ML18067A565

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LER 96-013-01:on 961115,DC Breaker Failed During Testing for as-found Trip Setting.Failure Caused by Oversight within Preventive Maint Program.Breaker Was Replaced & Tested
ML18067A565
Person / Time
Site: Palisades Entergy icon.png
Issue date: 06/03/1997
From: Toner K
CMS ENERGY CORP.
To:
Shared Package
ML18067A564 List:
References
LER-96-013, LER-96-13, NUDOCS 9706100155
Download: ML18067A565 (21)
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NRCFORM366 U.S. LEAR REGULATORY COMMISSION PROVED BY OMB NO. 3160..0104 (4195) EXPIRES 4130/98 ESll*TED BURDEN PER RESPONSE TO COMPLY W!1M THIS MANDATORY INFORMATION COUECTlON REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED ARE INCORPORATED

  • LICENSEE EVENT REPORT (LER) . INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-8 F33), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 205~

0001, AND TO ll!E PAPERWORK REDUCTlON PROJECT (315().011W, OFFICE OF (See reverse for required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503 FACILITYNAME(1) CONSUMERS ENERGY COMPANY DOCKET NUMBER (2) Page (3) 05000255 1of8 PALISADES NUCLEAR PLANT TITLE (4) LICENSEE EVENT REPORT 96-013, SUPPLEMENT 01 - DC BREAKER FAILURE DURING TESTING FOR AS-FOUND TRIP SETIING EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR I SEQUENTIAL NUMBER REVISION NUMBER MONTH .DAY YEAR FACIL[TY NAME

. DOCKET NUMBER 05000 FACILITY NAME DOCKET NUMBER 11 15 96 96 - 13 - 01 06 03 97 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF*10 CFR§: (Check one or more) (11)

MODE (9) N 20.2201(b) 20.2203(a)(2)(v) 50. 73(a)(2)(i) 50. 73(a)(2)(iii)

I I 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

I POWER.

LEVEL (10) 0 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71

  • 1~1~1111111 20.2203(a)(2)(ii)
  • 20.2203(a)(4) 50.73(a)(2)(iv) OTHER I

20.2203(a)(2)(iil) 50.36(c)(1) x 50.73(a)(2)(v) Specify In Abstract below or 20.2203(a)(2)CM 50.36(c)(2) *so.73(a)(2)(vii) in NRC Form 366A LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (Include Area Co~e)

KERRY A TONER, REGULATORY* AFFAIRS SUPERVISOR (616) 764-2000 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13}

CAUSE SYSTEM. COMPONENT* MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE

.. 'TONPRDS TONPRDS EC BKR W121 y SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR I YES If yes COMPLETE EXPECTED COMPLETION DATE x I NO EXPECTED SUBMISSION DATE (15)

ABSTRACT (Limit to 1400 spaces,' i.e., approxlmately. 15 single-spaced typewritten lines) (16)

On November 15, 1996, at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />, with the pi ant shut down for refueling, testing of a DC I molded-case circuit breaker (72-207) revealed that the breaker Y/Ould not trip on overcurrent. I Subsequently, another breaker (72-228) was removed from the same distribution panel for testing 1*

and failed to trip. Two additional hreakers (72-213 and 72-230), installed as spares in the same I panel, were then removed for testing and also failed to trip. These failures resulted in concerns I that all 72 DC molded-case circuit breakers installed in certain DC distribution panels might fail to I trip when subjected to a short circuit. I I

A number of these breakers act as isolation devices between 1E and non-1 E circuits, and provide I circuit protection coordinated with upstream protective devices. A fault in a non-1 E circuit could I potentially have caused the distribution panel supply fuse to blow, removing power from the panel I serving class 1E loads. As many of the non-1 E circuits powered from redundant panels run in I common cable trays, there was also a potential that a common mode failure could affect redundant I trains of 1E equ~pment. All 72 breakers in the affected DC distribution panels were replaced during 1.

the recent refueling outage. I 9706100155 970603 PDR ADOCK 05000255 S PDR

,., NRC. FORM 366a ** NUCLEAR REGULATORY COMMISSION 4195 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAM~ f1 \ PAGE f3\

I DOCKETf2\ L ER NUMBER 16\

YEAR . SEQUENTIAL REVISION NUMBER NUMBER 05000255 2 of 8 PALISADES NUCLEAR PLANT .

96 - 013 - 01 TEXT (If more space is required, use additional copie5 of NRC Form 366A) (17)

EVENT PISCUSSION Event Description LER 96-005 reported that breakers in two DC distribution panels were not coordinated with upstream panel fuses. During the recent refueling outage, new fuses were being provided in the circuits feeding the panels and the trip settings of three branch circuit breakers were being reset.

