ML19305C735

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LER 80-010/01T-1:on 800226,during Routine Operation,Power Supply Failure Initiated Transient Condition,Resulting in Reactor Trip,Turbine Trip & Engineered Safeguards Actuation. Caused by Failure of Nni Voltage Buffer Card 4-7-10(4)
ML19305C735
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 03/20/1980
From: Lancaster K
FLORIDA POWER CORP.
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML19305C731 List:
References
LER-80-010-01T, LER-80-10-1T, NUDOCS 8003310343
Download: ML19305C735 (20)
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f (3T71 lDuring routine power operation, a nnwer sunniv c,41 rn 4.,4-<,-,,,,

, ,a transient condition'that resulted in a reactor trip, turbine trin, and on I

, engineered safeguards actuation. A review of the transient revealed events rennr d-able under Technical specifications 6 9 1.9.b, 6.9.1.8, and 10 CFR 20.403. The 1 i

I i ,3 , , , j plant was maintained in Mode 3 (hot standhv) on nneurn! c irctil a t ion until forced RCS flow was initiated at 2107. This is the first occ,trrence l

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,,,o,gThe cause of th'is event is attributed to the fnflure of NNI vn1*,en knfco r I

, card 4-7-10(4). The failure resulted in the loss of the 24 VDC non-nucient t

, instrumentation "X" power supplv. The nower was restored to ebo *NT ,r  !

3 1444 on 26 February 1980. The plant was stabflized and n1n-ed in wvin s I t

,,,,,g(cold shutdown) on 29 February 1980.

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$. @ T . N o P " " h NA mac t.;sa cNov ao l 77l '*]' g; General media release 26 Feb. 1980. [ 'lIitIffIII tI i

,,M of preparer: . K. F. Lancaster ,. (9041 795-4486 i (SEE AT"Af"W5'n SUPP7Nm EGOICLCION SM) i l

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SUPPLEMENTARY INFORMATION Rsport No.: 50-302/80-010/01T-1 Facility: Crystal River Unit #3 Rsport Date: 20 March 1980 J

Occurrence Date: 26 February 1980 Conditions Prior to Occurrence:

Flux 98.6 OTSG "A" FRLV 242 inches RC pressure 2157 psig OTSG "B" FRLV 254 inches PZR level 202 inches OTSG "A" pressure 911 psig MU cank level 71 inches OTSG "B" pressure 909 psig Th "A" 599'F Main steam pressure 894 psig j T "B" 600*F Main steam temp. 589*F h

T "A" 557'F Condenser vacuum 1.76 T "B" 556*F Generated FSi 834 Reflow"A"73x106 lbs/hr DFT level 12.7 ft.

RC Flow "B" 73 x 10" lbs/hr Feedflow"A"5x10flbs/hr Letdown flow 48 gpm Feed flow "B" 5 x 10 lbs/hr OTSG "A" level (0P) 67% Feed pressure "A" 970 psig OTSG "B" level (OP) 65% Feed pressure "B" 968 psig i

Description of Occurrence:

At 1423 during routine power operation, the plant suffered a loss of the 24 VDC non-nuclear instrumentation "X" power supply. This resulted in the loss of the "X" power supply instrumentation and initiated a plant transient condition ending in a reactor trip, turbine trip, and high pressure injection. The plant was stabilized and maintained in Mode 3 (hot standby) on natural circulation until forced Reactor Coolant System flow was initiated at 2107. The plant was subsequently taken to Mode 5 (cold shutdown) at 0755 on 29 February 1980.

A review of the plant transient revealed events contrary to the following Technical Specifications.

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IDENTIFICATION / DESCRIPTION OF OCCURRENCE 3.1.2.9 The borated water storage tank shall be operable with a contained borated water volume of between 415,200 and 449,000 gallons.

During the plant transient and as a result of the high pressure injection actuation at 1426 on 26 February 1980, the borated water storage tank volume decreased from approximately 436,914 gallons to approximately 380,000 gallons.

