ML19322B481

Description of Cause & Correction of Damage to Control Rod Drive Mechanisms During Preoperational Testing of Oconee 1.

ML19322B481
Person / Time
Site:	Oconee
Issue date:	09/29/1971
From:	DUKE POWER CO.
To:
References
NUDOCS 7912030319
Download: ML19322B481 (50)

Text

o 7 DUKE P0WER C0MPANY OCONEE NUCLEAR STATION UNIT NO. 1 DESCRIPTION OF CAUSE AND CORRECTION  !

0F DAMAGE TO CONTROL ROD DRIVE MECHANISMS l DURING PREOPERATIONAL TESTING OF OCONEE NO. 1 l

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• Docket No. 50-269 Septeraber 29, 1971 7912030 3;/9 A

f I CONTENTS I. Summary Description of Incident II. Designation of Apparent Causes III. Plant Conditions at Time of Incident IV. Detailed Description of Incident V. Corrective Action A. Identification, Repair, and Replacement of Damaged Mechanisms B. Operating Procedures C. Plant Design

. Startup Test Program VI. Potential for Reoccurrence I

Figures 4

1. Reactor Coolant System Pressure History (5/21/71 to 6/14/71)

2. Modified (Center) CRD Closure Insert Assembly Tables

1. Summary of Control Rod Drive Operations (5/27/71 to 6/12/71)

2. Chronological Summary of Plant Operations (5/20/71 to 6/13/71)

3. Reactor Coolant Pump Operation and Spray Flow Summary l

APPENDIX A Control Rod Drive Mechanism Damage Evaluation Y

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1 DUKE POWER COMPAhT OCONEE NUCLEAR STATION UNIT NO. 1 DESCRIPTION OF CAUSE AND CORRECTION 0F DAMAGE TO CONTROL ROD DRIVE MECliANISMS DURING PREOPERATIONAL TESTING OF OCONEE NO.1 I. Summary Deceription of Incident Between May 27, 1971 and June 11, 1971 eleven (11) control rod drive mechanisms sustained internal damage when they were tripped with their hydraulic snubber regions either completely dry or only partially filled with water.

II. Designation of Apparent Causes A. Primary The primary cause of this incident was inadequate operating procedures for the existing conditions in that these procedures failed to incorporate basic precautions necessary to prevent gas accumulation under the reactor vessel head during certain preoperational tests.

B. Secondary The secondary cause of this incident was operator error in that con-sequences of unusual conditions prevailing during the incident period were not recognized and corrective action taken.

III. Plant Conditions at Time of Incident The reactor coolant system (RCS) was filled and (thought to be) vented.

Except during the RCS hydrostatic test, coolant pressure was maintained by loading the pressurizer with nitrogen in order to avoid solid water operation; pressurizer heaters were not available for installation.

Cold functional testing was in progress with the following major events occurring over the period of interest: Fill end venting of the RCS, Initial Reactor Coolant Pump Operations, RCS Cold Hydrostatic Test,

-Control Rod Drive (CRD) Minimum Latch-Run-Drop Current Tests, and cleanup operations to remove oil from the RCS, IV. Detailed Description of Incident Refer to Table 1 for. a summary of control rod drive (CRD) operations between May 27, 1971 and June 12, 1971.

The first indication of improper CRD operating conditions was received on June 9 when CRD G-9 was tripped for the first time (from approximately

, 1 35 percent withdrawal). Malfunction was indicated when this CRD failed to indicate withdrawal following trip.

On June ll, CRDs in Groups 1, 2, 3, 4 were tripped in groups from 100 percent withdrawal as part of the oil removal operation. Groups 1 and 2 operated normally; however, after tripping Group 3, trip time traces were obtained during a second trip to help explain why CRDs H-10, L-8, and H-6 would not relatch. The trace for CRD F-8 indicated a short trip time followed by a rebound of about two feet; however, subsequent operation of F-8 appeared normal. Group 4 CRDs were then tripped and trip time traces obtained; CRD H-8 experienced a fast trip and failed to relatch. At this time, the possibility of dry trip was suspected, and no further trip operations were conducted.