The first branch breaker to ha.Ve its setpoint reset was removed from service and turned over' for testing on November 15, 1996. At 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on November 15, with the plant shut down for refueling, testing of DC circuit breaker (72-207) revealed that the breaker would not trip on overcurrent. Subsequently, another breaker (72-228) from the same distribution panel was removed from service for testing and failed to trip. Two additional breakers (72-213, 72-230),

installed as spares in the same panel, were removed for testing and also failed.to trip. Further testing was stopped due to concerns with personnel safety, and the desire not to disrupt loads serving safe shutdown functions ..

These failures resulted in concern~ that all 72 DC molded-case circuit breakers (MCCBs) installed in DC distribution panels D11-1, D11-2, D21-1 and D21-2 (Enclosure 1) wou.ld not trip when subjected to a short circuit. A number of these breakers act as isolation devices between 1E and non-1 E circuits, and provide circuit protection coordinated with upstream protective devices to provide overall panel protection and selective tripping. A fault in a non-1 E circuit could potentially have caused the distribution panel supply fuse to blow, removing power from the panel serving class 1E loads. As many of the non-1.E circuits powered from redundant panels run in common* -

cable trays, there was also a potential that a common mode failure could affect redundant trains of 1E equipment.

Prior to restart from the recent refueling outage, all 72 MCCBs located in the DC panels were replaced. Subsequent to the outage, testing was completed for all 72 MCCBs on January 23, 1997. All magnetic-only MCCBs (44 total) failed to trip under any fault current. All thermal-magnetic MCCBs passed the as-found testing within setting sheet specifications. I I

Inspection of the operating and trip mechanisms revealed differences in the design of the two I breaker types. The common trip arm on the older magnetic-only breakers is made of a compressed I board type of material, while the thermal-magnetic breaker trip arm is made of hard plastic. The I magnetic-only breakers use one solenoid per phase under the tripping mechanism to carry the . I current; which in turn pulls in the common trip arm causing the breaker to trip. The thermal- I magnetic breakers use a copper stranded wire to carry the current. Current heats the wire causing I it to expand and push on a plate which actuates the common trip arm to trip the breaker. I Inspection of the new replacement breakers currently in service revealed the same design with the I common trip arm being made of hard plastic. I

NRC FORM 366a .S. NUCLEAR REGULATORY COMMISSION 4195 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME l1 l . DOCKETl2l LER NUMBER 16l PAGE l3l PALISADES NUCLEAR PLANT 05000255 YEAR I SEQUENTIAL NUMBER REVISION NUMBER 30F8 96 - 013 - 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

Event Extent 1*

I Thi~ failure-to;.trip condition is limited to magnetic-only breakers installed in DC panels D11-1, I D11-2, D21-1 and D_21-2. DC and AC systems were reviewed to determine proolem extent. The I major safety concerl') associated with these particular breakers is their 1E to n.on-1 E electrical I circuit isolation function, and the associated common mode failure for bo.th trains of DC power: I There are no other MCCBs in the plant other than those fr1 these DC distribution panels which I perform this function. * *

  • I r

Palisades DC System:

MCCBs within the DC syst~m. with the exception of those in the DC distribution panels, are include_d in the Periodfc and Pre.:.Planned Activity Control System which tests ttie trip setting of the breaker.: Th~re are two DC distribution panels besides. the four panels which are the subject of this LER. These panels, D11-A and D21--A, were installed in 1980 as part of Appendix R fire protection modifications .. There are n'o breakers used for electrical circuit isolation in either of these panels.

The breakers-in these panels are from a different man_ufacturer than those in the subject panels. :A review of industry expe_rience shows no similar* failures for the models installed in D11-A and _

D21-A.

During tt)e recent refueling outage, three breakers from these two panels were trip tested without signs of ttiis generic problem. Additionally, approximately 50% of the breakers in these two panels have been trip tested or tripped in service since 1989. No generic o_r abnormal trends have been detected during testing of these breakers. Additionally, one breaker from panel D21-Awas,taken apart for evaruation: The* internal trip mechanism for these breakers is a much sturdier design. The internal lubrication of these breakers is minimal and appears to be based on a suspended graphite 1*

in isopro.panol *solution. Based on this, it is concluded that the breaker deficiency does not extend I to these panels. * *

  • I Palisades AC System:

MCCBs in 120 VAC and 480 VAC systems were also reviewed. In the 120 VAC system, MCCBs are installed in* the output of the preferred AC inverters. These breakers are not used as 1E to non-1 E electrical isolation *devices in this application. The breakers function as thermal-only due to the current limiting design of the inverters. These MCCBs would be categorized as moderate risk and trip testing would not be recommended in accordance with the BWR_owners group recommendations (see "Industry Experience"). Further, these breakers are cycled on a minimum six year frequency when the inverters are removed from service to perform preventive maintenance. Current plans call for replacing these breakers during the next refueling outage when the inverters are replaced.