High pressure injection was terminated at 1550 on 26 February 1980.

I Relief for these Technical Specifications was granted as of 1715 on 26 February 1980 until the plant entered into Mode 5 (cold shutdown).

3.4.5 Each steam generator shall be operable with a water level between 18 and 360 inches.

Review of the plant conditions during the transient revealed that a reduction in feedwater caused by the failure of the "X" power supply resulted in steam generator "A" operation with a level less than 18 inches. It was determined that "A" OTSG effectively boiled dry approximately four (4) minutes into the transient at 1427 on 26 February 1980. Level was restored to greater than 18" at approxi-mately 1612 on 26 February 1980 using the auxiliary feedwater header via the bypass valves. -

3.4.6.1 The following Reactor Coolant System leakage detection systems shall be operable:

a) The containment atmosphere iodine radioactivity monitoring system.

b) The containment sump level monitoring system, and, c) the containment atmosphere gaseous radioactivity monitoring system.

3.4.6.2.d Reactor Coolant System leakage shall be limited to 10 gpm identified leakage.

The combined effort of the PORV opening on the NNI(X) 24 VDC power failure and safety valve (RCV-8) relieving to the Reactor Coolant drain tank following high pressure injection. actuation resulted in the rupture disc being blown and leakage exceeding the requirements of TS 3.6.4.2.d. Leakage rates are assumed to be consistent with the estimated HPI flow rates illustrated in Figure III-20 (at-tached). Leakage was maintained within Technical Specification limitations upon termination of HPI at 1550 on 26 February 1980.

The containment sump level monitoring system was rendered inoperable at approxi-mately 1430 on 26 February 1980 when sump level elevated off-scale high follow-ing high pressure injection actuation. The containment atmosphere radiation monitoring system was rendered inoperable upon isolation of the monitor sample line at approximately 1426 on 26 February 1980.

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l Containment atmosphere samples were taken in accordance with post-accident sampling and analysis of Reactor Building Atmosphere Procedure EM-303. The f

containment atmosphere monitoring system was returned to service at 2115 on g 26 February 1980. The containment sump level was returned to normal on 5 March 1980. The total volume accumulation in the sump as a result of the incident was approximately 42,810 gallons.

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Idsntification/ Description 2 J.P. O'Reilly, Director of Occurrence LER 80-10-OlT-1 Relief for surveillance requirements 4.4.6.2.b and 4.4.6.2.d of Techn. cal Specification 3.4.6.2 was granted as of 0915 on 27 February 1980 to continue until the Reactor Building sump was returned to normal.

3.6.1.4 Primary containment internal pressure shall be maintained between 17.7 and 12.7 psia.

The loss of Reactor Coolant via the pressurizer relief system through the Reactor Coolant drain tank resulted in the Reactor Building pressure increasing to approximately 19 psia. RB pressure was reduced to less than the Technical Specification limit of 17.7 psia at 1550 on 26 February 1980 and was stabilized at 15 psia at 1645 on 26 February 1980.

3.6.1.5 Primary containment internal air temperature shall not exceed 130*F.

The average containment temperature exceeded 130*F and reached a maximum average temperature of 142*F during the transient. It was determined that the

{ temperature transient remained above the Technical Specification limit for approximately one (1) hour.

3.4.9.2 The pressurizer temperature shall be limited to a maximum heatup and cooldown of 100*F in any one (1) hour.

Prior to the transient, the pressurizer temperature was 640*F. Approximately three (3) minutes into the transient following high pressure injection

. actuation, the RCS was " solid" and was relieving through the pressurizer code safety (RCV-8). At this time the pressurizer temperature was approximately equal to incore temperature. At 1523 on 26 February 1980 one (1) hour into the transient, incore temperature was 470*F. This temperature differential corresponds to a pressurizer cooldown rate of approximately 170*F in one hour.

