it involves a change to the Technical Specifications
! or an unreviewed safety question (10 CFR 50.59). The regulations also give guidance concerning what constitutes an unreviewed safety question and what reports and records are required as documentatio The licensee uses the 50.59 process in making temporary changes to plant equipmen :
Permanent plant modifications also receive a 50.59 review but are handled differently in that the licensee's engineering department uses a design change package (DCP) to implement the modificatio The inspector reviewed the licensee's administrative procedures, safety evaluation training records and selected temporary modification package Station procedures require safety evaluation training for those involved in the development and review of permanent modifications; there is no similar requirement for those who prepare safety evaluations for temporary modifications. The inspector's review of the licensee's training records indicated that many of the temporary modification safety evaluations were prepared and checked by individuals who had not received ths above train-in The inspector also noted that three members of the Onsite Safety Committee (OSC) which reviews the safety evaluations did not appear to have received the training which, as OSC members, is a procedural requirement.
l l
'
l
_ _ _ _ _ _ _ .
. .
.
The inspector reviewed selected safety evaluations for permanent modifica-tions (contained in the DCPs) and found them to be of good quality with adequate justification for stated conclusion The inspector also re-viewed the safety evaluations for a sample of temporary modifications and found them to be weaker than those in the DCPs. The temporary modifica-tion safety evaluations contained a much lower level of detail such that, in some instances, they did not appear to contain sufficient information to justify the stated conclusions. For example, one temporary modifica-tion was for a field wiring change to a level switch (2FNC-LS105) in the l Nuclear Spent Fuel Cooling and Cleanup System. The associated safety evaluation contained "no" answers to the ke/ screening questions without any justificatio In other cases, the only information provided for the conclusion that an unreviewed safety question did not exist was that the equipment being modified was not safety related. Modifications to non-safety related systems and structures may impact safety related equipmen Also, the support systems for some safety related equipment are not them-selves safety relate l One example of the above concern involves a temporary modification which involved the air dryer (21A) in the Emergency Diesel Generator Air Start-ing System (EDGASS) for the 2-1 diesel. The air dryer feature was re-quired by the NRC staff, as documented in NUREG 1057 "Beaver Valley 2 Safety Evaluation Report" (Section 9.5.6). Operating experience presented in NUREG/CR-0660 had shown that water accumulation problems in starting air systems had been one of the most frequent causes of diesel start fail- '
urcs and that the problem could occur in air-over piston started diesel ;
engine The licensee committed to install and maintain the dryers prior to fuel loa Air dryer 21A failed in May,1988 and was still out of service at the end of the inspection period over four months later. A temporary modification was performed which bypassed the failed d rye The associated safety evaluation noted that the dryer was not safety related and that the air flask would be periodically blown down to check for moisture. The BV-2 SER (NUREG 1057) noted that periodic blowdown of the air receivers would ,
not provide dry air and that water condensation could occur during airflow l to the engin The safety evaluation does not address this concer The inspector also noted that there was no apparent time limit on tempor- ;
ary modification One current temporary modification to the feedwater <
control system was installed on May 25, 1982. Another temporary change which modified LSHH-06107Al was installed in 1984 as was a modification which lif ted the control signal to one group of pressurizer heater It was not clear as to when a temporary modification must be replaced by a permanent plant change. In the interim, plant information documents such [
as system descriptions and prints may not fully and accurately represent t the in situ hardware and performance characteristic . .
c e
The above problems regarding training inconsistency, safety evaluation content and temporary modification duration are considered to be weak-nesses. The temporary modification allowing operation of the air com-pressor without the air dryer in service may represent an unreviewed safety question in that it may increase the possibility of malfunction of equipment important to safety. This issue is unresolved pending further evaluation by NRC of p evious assessment (s) (50-334/88-25-01 and 50-412/
88-19-02).