I.!=======================================================================~

.I

NRC FORM 366a .S. NUCLEAR REGULATORY COMMISSION 4195 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION

. DOCKET(2)

I FACILITY NAME l1 \ LER NUMBER 16) PAGE (3)

YEAR SEQUENTIAL REVISION NUMBER NUMBER 05000255 4 OF8 PALISADES NUCLEAR PLANT 96 - 013 - 01 TEXT (If more space Is required, use additional copies of NRC Form 366A) (17)

The 277/120 VAC breakers in lighting panels were also reviewed. Two panels were found to have Class 1E safety related functions. They include Containment Hydrogen Monitoring and Anticipated Transient Without Scram. Neither of these functions are required to achieve and maintain hot shutdown conditions. The feeder breakers to these panels were also reviewed and found to be dedicated to these panels.

The 480 VAC MCCBs are checked iri conjunction with the associated break~r preventive mainten*ance performed per maintenance procedure SPS-E-11. At a minimum, the MCCBs are cycled five times to operate the mechanical linkages and exercise the contacts. A records review showed that 84% of all 480 VAC breakers have been current-injection tested since 1986: No generic or abnormal trends have been detected during testing of these breakers.

  • Industry. Experience I I

An industry.review was conducted to identify recommended breaker preventive maintenance. A 1*

recent paper p*ublished by the* BWR owners group provides recommended maintenance for . I MC'CBs based on the risks associated with their faHure. A cbpy of their recommended test matrix is.

provided as Enclosure ~. (Enclosure 3 references the BWRpaper.) It is considered that MCCBs used as isolation devices would be in. the high risk ~tegory '.

  • Industry experience, as discussed both in the BWR owners group paper and in EPRI publications, suggests that periodic cycling is sufficient to maintain breaker function. Our own evaluation of the*

breaker models in the affected distribution panels concluded that simply actuating the breakers' on/off switches is not sufficient preventive maintenance, as it does not exercise the breaker trip mechanism. Newer breakers have "push-to-trip" buttons that sufficiently cycle the breaker's trip mechanism; h9Wever, these older breakers did not have this function. The only maintenance that could have identified this generic failure earlier would be over-current testing on a periodic basis.

CAUSE OE THE EVENT This event represents an oversight within the plant's preventive maintenance program to maintain avaHable and reliable, critical operating characteristics of a class of DC circuit breakers important I*

to safety. In the past, the need for maintenance on these breakers was discussed by engineering, I operating and maintenance staff, but omitted due to risk and potential impact on in-service loads. 1*

Notable breaker testing has been completed on other plant systems that have lower safety I significance associated with their removal from service. I I

I I

NRC FORM 366a .S. NUCLEAR REGULATORY COMMISSION 4195 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION I

FACILITY NAME C1) . DOCKETC2\

  • LER NUMBER 6\ PAGE C3\

YEAR . SEQUENTIAL REVISION NUMBER NUMBER 05000255 50F8 PALISADES NUCLEAR PLANT 96 - 013 - 01 TEXT (It more space is required, use additional copies of NRC Form 366A) (17)

In the p~st, it had beeri concluded that any need to perform preventive maintenance resutting in cycling the breakers open and closed was overcome by a greater need to maintain connected loads in continuous service due to the nature ofthe functions provided .. Loads supplied from these panels serve to maintain safe power operation or shutdown conditions. In addifion, the panel*

design makes it difficult to remove the breakers with the p~nel energized and there have been concerns for personnel safety and electrical equipment protection. It was consiqered that due to the simple design of these devices and the fact that there was no vendor recommended maintenance, there was a low probability of failure.

The technical root cause for this event was hardening of breaker internal lubrication which restricted required movement of breaker intern~I components to effect .breaker trip.