On June 13, coolant pressure was lowered to 60 psig and the CRDs were vented. Approximately 325 SCF of nitrogen (all of the vented gas) was collected verifying that several CRDs had been tripped with a significant quantity of gas present.

The RCS was then depressurized and drained allowing preliminary visual in-spection of all CRDS. This inspection revealed eleven damaged shim / safety drives and confirmed minor damage to four axial power shaping drives which had been difficult to uncouple prior to reactor vessel head removal. (The axial power shaping drives were damaged by the coupling tool during coupling /

uncoupling operations; the coupling tool has been redesigned to correct the deficiency.) As a result of this initial inspection, 19 CRDs (including four und maged shim / safety drives) were returned to the manufacturer for in-spection and analysis. Later, after inspection using formal criteria developed by the manufacturer, six shim / safety drives were removed from the reactor vessel head because of slight marking of the leadscrew upper extension.

Details of the manufacturer's inspections, analyses and findings are presented in Appendix A.

Plant evolutions prior to and during the period of CRD operations are summarized in Table 2, and reactor coolant system pressure history during the period is shown in Figure 1. The condition which precipitated the incident, i.e., accumulation of gas under the reactor vessel head, was generated by several events acting cither individually or in combination.

These events are highlighted in the following paragraphs.

Prior to RCS fill, the pressurizer spray valve was opened as part of the valve lineup. The sprav valve remained open for the duration of fill and during all subsequent operations.

Following RCS filling and venting, the pressurizer was loaded with nitrogen to establish the required pressure (approximately 400 psig) for reactor coolant pump operation. This mode was chosen because pressurizer heaters were still in the manufacturing process and solid water operation was not considered desirable. The RCS was not revented following initial reactor coolant pump operation because there was no requirement in the procedure being followed to do so. Failure to revent the system after circulation is considered a major contributory factor to the incident.

Pump testing was followed by system heatup to satisfy brittle fracture limitations in preparation for the RCS cold hydrostatic test. Coolant pump Al was used for this purpose and was the only pump operated following initial pump. testing. In further preparation for hydrostatic test, the nitrogen bubble in the pressurizer was collapsed allowing the RCS to become water solid. Shortly thereafter the RCS was depressurized. Prior to depressurization, coolant pump Al had been operated a total of 510 minutes circulating spray through the 400 psig nitrogen atmosphere in the pressurizer at about 170 gpm thereby producing an integrated flow of nearly one system volume (Refer to Table 3). If the worst case is assumed, i.e., that the RCS had become saturated with nitrogen, then some 3,500 SCF would have come out of solution upon depressurization to atmospheric pressure. (1) The actual situation was most likely one of only partial saturation. Nonetheless, circulation of spray through the nitrogen atmosphere at 400 psig followed by depressurization is considered as yet another contributory factor to the incident.

Upon completion of the RCS hydrostatic test, nitrogen pressure control was reestablished, and a trace quantity of oil was discovered in the RCS -

the source of which was determined to be the hydro test pump.

Cleanup operations consisted of adding a chemical agent (Triton X-100) followed by recirculation and feed and bleed. During this period, but prior to further coolant pump operation, a second depressurization took place. CRD operation as part of the cleanup procedure has been described previously; prior to CRD venting on June 13, a third depressurization took place on June 12 when the pressurizer power-operated relief valve was inadvertently opened. Between this depressurization and the second, coolant pump Al had been operated for 715 minutes resulting in an additional integrated flow of 1.3 system volumes through the 400 psig nitrogen bubble in the pressurizer.

V. Corrective Action A. Identification, Repair, and Replacement of Damaged CRDs All control rod drives were inspected either by the manufacturer or in place using criteria furnished by the manufacturer. Drives meeting any of the several criteria for identifying potential damage were re-moved from the reactor vessel head and returned to the manufacturer for comprehensive evaluation. Those drives removed were replaced by Unit 2 equipment. Appendix A describes in detail the manufacturer's inspection criteria, the extent and type of damage sustained by each mechanism, and the action necessary to restore each mechanism to its original condition.