3.7.3.2 Two independent decay heat closed cycle cooling water loops shall be operable.

At 1930 on 26 February 1980, decay heat closed cycle cooling pump DCP-1A failed,

, rendering "A" DHCCC loop inoperable. Consequently, the components being supplied i by DCP-1A were also rendered inoperable. These include DHP-1A, DHHE-1A, MUP-1A, BSP-1A, and RWP-3A. These components are identified below in the affected Technical Specifications.

3.5.2.a,b,c. Two independent ECCS subsystems shall be operable with each subsystem comprised on one operable high pressure injection pump (MUP-1A), one operable low pressure injection pump (DHP-1A) and one operable decay heat cooler (DHHE-1A).

3.7.4.2 Two independent decay heat seawater loops shall be operable. (RWP-3A).

3.6.2.1 Two independent containment spray systems shall be operable with each spray system capable of taking suction from the BWST on a containment spray actuation signal and manually transferring suction to the containment sump. (BSP-1A).

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4 Idsntification/ Description 3 J.P.0'Reilly, Director of Occurrence LER 80-10/0lT-1 As a result of the failure of DCP-1A, a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Action Statement was entered.

Subsequent investigation revealed that the pump-motor coupling on DCP-1A had failed.

The coupling was replaced with a spare and DCP-1A was returned to service at 0948 on 28 February 1980 following satisfactory completion af ECCS Pump Operability Surveillance SP-340 Relief for Technical Specifications 3.5.2, 3.7.3.2, and 3.7.4.2 was granted as of 0915 on 27 February 1980 to continue until the plant entered Mode 5 (cold shutdown).

3.3.3.5 The remote shutdown monitoring instrumentation channels shown in Table 3.3-9 of the Technical Specifications shall be operable with readouts displayed external to the Control Room.

A review of the transient revealed that the following remote shutdown instruments were inoperable as a result of the non-nuclear instrumentation "X" power supply failure at 1423.

MU-L4-LI Makeup tank level RC-1-LI-l Pressurizer level RC-4A-TI-2 RCS loop "A" Th RC-4B-TI-2 RCS loop "B" T h SP-1A-LI-3 OTSG "A"SU level SP-6A-PI-l OTSG "A" pressure DH-39-TI Decay heat cooler "A" outlet temp.

Power to these instruments was returned upon restoration of the NNI "X" power supply at 1444, on 26 February 1980.

3.3.3.6 The post-accident monitoring instrumentation channels shown in Table 3.3-10 of the Technical Specifications shall be operable with readouts and recorders in the Control Room.

A review of the transient revealed that the following post-accident monitoring instruments were inoperable as a result of the non-nuclear instrumentation "X" power supply failure at 1423, on 26 February 1980.

Pressuri'.er level LT-1 Pressurizer level LT-2 Pressurizer level LT-3 RC-4A-TI RCS 'aop "A" Th

RC-4B-TI RCS loop "B" T h

RC-13-FR RCS flow SP-6A-PR OTSG "A" outlet pressure SP-7A-FI FW flow "A" l SP-7B-FI FW flow "o" I

- Power to these instruments was returned upon restoration of the NNI "X" power supply at 1444 on 26 February 1980, 3.4.1 Both Reactor Coolant loops and both Reactor Coolant pumps shall be in operation.

Contrary to this specification, operation in Mode 3 (hot standby) proceeded without at least one Reactor Coelant pump or decay heat remotal pump.

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Id:ntification/ Description 4 J.P.0'Reilly of Occurrence LER 80-10-OlT-1 During the transient, following the HPI actuation from low RCS pressure, all Reactor Coolant pumps were tripped in accordance with Loss of Reactor Coolant or Reactor Coolant Pressure, Emergency Procedure, EP-106, and B&W small break guidelines. The plant was maintained on natural circulation until 2107 on 26 February 1980 when forced RCS flow was inititted.

