ML20090D436

Ro:On 721119,unit Inadvertently Tripped During Testing of Turbine Thrust Bearing Wear Detector.Turbine Trip Caused Reactor Scram W/Inadvertent Opening of Safety Valve & Pressurization of Drywell.Caused by Operator Error

ML20090D436
Person / Time
Site:	Nine Mile Point
Issue date:	11/29/1972
From:	Burt P NIAGARA MOHAWK POWER CORP.
To:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20090D435	List: ML20090D436
References
6605, NUDOCS 8303010120
Download: ML20090D436 (15)

Text

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O O NIAGARA MOHAWK POWER CORPORATION NBAGARA MOHAWK Nine Mile Point Nuclear Station Post Office Box 32 Lycoming, New York 13093 November 29, 1972

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lf l ch ne-J lla (r Mr. Donald J. Skovholt Assistant Director for Reactor Operations b; DEC 41972

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Division of Reactor Licensing a - 't, United States Atomic Energy Commission A, -

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Washington, D. C. 20545 'A; sN\.

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Dear Mr. Skovholt:

Re: Provisional Operation License: DPR-17 Docket No.: 50-220 On November 19, 1972 at 0527:30, Nine Mile Point Unit #1 while operating at 1820 Di (t) was inadvertently tripped during the testing of the Turbine thrust bearing wear detector, a routine weekly test.

The resultant turbine trip caused a reactor scram with the inadvertant opening of a safety valve for 9-seconds and subsequent pressurization of the drywell to 2.9 psig.

During the transient the maximum reactor pressure attained was 1083 psig or 7 psi below the actuation pressure of the first electromatic relief valvo and 153 psi below the normal actuation pressure for the safety valve.

Feedwater control system performance during the transient was acceptabic, maintaining the reactor water level at a minimum of -3.15 feet and a maximum of 1.43 feet. Normal reactor water level is 0-1.S feet. As a result no flooding of reactor vessel nozzles occurred and all safety systems could have operated, if needed, in their normal manner. The operator response during the transient was consistant with procedures and feedwater transient instructions provided to maintain reactor water level within the prescribed range.

Detailed analysis and calculations based upon operating recorders and computer monitoring systems provided the following:

i

1. The turbine - generator system functioned as designed with turbine bypassf available to limit reactor pressure rise resulting from turbine stop valve closure, See Enclosure 1.

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COPY SENT RECIot'A3 rG05

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, Mr. Donald'3kovholt November 29, 1972 U. S. Atomic 1:nergy Comission

2. Safety Valve #6267 operated for 9 seconds releasing approximately 290 gals. of coolent into the drywell floor drain system. The operation of the safety valve limited reactor pressure to less than the operating pressure of the first electromatic relief valve which is designed to provide pressure relief in this type of transient. As a result no steam was released to the torus via the electromatic relief system. Using the Nitrogen test and correlation to steam relieving pressures, -

it was found that Safety valve #6267 relieved at approximately 1080 psig instead of its design of 1236 psig. No apparent reason for the early relief was forthcoming from on site inspection. The valve will be sent to Dresser Valve Company for a thorough analysis. Although it was not indicated that the safety valve had been inadvertently adjusted during the 1971 maintenance overhaul, all safety valves were scribed on the adjusting nut to indicate in the future any deviations.

Enclosure 4 includes the past history of each safety valve .

and the results of the on site nitrogen testing.

3., Damage' in the drywell was limited to the upper section surrounding the affected safety valve and concerned mainly the insulation on the valve and vent piping to the drywell coolers. This is Jas would be expected. No higher temperatures than 10-15*F above nornal full power operation were noted in the lower portion of the drywell. Enclosure 2 provides a detailed description of the gaseous releases to the drywell and to the environment.

4. No emergency systems actuation parameters were reached during the transient. Subsequent analysis of the transient and the parameters plotted for the safety analysis for a turbine trip at this power lovel showed no inconsistancies in parameter response and range considering the safety valve opening excepting the recirculation system response. See Enclosure 3.

5. The recirculation system indicated a minimum flow of approximately 6 x 106 #/hr. 45 seconds following the turbine trip. This occurred with increasing reactor water level, decreasing (approximately zero) feedwater flow, and decreasing reactor pressure.

It would appear that void formation in the recirculation loops and/or flow sensing lines caused this phenomena. Collected data has been forwarded to General Electric for analysis and solution to this probica. The reduction in flow occurred over an 18 sec.

time span and presented no safety problem in as much as the reactor had already been shutdown.

l 6. The turbine trip was caused by the failure of a micro-switch l in the turbine thrust bearing wear detector test circuit. The micro-switch, which is activated when the wear detector mode switch is placed in the " test" position, blocks the thrust bearing wear detector turbine trip signal and provides an annunciator alam. The failure of the micro-switch to function enabled the turbine trip and did not alarm the annunciator.