B. Operating Procedures The following changes / additions to station operating procedures are being made to correct known deficiencies and to improve operating (1) Based on the solubility of nitrogen in water at 120*F from THE HANDBOOK OF CHEMISTRY, (Ninth Edition, 1956) assuming 12,000 ft3 RCS water volume.

techniques as regards accumulation of gas in the control rod drives.

1. Require venting the center CRD and as many additional drives es prac.tical during initial stages of RCS fill when the reactor vessel is filling.

2. Require venting the RCS from the loop high points and control rod drives immediately after initial coolant pump operation following fill. ,

3. Review the RCS fill procedure to assure proper isolation of connecting systems and to assure that lines discharging to the RCS are already filled, will be filled during RCS fill, or can be independently filled and vented up to the RCS boundary isolation valve.

4. Set a limit on the maximum permissible total gas concentration in the reactor coolant (tentatively 100 cc/kg).

5. Corresponding to Item 4, set a limit on minimum coolant pressure (as a function of temperature) such that gases present up to the limiting concentration will remain in solution.

6. Define those plant conditions and occurrences requiring that the RCS be revented to prevent gas accumulation in the reactor vessel head area; violation of the limits defined in Items 4 and 5 will be included.

7. Prohibit CRD operation, except as required for plant safety, when the reactor coolant system does not meet venting /reventing requirements.

C. Design Changes The following design changes will be made to allow collection of a cooled, pressurized reactor coolant sample and to facilitate control rad drive venting.

1. Add a sample connection in the coolant letdown line between the letdown coolers and the pressure reducing block orifice.

2. Provide new CRD vent tools to permit venting above the minimum pressure required for coolant pump operation.

3. Provide a special CRD closure insert assembly as shown in Figure 2 to permit venting the center CRD under all plant operating conditions.

4. Provide a suitable apparatus to allow quantitative measurement of gas which might accumulate in the center CRD during hot functional and power testing.

D. Startup Test Program

1. During the hot functional and power testing the center CRD will be monitored to detect any buildup of gas.

I

2. As a precautionary measure, planned trips of the center CRD will be monitored to obtain trip time traces in order to detect un-usual operating conditions.

VI. Potential for Reoccurrence

In order to evaluate the likelihood of dry _ trip reoccurrence, two questions need to be considered

(a) What caused the gas accumulation in the CRDs,

. and (b) Will the proposed corrective action prevent gas accumulation?

, Clearly the adverse environmental conditions under which several CRDs were tripped is attributable to one or more of the following causes:

1. Improper RCS venting, in that venting was not performed following coolant pump operation.

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2. Improper use of nitrogen to pressurize the reactor coolant system in that the pressurizer spray valve remained open for a prolonged period i without the means or the requirement to measure coolant total gas content, thereby allowing a large quantity of nitrogen to become dissolved during coolant pump operation.

3. Failure to vent the reactor coolant system following repeated de-i pressurization thereby allowing gas escaping from' solution to collect in part under the reactor vessel head.

I Reoccurrence of these errors will be prevented by adherence to operating procedures, revised as stated in Section V.B.

As a further measure, the center control rod drive will be monitored during startup as stated in Section V.D to detect otherwise unexplained buildup of gas under the reactor vessel head; if such buildup is shown, appropriate methods will be incorporated to protect the control rod drives during extended power operation.

In summary,-Duke Power Company submits that (1) the causes of this in-

cident have been determined, (2) adequate corrective measures have been

taken, and (3) the potential for reoccurrence is correspondingly slight.