3.4.3 A cinimum of one pressurizer code safety valve shall be operable with a lift setting of 2500 psig 1%.

Contrary to this specification, pressurizer code safety RCV-8 lifted at 2400 psig.

Reactor Coolant System pressure control was established at 1453 on 26 February 1980 using normal makeup and letdown.

3.1.1.2 The flowrate of Reactor Coolant through the Reactor Coolant System shall be 2 2700 gpm whenever a reduction in Reactor Coolant System boron concentration is being made.

At 1453 on 26 February 1980 while re-establishing normal letdown and makeup flow to maintain Reactor Coolant system pressure control (on natural circulation) an apparent reduction in RCS boron concentration was equal to, or greater than the RCS boron following high pressure injection.

3.4.8 The specific activity of the primary coolant shall be limited to less than or equal to 1.0 microcuries per gram Dose Equivalent I-131.

At 0058 on 27 February 1980 it was discovered that the Reactor Coolant Dose Equivalent Iodine-131 was 1.45 microcuries per gra.s. The four hour sampling frequency was initited per the Action Statement re uirements. The sampling interval was terminated at 0110 on 28 February 198( -hen the DEI-131 was determined to be .755 microcuries per gram.

Reference the following Supplemental Information Data in regard to this event.

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l Analysis of the Occurrence:

The loss of electric power is an analyzed design basis event for the CR-3 FSAR. The following criteria for reactor protection are required for this event:

a) Fuel damage will not occur from an excessive power-to-flow rate.

b) Reactor Coolant system pressure will not exceed code pressure limits.

c) The resultant doses are within 10 CFR 100 limits.

The February 26, 1980 transient involved a partial loss of electrical power.

The applicable protection criteria appear to have been met throughout the event.

The sequence of events following the power supply failure in the NNI "X" cabinet is similar in nature to a loss of feedwater type transient whose consequences have been bounded for the FSAR analysis by the design basis event of a feedwater line break. The Reactor protection criteria for the feedwater line 13eak are:

a) The core shall remain intact for effective core cooling.

b) The Reactor Coolant System pressure shall not exceed code pressure limits of 110% of 2500 psig; i.e., 2750 psig.

The major difference between the February 26, 1980, event and the bounding feedwater line break event analyzed for the FSAR are itemized below:

1. The main feedwater valves were closed over a 25-30 second time interval rather than the conservative FSAR assumption of an instantaneous rupture of the main feedwater header at the steam generator inlet nozzles.
2. The combination of the opening of the PORV, the ICS demand for 103% full power and feedwater runbacks caused the Reactor trip on high pressure to occur at 10-25 seconds into the event rather than at 11.8 seconds for the FSAR event.
3. The opened PORV allowed a rapid depressurization to 1500 psi such that the ESFAS was actuated and two HPI pumps were started. The PORV and spray block valves were closed about five (5) minutes into the event and the RCS code safety valves actuated about 13 minutes into the event. This was much slower than 22.3 seconds for the FSAR events.
4. The operator initiated emergency feedwater at approximately nine (9) minutes into the event rather than the 15 minutes which was demonstrated for the FSAR as an acceptable time for operator action to prevent core or Reactor Coolant boundary damage.
5. Main feedwater was supplied "B" 0TSG at about four (4) minutes (approximately 1400 gpm) rather then being totally lost as analyzed in the FSAR.
6. The maintenance of full HPI flow for 32 minutes provided a significantly greater heat removal capacity than assumed in the FSAR where only relief through the safety valves of the expanding RCS system water is considered.

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6 Analysis of the Occurrence: (page 2)

For both events, core coverage was maintained, o fuel damage occurred and the Reactor Coolant System pressure remained within code allowable limits.

The safety evaluation criteria were met.