Numerous magnetic-only breakers wer~ disassembled following testing. Their solenoid coils had been destroyed due to their inability to isolate fault current The breakers' internal$ were sticky due to solenoid coil insulation breakdown. The internal lubrication was not excessive co*mpared to* the lubrication in the thermal~magnetic breakers. However, the spring-actuated trip latch on the magnetic-only breakers was stiff and did not move freely, unlike the latch on the thermal-magnetic breakers. After cleaning the joints with authorized cleaning lubricant and manually exercising the latch, the magnetic-only bre~kers would trip in a normal manner. The apparent cause of the magne.tic~only breakers' *failure is hardening of the lubricant on the pivot points. Lubricant hardening is likely the r0$ult of resin leaching from the. solenoid coils.

  • SAFETY SIGNIFICANCE The* safety significance of DC MCCBs failing to trip with potential loss of redundant 125 VDC distribution panels (011~1. 011-2, 021-1 and 021-2), requires *assessment of the capabifity to achieve and maintain safe reactor shutdown. A common mode failure which challenges distribution panel branch MCCBs would be the result of external condi!ions such as fire, pipe break or missile_s in areas of the plant where cables fro.m redundant panels are routed together in close proximity to one another. These cables could be shorted resulting in automatic isolation of redundant DC panels due to protective action of upstream panel fuses should the branch MCCBs fail to trip.

Loss of the DC distribution panels could not have resulted as a consequence of a design basis accident (OBA); either a loss of coolant accident or a main steam line break. There are no I unprotected (unfused) cables associated with the faulty MCCBs routed in areas of the plant I subject to a harsh environment. Thus, the panels would not have been affected as a direct result I of a OBA. Further, given the locations of unprotected (unfused) cables in relation to high energy .I piping or rotating equipment, it is not conceivable that any initiating event that would have resulted I in a loss of the redundant DC panels would also have resulted in initiating a OBA.

I

NRC FORM 366a .S. NUCLEAR REGULATORY COMMISSION 4195 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME l1 l DOCKETl2\

05000255 I

YEAR ILER NUMBER 16\

SEQUENTIAL .

NUMBER REVISION NUMBER PAGE l3\

60F8 PALISADES NUCLEAR PLANT 96 - 013 - 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

Regarding the effects of fire, our design basi~ precludes consideration of a OBA concurrent with I an area fire. The Post Fire Safe Shutdown Analysis ahd associated operating procedures address I postulated area fires; assuring .that equipment needed to achieve safe reactor shutdown is I preserved and utilized. Assessment of the ability to achieve and maintain safe'reactor shutdown I in the event of loss of one or both DC power trains follows. (Refer to Enclosure 1 for DC system .I configuration.) I I

Failure of a 125 VDC breaker to trip when subjected to a short circuit can lead to loss of a DC I distribution panel. The panel would be lost when the fuse blows upstream of the failed breaker. I Additionally, if the cause of the event occurs in an area common to circuits of all four of the I distribution panels, it is possible to have multiple short circuits. Due to the assumed proximity of I the cables, the event could cause the upstream fuses in each of the four distribution panels (011- I 1, 011-2, 021-1, 021-2) to blow. *

  • I I

The loss of a single 125 VDC distribution panel, or the loss of two panels of the same train, is .I accounted for by the design configuration of the 125 VDC system. The system contains sufficient I redundancy so the *loss of one complete train of 125 VDC power does not impact the ability of the I operators to maintain the plant in a safe condition. The plant Off Normal Procedure, "ONP-2.3 I Loss of DC Power, is written with the assumption that only one 125 VOC distribution train .is available. The ONP, therefore, has guidance that bounds the loss of one or two 125 VDC distribution panels of the same train.

  • Off Normal Procedure ONP-2.3 is not designed to be used when the loss of DC power extends into both 125 VDC trains. A postulated fire in an area such as the Turbine Building, could result in multiple blown fuses in redundant 125 VDC distribution panels since the area contains circuits
  • from each of the panels. The postulated situation is similar to the result of a postulated fire in the Cable Spreading Room, where the following is lost: *
  • Both* 125 VDC Control Centers (020 (right and left) and 010 (right and left))

All four Battery Chargers (#1,#2,#3,#4)

All four Inverters (#1,#2,#3,#4) and Preferred AC panels (EY-10,EY~20,EY-30,EY-40) I All four 125 voe distribution panels (011-1, 011-2, 021-1, 021-2) I I

Since DC power is required for safe shutdown, the plant design configuration includes two I additional 125 VDC distribution panels (011-A and 021-A) that are not located in the Cable I Spreading Room nor have circuits running through the Turbine Building. This design ensures I sustained 125 voe power is available for vital equipment following a design basis fire. The 125 I VDC panels 011-A and 021-A do not contain breakers of the same type as those connected to the *1 four 125VDC distribution panels 011-1, 011-2, 021-1 021-2. Therefore, the 011-A and 021-A I

NRC FORM 366a U.S. NUCLEAR REGULATORY COMMISSION

..4195.