1 Storage Battery Adequacy Audit The inspector conducted an audit of onsite storage batteries at both Units 1 and 2 during this period. The purpose of this audit was to review the licensee's program to assure that storage batteries will, in accordance i with the current licensing basis, remain operabl This audit was per-formed in accordance with NRC Region I Temporary Instruction (RTI) 87-07, Storage Battery Adequacy Audit. To assist the licensee in effectively i providing information to the inspector, Attachment 2 (Storage Battery
- Inspection Sample) of RTI 87-07 was provided to the licensee. The attach-ment is also provided as an enclosure to the current inspection report cover letter. This audit will be performed in two phases and will be completed during the next routine resident inspectio The inspector reviewed battery documentation and found that seismic life-time and qualification had been e sual i shed, The seismic lifetime was l established by both analysis and testing of actual equipment that had been artifically aged per the appropriate IEEE Standard Specific require-ments for maintaining the seismic qualification had also been established by the licensee. Criteria for assuring battery and rack seismic qualifi-cation are defined and used for periodic inspections and cell replacement Battery electrical sizing is confirmed to be sufficient to handle the DC load profiles with suitable margin by testing and calculations, which are i
periodically updated to reflect system changes. All safety related bat-teries were ordered with a 24*. margin for aging and a 19*. ma rg i n for operating down to a temperature of 50 Each battery room is physically independent. Provisions for assuring ade- !
'
quate battery ventilation during normal operation, outages, charging and discharge are administratively controlle Periodic surveillance tests also require that the ventilation system flow is checked. Administrative controls are in place to ensure that ventilation enclosures are not obstructed such that ventilation flow could be impede The inspector conducted detailed tours of all safety related battery room No significant deficiencies itere identifie .
e
.
13. Unresolved Items Unresolved items are matters about which more information is required in order to .sscertain whether they are acceptable items, violations or deviation Unresolved items are discussed in sections 9 and 11 .
14. Meetines Periodic meetings were held with senior facility management during the course of this inspection to discuss the inspection scope and finding A summary of inspection findings was further discussed with the licensee at the conclusion of the report period on October 11, 198 , - _ _ _ - - _ - - - - - _ _ - - - - - _ _ _ _ _ - - - - - - - _ - - - _ - - . _ - - - - - - - - . - - - - - - - - - - - - - - _ _ _ - - - - - - - - - - - - _ - - - ----_-- -_____
- *
ENCLOSURE 2 e
l
.
%
ATT,ACHME_NT 2 (RTI 87-07)
STORAGE BATTERY INSPECTION SAMPLE The following identifies the wet cell battery inspection sample. It may be provided to the licensee for more efficient identification of data relevant to assessing compliance with the current licensing basi . C_eneral Battery Information Document the below information for batteries which carry vital load (1) Qualified, or design, seismic lif (2) Qualified, or design, electrical lif (3) Ag (4) Time in servic (5) Plans for replacement, Previous Licensee Actions Identify actions taken on the following IE Information Notices: 63-11, Possible seismic Vulnerability of Old Lead Storage Batteries. :%-83 Various Battery Preblems: 85-74, Station Battery Problems; an., e6-37, Degradation of Station Batterie . Seismic L_itet_ine and Ovali,f,i:ation For batteries su;)plytrg vital loads, identify the follcwing informatio (1) Licensee and/or manufacturer's establishment of seismic lifettn This traybe through cocumentation allowing verification by competent personnel other than the qualifiers and containing design specifica-tiens, the cualtfication method, results, and justifications (ref:
IEEE 535-19e6).
(2) Seismic Qualification maintenance. Identify how the criteria for assuring that the battery and rack will maintain saismic qualifica-tion are defined, available, and used for periodic inspections ard cell replacements. Identify the criteria for determination of seismic end of life based upon the in-service condition of t.he batter A2-1
- ------._____ __ ___ _ _
. -
l'
s
.
.
4. Efectrical String and Qualification For batteries supplying vital loads, identify the following informatio (1) Confirmation that the battery size is sufficient to handle the load profile with a suitable margi (2) The means of tracking and control of battery loads such that the-batteries and their replacements will have sufficient capacity throughout design life, if worst case electrolyte temperature and other worst case conditions exist when the battery is called upon to perform its design functio (3) The provisions for consideration of the ef fect of jumpered out cells upon the ability of a battery to perform under worst case condition . Battery Ventilation and Protection From lenition Hazards For batteries carrying vital loads, identify the followin (1) The provisiens for assueing adequacy battery ventilation during noreal operation, outages, charging, and discharg (2) Adequacy of checks of battery ventilation fle (3) Acequacy of controls over battery ventilation impediments such as enclosing the battery space or its ventilation with plastic sheetirg, or any other ventilation obstructions, during cutages and other period (a) Adequacy of hydrogen cetection equipment and its calibration ar$ use, or of the technical justification f er not using such equipmen (5) knowledge of the hydrogen hazard on the part of plant management, operating snift management, and perscnnel who access the battery space (6) Prohibition of het work and smoking in battery spaces. including checking the spaces for the residue of such activit (7) Assurance that battery cells are secured, with post-to-case anc tep-to jar seals tight. Thermometers should not be left in cells after temperatures are measured. Caps on the filler openings should be properly secured when not required to be off. (Cells should be vented only through the flash arrestors.)