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4 TABLE 1 St?t1 ART OF CONTROL ROD DRIVE OPERATION (MAT 27, 1971 - J1TNE 12. 1971) Sheet 1 of 4 1:RD Identificat ton Core Date 40 2 . I,oc. CRF 5/27 5/31 6/1 6/2 6/8 6/9 6/10 6/11 -

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17 K-Il 1 SEV. T-10 3T-10. 1T-35 2C-100,17-100 j 15 E-9 i SEV. T-10 10T-10, 4T-35 2C-100. IT-100 16 C-11 1 3T-10, 17-35 2C-100. IT-Ino 14 E-7 1 3T-10. IT-35 2C-100. IT-100 18 31 - 9 1 3T-10. IT-35 1C-100. IT-100 21 C-5 1 3T-10. 1T-35 1C-100.17-100  !

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66 P-10 2 3T-10. 17-35 2C-100. IT-100 23 H-12 2 3T-10, IT-35 2C-100. IT-100 24 N-8 2 3T-10. IT-35 2C-100. IT-100 62 B-6 2 3T-10, IT-35 2C-100. 1T-100 67 P-6 2 3T-10. 17-35 IC-100. IT-100 69 F-2 2 3T-10, IT-35 IC-100 IT-100 25 H-4 2 3T-10, IT-35 IC-100, IT-100 68 L-2 2 3T-10, IT-35 IC-100, 1T-100 I

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CRD Identification Core .Date Nor, Loc. CRP 5/27 5/31 6/1 6/2 6/8 6/9 6/10 6/11 6/12 13 L-6 ,7 3T-10, IT-35 IC-100 49 N-4 7 3T-10. 17-35 1C-100 Legend: First Digit = Number of Occurrences (SEV.' = Several)

T = Trip C = 4ycle Last Digit = Percent of full withdrawal at which CRD was tripped or to which CRD was cycled.

Notes: (1) Trip Time Trace obtained (Where two trips were performed, the trace was obtained during the second trip.)

(2) CRD Malfunction Af ter Trip (3) Torque Taker Tangs Sheared off (4) CRD Failed to Latch (5) Fast Trip Time Indicated (6) Torque Taker Tangs Bent (7) No Operational Signs of Damage (8) only Slightly Damaged s

6- - __ . _ _ _ _ _ _ _ _ _ _ _ _ _ - . _

TABLE 2 CHRONOLOGICAL

SUMMARY

OF PLANT OPERATIONS (5/20/71 TO 6/13/71)

May 20, 1971 Commenced initial RCS fill

. May 21, 1971 Completed CRD venting (static head)

Completed venting loop high points (25 psig N2)

Depressurized to repair misc. leaks May 22 - 24, 1971 Maintenance (leak repairs)

May 25, 1971 Pressurize RCS to 25 psig May 26, 1971 Completed venting loop high points Completed venting CRD's Increase pressure for initial RCP operation May 27, 1971 Depressurize to repair RCP A2 seal bypass flange Resume pressurization for RCP operation Commence CRD operations for minimum latch-run-drop current testing -

refer to table 1.

May 28, 1971

, RCS pressure at 410 psig Completed one (1) minute run of each RCP Pump Al run for 13 minutes-No CRD operations May 29, 1971

1. Completed.one (1) hour.run. tests on each RCP

2. Pump Al'run.for 72 minutes

3. No CRD operations

May 30, 1971

1. Commence heatup for hydro (Pump A1)

2. Pump Al run for 218 minutes

3. No CRD operations May 31, 1971

1. Pump Al run for 207 minutes (prior to solid plant)

2. Collapsed N2 bubble in pressurizer - RCS in solid water condition

3. CRD operations per Table 1.

June 1, 1971

1. RCS depressurized while on solid water at approximately 400 psig.

Approximately one (1) RCS volume sprayed through No bubble at approximately 400 psig prior to this depressurizatIon.

Total run time Pump A1 = 510 minutes

2. Hydro test at 3125 psig completed

3. CRD operations per Table 1

4. No pump operations June 2, 1971

1. RCS pressure reduced to 400 psig and N9 bubble reestablished in

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2. CRD operations per Table 1.