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION I

FACILITY NAMF f1 \ DOCKET<2l LER NUMBER 16\ PAGE f3\

YEAR SEQUENTIAL REVISION NUMBER NUMBER 05000255 70F8 PALISADES NUCLEAR PLANT 96 - 013 - 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) breakers are not subject to the breaker failure mode as described in this LER for the breakers I connected to the four distribution panels. 011-A and D21-A breakers will remain energized during I the event which may cause isolation of the four distribution panels. * * . .I The Post Fire Safe Shutdown Analysis (PFSSA) is a bounding analysis for loss of multiple 125 I VDC distribution panels. The PFSSA is a conservative analysis for use in this event assessment in that the analysis assumes .a loss ofoffsite power occurs coincident with loss of.the DC.

distribution panels 011'-1, 011-2, ,P21-1 and 021.:2. In the event considered here,

  • where a cemmon mode failure results in isolation and loss of four 125 voe distribution panels, the 125 .

VOC Control Centers, Chargers, Inverters,. and Preferred AC buses remain energized; *loss of offsite power is not a direct result of this event. Evaluation of the PFSSA has shown that for this, and more significant events, sufficient equipment is available to safely shutdown the plant and maintain the plant in Hot Shut~own for an extended per,iod..

  • Even with the loss of all fdur.125 VDC di~trib1:.1tion p~ne.ls (011-1,:011-2, 021-1 and 021-2),

sufficient equipment remains available to mai!ltain the plant'in a safe condition. Off Normal Procedure ONP-25.2, "Alternate Safe Shutdown* Procedure", writte*n for an inability to maintain control of the plantfrorri the Contr~I Room, provides guidance necessary for operatqrs to respond to an* event such .as that. de.scribed .in this LER.. Enclosure .4 provides the evaluation ofthe PF SSA and describes how each critical safety function is affected by the loss of 125 VDC. power. The . I*

enclosure also explains why each safety'function i~ not adversely affected.

  • I I

CORRECTIVE*ACTION ** I I

Correctiv~ aCtions taken include.breaker replacement and testing, determining the extent of the t.

deficiency throughout the plant's. electrical system, ~nd. identification of vendor recommendation~ I" for periodic maintenance. Actions to be taken include the development of effective periodic I breaker .

maintenance

. and rev.iew. ~

for other eql,iipment .potentially not maintained as necessary-.

.I I

Actions Completed J I

1: All 72 MCCBs in. DC distribution panels 011-1,* 011-2,.. 021-1 and 021-2 were replaced and .I

.. successfuliy trip tested prior to' startup from the recent refueling outage. I** .

I'

2. A review of all other plant installed MCCBs was completed prior to startup from the recent
  • l
  • refueling outage to determine the extent of the problem. The review revealed the problem I
was limiteq to DC panels 011-1, 011-2, 021-1 and 021-2. I I

NRC FORM 366a .S. NUCLEAR REGULATORY COMMISSION 4195 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION I

S:ACILITY NAME f1 \ DOCKETf2) LER NUMBER 6) PAGE f3\

YEAR SEQUENTIAL REVISION NUMBER NUMBER 05000255 80F8 PALISADES NUCLEAR PLANT 96 - 013 - 01 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

3. Subsequent to the recent refueling outage, all 72 MCCBs removed from DC panels D11-1, I 011-2, D21-1 and D21-2 have been as-found trip tested. The testing revealed that the failure I to trip was limited to the magnetic-only breakers in these DC panels. I I
4. A review of industry experience regarding MCCB maintenance was conducted to obtain 1.

information with which to develop the testi.ng program for MCCBs ~t Palisades. *1 I

Actions To Be Completed I I

1. For high safety significant systems (defined by Palisades Procedure EM-25 "Maintenance I Rule"), system engineers will perform reviews to determine if preventive maintenance (PM) is 1-required but not being performed. The reviews will consider sources such as plant. 1 Preventive Maintenance Optimization, Deferred Maintenance, and Industry Experience as l applicable.* Objectives are to determine: 1) if justification is adequate for not having I implemented pre-identified plant PM, 2) if deferred plant PM should be reactivated to I mairitain system integrity, and 3) if any additional PM should be performed as indicated by I conditions or events throughout the industry. I

. Following completion of these reviews, the Maintenance Rule Expert Panel will review the I

  • input from each system engineer to validate the findings. The system engineer and expert I panel reviews will be completed by* July 1,* 1997. I J
2. By July 1, 1997, a review of open industry experience documents greater than one year old I will be completed to determine if there are any significant items similar to the MCCB issue I which have not been resolved in a timely manner. - I 1*
3. A program will be developed by December 1, 1997 for periodic testing of MCCBs. This .I schedule assures that results of equipment maintenance reviews identified above will be I, _

available and factored into testing program development as necessary. I I

PREVIOUS ~IMILAR EVENTS _

Licensee Event Report 96-005, "DC Panels ED-11-1 and ED-21-1 Breaker/Fuse Coordination - I.