Although apparently initiated by a single fault, this transient involved a series of multiple, propagating abnormalities through the interaction of the

., various systems which are affected directly or indirectly by the interruption of the NNI "X" power source. These interactions led to:

a) Loss of some instrumentation and annunciator indications to the plant operator. Presentation of some confusing and invalid information b) Presentation of some invalid plant signals to the integrated control system, which acted upon them to increase Reactor power, decrease feedwater flow and increase steam flow.

c) The pressurizer PORV was opened and latched at a system pressure below its setpoint.

d) Automatic initiation of emergency feedwater did not occur, despite reduction of main feedwater.

The maintenance of plant conditions within the safety criteria required the action of automatic safeguards (Reactor trip, high pressure injection, Reactor Building isolation), safeguards features (safety relief valve), and/or operator action (start emergency feedwater) in response to items (b)(c), and (d). Item (a) reduced the operator's ability to respond.

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7 Corrective Action:

Corrective action as a result of this event was identified in our letter of 12 March 1980 to Mr. Harold Denton.

Failure Data: This is the first occurrence of this type report.

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Enclosure 1 (pg. 1)

REACTOR PO'ER HISTORY OF PRIOR FORTY-EIGHT HOURS ITEM 1 EVENT DATE: 27 FEBRUARY 1980 l

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DATE 2,25,80 L}

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$ 16 51.56 8856 45.96 60.20 1.310 1.122

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IE AVERAGE DAILY TURBINE GROSS HEAT RATE 9985 BTU /KUH

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e 6 87.37 9933 88.66 S3.80 1.058 .9.85

,y 7 S9.59 10416 95.68 99.20 1.037 .936 5 93.-39 7t31 75-70 99-SG 1.C37 .9?S

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Enclosure 1 (pg 4)

FUEL BURNUP BY CORE REGION ITEM 2 EVENT DATE: February 27, 1980 I

Enclosure 1 (pg 5)

ITEM 2 The burnup was calculated at 166.494 EFPD for the three (3) enrichment regions.

REGION NUMBER OF FA BURNUP B 61 20,046 MWD /MTU C 60 15,748 MWD /MTU D 56 4,434 MWD /MTU Avg. 13,650 MWD /MTU l

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Enclosure 1 (pg 6)

CLEANUP FLOW HISTORY ITEM 3 EVENT DATE: February 27, 1980

Enclosure 1 (pg 7)

ITEM 3 Cleanup flow history forty-eight (48) hours prior to the first sample in which the limit was exceeded indicated a letdown flowrate of forty-five (45) gpm.

DATE/ TIME F1.0WRATE 2/25/80 @ 1420 47 gpm 2/26/80 @ 0010 100 gpm 2/26/80 0 1415 47 gpm 2/26/80 @ 2135 54 gpm 2/27/80 @ 00.18 60 gpm

Enclocurs 1 (pg 8)

DEGAS OPERATIONS ITEM 4 EVENT DATE: February 27, 1980 NO DEGAS OPERATIONS.

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Enclosure 1 (pg 9) l l

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TIME DURATION WlEN DEI-131 EXCEEDED 1.0 uCi/ gram and I-131 ANALYSIS RESULTS ITEM 5 Event Date: February 27, 1980

Enclosuro 1 (pg 10) 4 l

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ITEM 5 -

l As per Technical Specification 3.4.8, the four (4) hour sampling frequency as depicted on the Table below was initiated at 0058 on 27 February 1980, and the Dose Equivalent I-131 was 1.45 microcuries per gram. The four hour sampling frequency was terminated at 0110 on 28 February 1980 when the DEI-131 was determined to be .755 micro-curies per gram. The time duration when the specific activity of primary coolant exceeded 1.0 microcuries per gram Dose Equivalent is 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> 12 minutes.

q DATE TIME DEI-131 ( Ci/ gram)  ;

2/27/80 0058 1.45 2/27/80 0615 1.1 2/27/80 0855 .979 2/27/80 1255 .944 2/27/80 1755 1.07 2/27/80 2110 .818 2/28/80- 0110 .755 2/28/80 0815 .620 2/28/80 2020 .526 4

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