3. No pump operations June 3, 1971

1. Presence of oil in RCS confirmed

2. Suspended CRD minimum latch-run-drop current testing

3. Begin cleanup operations (feed & bleed). Lowered pressure to 200 psig.

4. No pump operations June 4, 1971

1. Continue cleanup

2. Increased pressure to 400 psig

3. Depressurization. No pump operations since last depressurization on June 1.

4. Increased pressure to 400 psig

5. No pump operations

6. No CRD operations June 5, 1971 ~l

1. Pump Al run for 103 minutes .

2. Continue cleanup  !

3. No CRD operations

.l June 6, 1971 4

1. Pump Al run for 220 minutes

2. Continue cleanup

'3. No CRD ene acions

. 4. Commence ..ydro S.G. "B" i.

June 7, 1971

1. No pump operations

2. No CRD operations

3. Continue cleanup

4. Continue hydro S.G. "B" June 8, 1971

1. Complete hydro S.G. "B"

2. Resume CRD minimum latch-run-drop current testinr CRD operations per Table 1 (CRD F-8 would not latch.)

l 3. No pump operations-i June 9, 1971

1. Fump Al run for 208 minutes

2. CRD operations per Table 1 - CRD G-9 was inoperable following 35% trip

3. Continue cleanup June 10, 1971

1. Pump Al run for 184 minutes

2. Hydro tested S.G. "A"

3. CRD operations per Table 1 (CRD F-8 successfully latched) Completed CRD minimum latch-run-drop testing

4. Continue cleanup

. June 11, 1971

1. No pump operations

2. Continued cleanup operations - cycled Groups 1, 2, 3, 4 CRD's (by Group) and conducted 100% trips (by Group) per Table 1. CRD's H-10 , H-6 and L-8 would not relatch; therefore obtained trip time traces on several rods per Table 1. Fast trip times noted on CRD's F-8 and H-8.

I 3. Gas suspected in CRD's.

June 12, 1971 l

1. No pump operations

2. CRD operations per Table 1; no trips. Depressurization due to inadvertent opening of pressurizer. power relief valve.

3. Pump Al run for 715 minutes. Since last depressurization on June 4 sprayed an additional 1.3 RCS volumes through N2 bubble at 400 psig.

June 13, 1971

1. No. pump operations

2. Vented CRD's at 60 psig RCS-pressure - 325 SCF collected I

TABLE 3 REACTOR COOLANT PUMP OPERATION AND SPRAY FLOW

SUMMARY

Pump Al Total Spray Flow Period Run Time (min) (Gal) (RCS Volumes)

Initial Pump Operation to First Depressurization 510 86,700 0.97 i

I Initial Pump Operation to Second Depressurization 510 86,700 0.97 ,

Initial Pump Operation to Third Depressurization 1225 208,250 1.3 NOTES: (1) Spray flow with RCP Al running = 170 GPM (2) Assumes 12,000 ft3 RCS water volume i

APPENDIX A CONTROL ROD DRIVE MECHANISM (CRDM) DAMAGE EVALUATION I. General Reasons for CRDM Removal and Damage Summarv A. Eleven (11) shim safety drives were initially removed from the Oconee 1 head because they showed evidence of damage from being tripped with the snubber region either completely dry or only partially filled with water. This condition would prevent hydraulic snubbing action, thereby producing excessive deceleration of the drive line to a point of causing damage to the mechanism. The following is a brief de-scription of the damage to these eleven drives; detailed inspection results are contained in the attached CRDM Inspection Data Sheets.

1 Five (5) drives had the torque cap tangs sheared off (G-9, H-10, L-8, H-6, H-8). These drives will have a number of internal drive components replaced and other reworked to the original condition.

2. Four (4) drives had the torque tube cap tangs bent (G-7, K-9, K-7, F-8). In these drives, a lesser number of internal drive com-ponents will have to be replaced and/or reworked.

3. Two (2) drives had brinelling of the support ring load bearing surface with a very low degree of internal damage (D-8, B-10).