Appendix R Enhancement Analysis" I

ENCLOSURE1 CONSUMERS ENERGY COMPANY PALISADES PLANT DOCKET 50-255 LICENSEE EVENT REPORT 96-013 SUPPLEMENT 1 PALISADES DC SYSTEM CONNECTION DIAGRAM 1 page

. I PALISADES DC SYSTEM MCC1 MCC2 MCC2* MCC1

.i CHGR1 CHGR3 CHGR2 CHGR4 72-18 72-15 ,!

  • 72-12* 72-28 72-25 72-22 DC BUS D-10 R 72-10 D-10 L . DC BUS D-20 R 72-20 D-20 L 72-16 72-11 72-26 . 72-21

~*

D11-1 D11-2 D21-1 D21-2 INVERTER 1 DIST PNL DIST PNL I INVERTER3 INVERTER 2 DIST PNL DIST PNL INVERTER4 Y-10 Y-30 Y-20 Y-40 TRB BLDG

._____________________________________________________ &OTHER

,___________________________________________________________ ROUTES

)

I

ENCLOSURE 2 CONSUMERSENERGYCOMPANY PALISADES PLANT DOCKET 50-255 LICENSEE EVENT REPORT 96-013 SUPPLEMENT 1

  • ewR OWNER'S GROUP

. RECOMMENDED T,EST MATRIX

} .

1 page

BWR OWNER'S GROUP MOLDED CASE CIRCUIT BREAKER TEST MATRIX RISK TEST SCOPE TEST SIGNIFICANCE FREQUENCY High

  • Overheating inspection 6 -*10 years
  • Mechanical operation
  • Enclosure inspection

.

  • Overload trip test
  • Instantaneous over-current trip test
  • Shunt trip or Undetvoltage test (where applicable)

Medium .* Overheating inspection 9-20 years

  • Mechanical operation
  • Enclosure inspection
  • . Low None required beyond pre-installation No periodic testing testing required

-, , .. -~ ,.* : . .

ENCLOSURE 3 CONSUMERS ENERGY COMPANY PALISADES PLANT

  • DOCKET 50-255 LICENSEE EVENT REPORT ss*-013SUPPLEMENT1 PLANT AND INDUSTRY DOCUMENTS REVIEWED 1 page

LICENSEE EVENT REPORT 96-013 SUPPLEMENT 1 PLANT AND INDUSTRY DOCUMENTS REVIEWED 1

  • LER 96-005 "DC Panels Breaker/Fuse Coordination"
2. Maintenance Order Work History for DC Breakers
3. NEMA AB-4-1991 "Gui.delines For Inspection and Preventive Maintenance of Molded Case Circuit Breakers Used In Commercial and Industrial Applications"
4. EPRI NMAC NP-7410-V3 "Molded Case Circuit.Breaker Application:and Maintenance
  • . -Guide"
5. BWR Owner's Group OG96-732-185 "Final Report on Molded Case Circuit Breaker Maintenance Program" *
6.
  • ENCLOSURE4
  • CONSUMERSENERGYCOMPANY PALISADES PLANT
  • . DOCKET 50-255 LICENSEE EVENT REPORT 96-013 SUPPLEMENT 1 .

. CRITICAL SAFETY FUNCTIONS 6 pages

  • . CRITICAL SAFETY FUNCTIONS REACTIVITY CONTROL The capability to manually trip, and the availability of the automatic trip functions, are always maintained regardless of 125 VDC status. Reactivity control is not initially affected since, if the reactor is reset, it will automatically trip and all full length control rods-will be inserted into the core. Once vital AC power is restored, additional shutdown margin can be achieved using the guidance of ONP-25.2.
  • PRIMARY COOLANT SVSTEM (PCS) INVENTORY CONTROL PCS inventory control*is not initially affected, since charging and letdown are simultaneously stopped. Makeup to the PCS is dependant on restoration of AC power. Once AC power is restored, procedural guidance is provided for reestablishing the charging system. Instructions are also provided to ensure sufficient concentrated boric acid is added to the PCS for shutdown margin considerations. Even without any makeup to the PCS, subcooling is maintained until sucf1 time that either the charging system is restored or the High Pressure Safety Injection system is able to provide makeup. *

.COMPONENT AFFECTED EFFECTS FROM LOSS OF DC ALTERNATIVES POWER Charging System Charging Pump breakers fail as- Can manually operate breakers is; P-55A, P-55B, P-55C at the 480 VAC load center (when AC power is available) .