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O Nevstbar 29, 1972 Mr. Donald Skcvholt I U. S. Atomic Energy Commission l

6. 'Iho operator did not recogni:o the significance of the lack '

of the annunciator and when attempting to proceed with the ,

test, tripped the turbine.

Following a thorough review of the transient the following corrective actions were taken prior to start-up of the reactor.

1. To provide the most reliable set of safety valves for the reactor.

a. The 7 existing spare safoty valvos that had recently been steam tested were installed matching relieving pressure for relieving pressure (including the safety -

valve that relieved early).

t

b. The remaining 9 safety valves on the head were nitrogen tested and correlated to steam relieving pressures.

Those that met the criteria were placed back on the vessel head,

c. Two valves were needed frcm the 7 safety valves initially removed. Those were nitrogen tested and correlated to steam providing 16 reliable valves to be installed on the vessel head.

. 2. To eliminate the possible reoccurance of a trip resulting from the thrust bearing wear detector test,

a. A circuit was installed which will provide a light at the bearing wear detector test switch in the control room to indicate operation of the micro-switch,

b. The micro-switch that failed was replaced.

3. To determine the effect of the transient on other operating systems, and the drywell,

a. The transient results vere compared with the safety analysis to assure that the response of all systems to the transient was in agreement with design, considering the inadvertent operation of a safety valve,

b. General Electric Company was provided with sufficient data to determine what can be done to eliminate the void effect seen in the recirculation system loop and/or flow instrumentation lines.

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. 4 .

l Mr. Donald Sksyholt

. Nsvem 29, 1972 U. S. Atomic Energy Commission

3. c. Electrical equipment in the drywell, including cable insulation was inspected for possible damage. No damage was found,

d. Damage occurring in and around the safety valve discharge was repaired.

e. A post-start-up hydrostatic test at 1000 psig was performed.

4. An analysis and review of these corrective actions was undertaken by the Site Operations Review Committee prior to Station start-up.

Results of the review by the Site Operations Review Committee 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to station start-up showed:

1. That no undue safety or radiological hazard was presented to the general public.

2. No unreviewed safoty question exists and all systems with the exception of those now under study as previously mentionod, performed their intended function in their designed manner.

3. All safety valves would in the future be required to

. have a nitrogen correlation particular to that valve, to enable testing just prior to the installation on the vessel head.

Very truly yours, b f P. Allister Burt General Superintendent Nuclear Generation PAB/cm Enclosures.

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ENCLOSURE 1 Tho turbine-generator system operation is based upon the ability of the bypass system to operato upon control valvo closing thus limiting the pressuro in the reactor vossol. Following closure of the turbino stop valvos the bypass system should bo in operation within .3 seconds.

A detailed analysis of the system oporation requires a distinction between the transient timo (in millisoconds) and computer time (in seconds) and the correlation between the scanning frequency of both.

An analog point is scanned normally in a prescribed group sequence, that is, the software selects an analog point on each one of the analog switch matrix terminations and scans these points as a group. Therefore ,

knowing the hardware timing limits and the software conversion timing it is possible to tell the approximate time within a one second time frame that the individual point was scanned. In addition a check can be made upon this value using the API priority structure. The Automatic

. Priority Interrupt can recognize an event occurrance within .5u sec.

Therefore knowing the time in milliseconds that a particular parameter exceeded its alarm point (such as in the sequence of events which are all API generated) a back fitting to the post mortem log can be made to detemine the fraction of the second that the parameter exceeded its alarm value. The sequence of event log indicates that the computer timo changed from 5:27:31 to 5:27:32 between 1369 ms and 1401 ms into the transient. Therefore it is exactly correct to say that the trip occurred at 5:27:30 and 599-631 ms. The reactor high pressure sequence of event point went into alarm (1080) 796 ms into the transient or at 5:27:31 and 395 ms computer timo. On the post mortem log wide range reactor pressure (which was scanned at 5:27:31 and approx. 400 ms) indicated

, abovo 1080 psig where as the narrow range reactor pressure (which was scanned at 5:27:31 and approx. O ms) indicated 1013 psig. Therefore by determining the computer time in us that the analog point was read, a propor analysis of the post mortem can be mado. Applying this same analysis to the bypass valve position, shows that the bypass valves are fully open within the 5:27:31 second time frame.