4. Three (3) of the above drives also showed evidence of coupling damage (K-7, F-8, G-9). The improper coupling damage was dis-covered after the drives were removed from the vessel head.

B. Three (3) shim safety drives were removed from the vessel head to gain access which permitted working on the level of the reactor vessel head and thus expedite drive replacement. One drive was removed because of indications of a loose dummy guide assembly support plate. These four mechanisms (F-10, C-9, E-9, H-14) did not exhibit any damage and were subsequently used to verify the damage criteria developed by DPSC.

C. Six (6) shim s,afety drives (M-7, K-5, M-5, D-4, B-8, F-14) were found to have abnormal markings on the upper leadscrew extensions and male couplings. Detailed examination revealed that raised metal (possibly due to handling during installation) on the inside diameter of the torque takers caused the markings, and was not the result of a " dry trip." These drives were removed from cne head at a later time as a result of the inspection criteria developed by DPSC, i.e., (to look for any unusual or abnormal markings on accessible components). The abnormal markings observed would not affect normal drive operation.

D. Four (4) axial power shaping drives (D-6, N-6, L-4, F-4) were removed because of binding of the leadscrew nut which would not allow normal i A-1 l

l

coupling and uncoupling of the CRDM leadscrew with the dummy guide assembly. The damage consisted of slight metal upset on the upper leadscrew extension, male coupling, and torque taker assembly. This damage.was caused by the coupling tool which has since been modified.

II. Drive Inspection Acceptance Criteria A. General Mechanisms identified in I.A and I.B (above) were returned to the manufacturer, Diamond Power Specialty Company (DPSC), to see if criteria could be developed which would enable an external inspection to verify whether or not any damage had occurred to the remaining mechanisms.

A criteria was developed by DPSC from: (1) detailed inspection of four (4) CRDM's in their own shop; (2) preliminary inspections of other drives that were taken off the Unit I head at the Oconee Site; and (3) an " order of failure" stress analysis. This investigation revealed a standard progressive mode of failure for a drive that undergoes a dry trip of any degree of severity. Certain parts of the drive, available for inspection without renoval of the drive from the head, will fail or show abnormal loadings in a progressive mode as the kinetic energy of the falling drive line is absorbed by the internal drive members.

B. Inspection Procedure and Criteria The following procedure and criteria were developed by DPSC and formed the basis for inspection of remaining shim-safety drives on the Oconee 1 head.

1. Remove and inspect the 703291-1146 support ring for brinelling of the surface contacted by the 703277-1134 torque tube cap tangs.

2. Inspect the surface of the 703227-1152 motor tube that is con-tacted by the 703291-1146 support ring for brinelling of the support !

ring into the motor tube surface.

3. Inspect the closure insert, P/N 705039-1106, uon the diameter that pilots into the 703277-1134 torque tube cap for brinelling of l the' closure insert from the tangs on the torque tube cap.

4. Inspect each leadscrew assembly, P/N 703267-1054, for damage.

In particular the male coupling shall be inspected for abnormal markings and the upper leadscrew extension shall be inspected in the area where the 703278-1044 torque taker fits.

A-2 i

l l

Criteria:

If brinelling of any of the surfaces outlined in Items 1, 2, or 3 is seen or if the leadscrew assembly is abnormally marked, the drive shall be considered to have been damaged and will require removal from the head for disassembly and complete inspection.

If no brinelling of the surfaces outlined in 1, 2, or 3 has occurred and no abnormal marking on the lead screw assembly is noted, then the drive is considered acceptable for use, and may be left on the vessel head.

C. Verification of Criteria These criteria were verified by DPSC by applying them to a number of drives shipped from the Oconee Site that had not previously undergone a detailed inspection. In all cases where the criteria inspection showed no signs of external damage, the detailed in-spection of internals also showed no evidence of dry trip damage.

Upon this basis, the criteria for leaving drives on the head was shown to be valid.