Charging System Stop valves fail No response is required. Open

' Open; CV-2111, -2113, -2115 is the desired position for establishing charging t~ the PCS Letdown Systeni .. Letdown orifice valves.Close .. Letdown is not required since charging can be controlled manually.

Primary' Coolant Pump Normal controlled Bleed off PCP controlled Bleed off will be Controlled Bleed off valves to the Volume Control established through the alternate Tank fail Closed; CV-2083, pathway.

-2099. Alternate controlled .

Bleed off valve fails Open; CV-2191

  • 2 PRIMARY COOLANT SYSTEM PRESSURE CONTROL The loss of AC power combined with the loss of all four 125 VDC distribution panels (011-1, 011-2, 021-1 and 021-2) will limit the ability to control PCS pressure. The goal for pressure control without the use of heaters or sprays, is to control PCS temperature as stable as
  • pos~ible. Maintaining a stable PCS temperature (using the Steam Generators) will minimize the Pressurizer in-surges and out-surges, which complicate PCS pressure control.

Once AC power is restored, Pressurizer heaters can be reestablished to aid in maintaining PCS subcooling. Without the capability to go onto Shutdown Cooling (due-to the lack of AC

  • power) the procedures will direct the operators to remain in a* hot condition until AC power is restored.

Comoonents affected Effects from loss of DC oower Alternatives Pressurizer Auxiliary Spray Auxiliary Spray valve fails Main Pressurizer Spray, if 4160.

  • Closed; CV-2117
  • VAC power is available. If 4160 VAC power is not available, then the ability to lower PCS Pressure will be* limited. -*

Pressurizer Heaters Lose control power to . . . . .. Can .manually trip breakers at * '

pressurizer heater breaker. the switchgear if 2400 VAC 152-305 on 2400 V bus 1E. power is available to bus 1E.

Breaker 152-305 will not trip

  • automatically on Low Pressurizer Level.

Power Operated Relief Valves PORVs fail Closed; PCS code safety valves are (PORVs). PRV-10428,-10438. available for pressure control.

Head vent valves fail Closed; PCS cod~ safety valves are Reactor}iead Vent System .. -*- PRV-1067,-:1068,-1069,-1070, . available for pressure control.

.. . -1071,-1072.

Control of PCS temperature may be enhanced. using the Hogging Air Ejector as a steaming path. The plant is maintained in Hot Shutdown until such time as sufficient equipme*nt is available to support a cooldown.

Component Affected Effects from loss of DC Power Alternatives Main Feedwater Pumps Lose the ability to trip both main The operators can trip the main feedwater pumps from the feedwater pumps locally.

Control Room. Continued feedwater pump operation may cause excessive cooling of the PCS.

Auxiliary Feedwater Auxiliary Feedwater pump P-88 None required, P-88 operation is steam supply valve fails Open desired to maintain steam and P-88 starts. CV-05228 generator levels.

Auxiliary Feedwater Automatic Low level initiation is not required initiation on low steam generator since P-88. starts upon loss of level is lost. DC power. Additionally, AFW pumps P-8A and p.:8c can be controlled manually.

Atmospheric Steam Dump Atmosphe~c Du_mp Valves fail The Steam Generator code .

Valves (ADVs) Closed, safety valves will function to limit CV-0779,-0780,-0781,-0782. Steam Generator pressure and Primary Coolant System temperature. Procedures will direct operators to use hogging

-- .- .. air ejector steam supply for heat .

  • .. . removal when cooling down .

Turbine Bypass Valve {TBV) . TBV fails Closed None are required. Since it is assumed a loss of condenser vacuum accompanies the event, TBV use is not *desirable.

Main Steam Isolation Valves MSIVs cannot be closed Can close MSIVs manually at (MS IVs) electrically, CV-0501,-0510 the solenoid valve cabinets in the Turbine building or 1 D Switchgear room.

Main Turbine Protection Lose ability to trip the Turbine Can locally trip the Turbine at electrically from the Control the Turbine Pedestal.