(8) The means of assuring proper elimination of water-carrying pipes (e.g.,
HVAC lires) f rom battery spaces, especially those which may carry salt wate A2-2
__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l
. .
f :
W
'. t
(9) The means of positive control over the quality of water added to the batteries to assure that the manuf acturer's recomendations or an appropriate licensee standard are net or exceede i (10) The assurance of elimination of combustibles, and loose equipment and conductors, from battery space . Electrolyte Tempera _ture Control
,
For batteries supplying vital loads, identify the adequacy of the followin (1) Avoidance of localized heat sources such as direct sunlight, radiators, steam pipes, and space heater (2) That the location / arrangement provides for no more than a SF difference in cell temperature, as confirmed by measure. rents representative of operating condition If this is not the case, then the licensee and manufacturer should have identified the i consequer.t impact on expected battery and individuel cell capacity and life, ard surveillance procedures should reflect the additicnal allowable temperature variatio I 7. Charging ;
For batteries carrying vital loads, identify the adequacy of the !
followin ;
(1) Provision for a f reshening charge af ter more than 3 months of being l o' crer circuit, unless determined by the manuf acturer to be l unnecessary to assure rated capacity throughout lif [
(2) AccompitsMent of equalizing charges at 18 month intervals, and wher [
the corrected spect fled gravity (W) of an individual cell is more j than 10 potrt (0.010) below the average of all the cells, and when '
the aserage corrected SG of all cells drop more than 10 points below the average installation value, and if any cell voltage is below 2.13V (Specific manuf acturer's provisions and assessment may allow the ron- t performance of some of these recomrnended charges, or may provice dif ferent criteria.)
(3) Control over Dati,ery water quality such that specified purify is confirmed before addition, that water added just prior to charging ,
is added only t,o bring the electrolyte up to the prescribed minimum (to prevent overflow during charging), and that water added after and !
between charges not bring the level above the prescribed maximum (unless manuf acturer's instructions provide for other water addition '
measures). l I
A2-3 ;
_ . _
_ . ._ - - . .
._ .
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.. .
-
.
Ihat routine float and final end of charge SGs not be taken before 72
'
(4)
hours of float operation af ter completion of the charge and the last water addition, unless the manufacturer's instructions provided otherwise. (The need is for measurement of representative cell levels and average them.)
(5) Establishment and maintenance of float voltage on accordance with the manufacturer's instruction ,
(6) Assursace that single-cell charger use does not violate Class 1E independence from non-class lE equipmen . Performance _ Tests and Replaceme Q Criteria For batteries carrying vital loads, identify the followin (1) Initial acceptance testing which demonstrates the ability of the batteries to meet the manufacturer's ratin l
,
(2) Service testing which cemonstrates the ability to carry the load profile with an appropriate margin for worst case conditions, including end of life loss of capacity under the worst case '
.
electrolyte temperatur i (3) Accomplishment of a performanc9 test (capacity test discharge) -
within the first twc years of service and at 5 year intervals until signs of degradaticn are evident or 85% of the qualified service life is reache (4) Annual performance testing of batteries which show signs of degradation or which have reached 85'. of the qualified service life is reache (5) Enc of electrical life criteria which consider the rapid end of life drop-of f in capacity, worst case state of charge during float service, worst case electrolyte temperature, current 0; loads, and the time needed to replace the battery while it can s*ill handle worst case condition . Other Safety-Significant Wet Cell Batteries For safety-significant wet cell batteries not used for vital loads, show how the raintenance program periodically i ermines the ability to perform the cesign function and provides 4 timely replacement of batteries and for maintaining associated equipment (e.g., chargers).
i A2-4 i_