As an independent verification of this fact, the transient recorder shows the bypass valves 100% open approximately .3 seconds following closure of the stop valves. Comparing the above results with the safety analysis for turbine trip with failure of the bypass system, Neutron flux would peak at 163%, .83 seconds following the trip. No alare.s ,

indicating neutron flux exceeded its trip point of 120%, were in evidence during the November 19th transient, a further indication of proper action of the bypass system.

.' ENCLOSURE 2

1. Total fission gases rolensed from 0530 hrs.-----------.> 2400 hrs. 714 _ _C1.

2. Average release rate fission gases from 0530 hrs.------------> 2400 hrs. 10,715 pCi/sec.

3. Average release rate during drywell purge 0900 hrs.----------> 1100 hrs. 54,120 pCi/sec.

4. Max. release rate during drywell purge.

time 0915 hrs. 131,350 pCi/sec.

5. I released during drywell purgo from 0900 hrs.---------- > 1100 hrs. 6.5 x 10-4 Ci.

6. I release rate during drywell purge.

0900 Hrs . -------- > 1100 Hrs. 9 x 10-2 pCi/sec.

7. Fission gases concentration in drywell before purging N 1 x 102 Ci/cc

8. I Concentration in drywell before purging s 2 x 10-8 pCi/cc

9. Wind direction was from the south east. Velocity approx.

20 MPH. e

O O ENCLOSURE 3 Figure 1 shows the vital parameter changes that occurred as a result of the turbine trip from 1820 M1 (t) on November 19, 1972.

The comparison with the safety analysis (submitted February 28, 1972) shows an acceptable transient with no significant variation of measured values of thermal, nucicar or hydraulic characteristics from the predicted safoty analysis considering the safety valve operation. The rceirculation flow reduction, the suspected result of void generation in the loops and/or flow sensing lines, is being analyzed by General Electric and Nine Milo Point Site Personnel. In any event the change in recirculation flow is beyond the time when it would be considered significant to the safety analysis (16 seconds) . The excellent response of the bypass system to limit the resultant reactor pressure spike and corresponding neutron flux peak to better than acceptable values (less than 105% on the APRM's) indicates the effort to the " Fine Tune" this system in the past, paid dividends during the transient.

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. . O O ENCLOSURE 4 The following is a history of the 16 safety valves residing on the reactor vessel head during the transient of November 19, 1972, plus spares.

Following the history is the data, generated as a result of the nitrogen testing program. Figure 1 shows the present safety valve alignment and Figure 2 shows the nitrogen correlation used to determine the steam relieving pressures.

1. 6250 Pressure: 1218 lbs.

1. 1969 Installed Position N7F.

2. 1971 Removed, dismantled, cleaned, lapped seat and disc. X-rayed disc.

3. 1972 Installed Position N7J.

1. 6253 Pressure: 1227 lbs.

1. 1971 Installed Position N7D.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc. reassembled -

and installed position N7D.

1. 6254 Pressure: 1245 lbs.

1. 1969 Installed Position N7N.

2. 1971 Removed, dismantled, cleaned, lapped seat and disc. Reassembled and installed position N7N.

3. 1972 Removed, dismantled, cleaned, lapped seat and disc. Dye checked nozzle. Reassembled and installed Position N7N. .

1. 6255 Pressure: 1236 lbs.

1. 1971 Installed Position N7C.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc. Reassembled.

Shipped to Dresser Industries to check popping pressure. Also to test with nitrogen. .

1. 6256 Pressure: 1236 lbs.

1971 Installed Position N7E.

1.

2. 1972 Removed, dismantled, cleaned, lapped seat 6 disc. Reassembled and installed position N7E.

1. 6267 Pressure: 1236 lbs. I

1. 1969 Installed position N7K. l-

2. 1971 Removed, dismantled, cleaned, lapped seat and disc. Reassembled and installed position N7K.

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1. 6280 Pressure: 1236 lbs.

1. 1969 Installed Position N7E.

2. 1971 Removed, dismantled, cicaned, lapped sent 6 disc. Reassembled and installed position N7J.

3. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Dye checked nozzle. Reassembled.

Shipped to Dresser Industries to check popping pressure.

Also to test with nitrogen.

1. 6291- Pressure: 1227 lbs.

1. 1971 Installed position N7G.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Reassembled and installed Position N7G.

1. 6292 Pressure: 1254 lbs.

1. 1969 Installed Position N7T.

2. 1971 Removed, dismantled, cleaned, lapped seat and disc.

3. 1972 Dismantled, Dye checked nozzle reassembled.

Shipped to Dresser Industries to check popping pressure.

Alsc, to test with nitrogen.