Room. If not tripped, eventually the PCS will be overcooled.

  • 4 CONTAINMENT ISOLATION Containment Isolation is not adversely affected by the loss of 125VDC power as all isolation valves are DC powered and are designed to fail in the safety position.

Components Affected Effects from loss of Power Alternatives Containment Isolation Valves The valves fail Closed, None are required since Closed is CV-0737 ,-0738,-0739,-0767, the Safety position.

-0770,-0771,-1001,-1002,-1004,

-1007 ,-1036,-1038,-1044,-1045,

-1064,-1065,-1101,-1102,-1103,

-1104,-1358,-1501,-1502,-1503,

-1910,-1911.

Component Cooling' Water Valves fail Open; None since Open is the Safety Containment Isolation Valves CV-0910,-0911,-0940. position and no accident is assumed.

Service Water' System Valves fail Open; CV-0824,-0847. None since Open is the Safety

  • qoi:itainment lsolationValves , position and no accident is assumed.

Valves fail closed and monitors Containment Hydrog.en Monitor are not available .. Sampling the. Containment' for Hydrogen is not required since no accident is assumed.

  • 5 VITAL AUXILIARIES WATER AND AIR Maintaining the Service Water System (SWS), Component Cooling Water System (CCW),
  • and Instrument Air System is dependant upon restoration of AC Power. Until the AC power is restored, the loss of the DC distribution panels (011-1, 011-2, 021-1 and 021-2) will result in control valves failing to the safe position. Once AC power is restored, procedural guidance will direct the operators to reestablish the vital air and water auxiliary systems.

Component Affected Effects from loss of DC power Alternatives Seryice Water System (SWS) The following valves fail Open; CV-0824,-0847,-0876,-0877, Can manually operate CV-0824.

-0826,-0844,-0857,-1319, None required as these valves CV-1359 fails Closed. fail to the desired positions.

Component Cooling Water The following valves fail Open; None required as these valves System (CCW) CV-0913,-0937 ,-0938,-0945,-09 fail to the desired position.

46,-0947,-0948,.,.0949,-0950.

Service/Instrument Air. System Service Air isolation valve There is no safety .related Closes; CV-1212. function for this valve once it is closed.

Feedwater Purity Air system This is the desired position*on'ce cross-tie valve fails Open; AC power is restored.

CV-1221.

Instrument Air to Containment . None required as this is the fails Open; CV-12~1.

  • desired position for a loss of 125 voe.

Standby air compressor will not When AC power is restored start. these may be manually

- operated.

Air compressors C-2A and C-2C When AC power is restored this will not unload. compressor C-28 can be placed in service.

  • 6 VITAL ELECTRICAL AUXILIARIES It is assumed that allfour 145VDC distribution panels (D11-1, D11.-2, D21-1 and D21-2) are lost, however; the safe .shutdown 125VDC .distribution panels (D11-A and D21-A) are n_ot
  • affected. D11-A and D21-A power vital DC circuits for Class 1E 2400 VAC buses 1c *and 1D, Diesel. Generators 1-1 and 1-2, and other control and instrument circuits for necessary for safe plant shutdown. The goal of each. of the Off-Normal and Emergency procedures is to establish necessary AC power while conserving. the batteries. Battery conservation is
  • accomplished by maintaining discharge amperes below the limits assume'd in the station battery profile.

In th.e event that offsite power were lost, the Diesel Generators would not automatically start and load onto the safety related buses. However, control power required to manually *start .

and load the Diesel.Genemitors would be available.. Off-Normal.Proeedure ONP-20, "Diesel Generator Manual Control", provides di'rection regarding re-energizing the safety related

  • 24.00VAC buses. * *
  • Comoonent Affected Effects from loss of DC power Alternatives Diesel Generator 1-1 * .. Won't start on Bus *1 C Can manually st~rt the DIG .

.. .. * *undervoltage: OutPut breaker Can manually close the output .

.- will not auto close . breaker.

Diesel G.enerator 1.:2

  • Won't start on Bus 1D Can manually start the DIG.

undervoltage. Outp_ut breaker Can. manually close* the output

.. .. will not auto close. breaker.*

Bus 1c *

  • Won't fast fransfer to startup Can manually transfer to startup

.. power. power.

~...; Bus 1D ... - -**** _,, Won't fast transfertO startup Can manually transfer to startup

  • .. . power. . .

.. power .

480 VAC load centers Los.e control power to breakers *Can manually operate breakers at the sWitchgear.

4160 Volt bu~es Lose control power for breakers Can manually operate breakers

' at the*switch!lear.

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