1. 6297 Pressure: 1245 lbs.

. 1. 1969 Installed Position N7S.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Reassembled and installed Position N7S.

1. 6298 Pressure: 1254 lbs.

1. 1969 Installed Position N7U.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Reassembled and installed Position N7U.

1. 6301 Pressure: 1254 lbs.

1. 1969 Installed Position N7R.

2. 1972 Removed, dismantled, cleaned, lapped sent and disc.

Dye checked nozzle. Reassembled and installed position N7R.

l #6303 Pressure: 1254 lbs.

1. 1971 Installed position N7T.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Reassembled and installed position NTr.

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1. 6313 Pressure: 1245 lbs.

1. 1969 Installed Position N7H.

2. 1971 Removed, dismantled, cleaned, lapped seat and disc.

X-rayed disc. Reassembled.

3. 1972. Installed Position N7H.

1. 6316 Pressure: 1245 lbs.

1. 1971 Installed Position N7H.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Dye checked nozzle reassembled.

Shipped to Dresser Industries to check popping pressure.

Also to check with nitrogen.

1. 6317 Pressure: 1227 lbs.

1. 1969 Installed Position N7D.

2. 1971 Removed, dismantled, cleaned, lapped seat 6 disc.

3. 1972 Dismantled, dye checked nozzle.

Shipped to Dresser Industries July 12 for testing popping pressure. Also to check with nitrogen.

1. 6319 Pressure: 1227 lbs.

l

, 1. 1969 Installed Position N7G.

2. 1971 Removed, dismantled, cleaned, lapped seat and disc.

3. 1972 Dismantled dye checked nozzle reassembled.

Shipped to Dresser Industries for checking popping pressure.

Also to be tested with nitrogen.

1. 6325 Pressure: 1227 lbs.

1. 1969 Installed Position N7B.

2. 1971 Dismantled, cleaned, lapped seat and Disc.

Reassembled and Reinstalled Position N7B.

3. 1972 Dis..tantled, cleaned, lapped seat and disc.

Reassembled and Reinstalled Position N7B.

1. 6520 Pressure: 1218 lbs.

1. 1969 Installed Position N7J.

2. 1971 Removed, dismantled, cicaned, lapped seat and disc.

3. 1972 Dismantled dye checked nozzle - Reassembled and installed us position N7A.

1. 6521 Pressure: 1218 lbs.

1. 1970 Installed Position N7A for increase in power.

2. 1972 Removed, dismantled, cleaned, lapped seat and disc.,

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1. 6522 Pressure: 1218 lbs.

s .

1. 1971 Installed Position N7F.

2. 1972 Dismantled, cleaned, lapped seat and disc. Reassembled and reinstalled position N7F.

1. 6524 Pressure: 1218 lbs.

1. 1969 Installed Position N7M.

2. 1971 Removed, dismantled, cleaned, lapped seat and disc.

Roassembled and Reinstalled Position N7M.

3. 1972 Removed, dismantled, cleaned, lapped seat and disc.

Reassembled and Reinstalled Position N7M.

1. 6535 Pressure: 1236 lbs.

1. 1969 Installed Position N7C.

2. 1971 Removed, dismantled, cleaned, lapped seat and Disc.

3. 1972 Dismantled dye checked nozzle.

Reassembled and Insta11ee Position N7C.

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. . O TESr oxTi O Nov. 21, 1972 TEST 1 TEST 2 TEST 3 VALVE SET N2 N2 N2 S/N PSIG PSIG A PSIG A PSIG A 6267* 1236 1025 211 1027 209 1026 210 6267* 1236 1018 218 1020 216 6256 1236 1150 86 1135 101 1140 .96 6254** 1245 1 215 30 1210 35 1205 40 6301 1254 1 155 99 1150 104 1145 109 6325 1227 1 150 77 1125 1 02 1125 102 6524 1218 1 140 78 1105 113 1 105 113 6291 1227 1165 62 1160 67 1 160 67 6267* 1236 1 025 211 1025 211 Nov. 22, 1972 TEST 1 TEST 2 TEST 3 N2 2

VALVE SET N2 N S/N PSIG PSIG A PSIG A PSIG A 6520 1218 1135 8! 1135 83 1135 83 6313 1245 1185 60 1175 70 1 175 70 6250** 1218 -1215 3 1185 33 1190 28 6522 1218 1125 93 1 100 118 1100 118 6298 1254 1 170 84 1180 74 1 175 79 6297 1245 1165 80 1165 80 1160 85 6267* 1236 1050 186 1 025 211 1025 211

• Safety Valve that Relieved Early.

• • Valves Rejected on N2 Testing.

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