III. CRDM Testing with Gas Volume in Motor Tube A. Description of Test To further substantiate the inspection criteria and thereby demon-strate that most of the Oconee 1 CRDM's had not been subject to excessive deceleration forces, a test was conducted to demonstrate the snubbing action of a partially filled motor tube. This test also demonstrated that the first observable damage to a CRDM subject to a dry trip is as listed in II.B (above).

The tests were run at DPSC using an undamaged CRDM which had been returned from Oconee, disassembled, and inspected. The drive was then reassembled, put on the test autoclave, and a full " production acceptance test" run. The results of this production acceptance-test were nearly identical to those of the original production acceptance test which was run when the drive was first shipped to the Oconee Site. This demonstrated that the drive was a completely functional mechanism. Upon completion of the production acceptance testing, the drive was torn-down and again inspected to ensure the start of the gas volume test with an undamaged drive.

.The gas volume test consisted of:

1. Obtaining reference trip data with the CRDM and autoclave com-pletely filled with water by making traces of various stroke length trips at atmospheric pressure and ambient temperature.

2. Lowering the water level in one foot increments and making trip traces at each water level.

A-3

3. Each time the drive was tripped, the traces were examined and "g" loading computed. Once the terminal velocity of the drive line began to exceed normal limits, the drive was opened and inspected for damage after each trip.

4. This procedure was followed until the test was run to the

" threshold of damage" for the drive components.

5. This test was then repeated-for various pressures and temperatures to compare the results of various gas volumes (motor tube water level).

4 B. Test Results The results of the testing at DPSC are summarized as follows:

1. At 2000 psig and 550*F, the 10g limitation occurred at a water level of 10.0 feet from the top of the drive. No visible damage was incurred in this testing.

2. At 400 psig and ambient temperature, the 10g limitation occurred at a water level of 10.0 feet from the top of the drive. First indications of damage occurred at a water level of 11.0 feet from the top of the drive. This first indication of damage was in the form of a very slight brinelling of the support ring surface which is contacted by the torque tube cap tangs. -

3. At atmospheric pressure and ambient temperature, the 10g limitation occurred at a water Tavel of 4.0 feet from the top of the drive.

First indications of damage occurred at a water level of 10.0 feet from the top of the drive. This first visible damage was the same as described in (2) above.

C. Test Conclusions The test results outlined above have been reviewed and indicate the following conclusions:

1. With the reactor at rated *emperature and pressure with a 130 pound CRA and no RC pumps on the line (worst case), a full stroke trip will not result in a deceleration greater than 10g's provided the entrapped gas o'ubble is not greater than 10.0 feet from the top of the drive (12.0 feet from the top for start of damage).

2. With the reactor at 400 psig and at ambient temperature with a 130 pound CRA and no RC pumps on the line (worst case), a full stroke trip will not result-in a deceleration greater than 10g's provided the entrapped gas bubble is not greater than 10.0 feet from the top'of'the drive (11.0 feet from the top for start of damage).

A-4 e.

3. With the reactor at atmospheric pressure and at ambient temperature with a 130 pound CRA and no RC pumps on the line (worst case), a full stroke trip will not result in a deceleration of greater than 10g's provided the entrapped gas bubble is not greater than 4.0 feet from the top of the drive (10.0 feet from the top for start of damage).

4. The test program confirms the first visible indication of damage sustained to the drives, and proves the accuracy and validity of the acceptance criteria.

IV. Discussion and Conclusion The CRDM inspection data sheets, along with data from the Oconee Site, were utilized to put together a " core layout" diagram showing various types of information.

Measurements of gas volume vented from the drives were used to calculate the size of tue bubble in the vessel head. The diagram also shows the nozzle positions where drives were removed. This information, including gas volumes vented from each individual drive, were assessed to determine the distribution of the gas in the vessel head and drives when each drive was tripped. This evaluation was not conclusive due to the pressure perturbations that went on in the RC system, and the possible variation and re-distribution of the gas volume during these pressure changes. It can be readily seen from the core layout diagram that most of the drive damage occurred in the center of the vessel head, which would be expected if the gas bubble were as shown. This information tends to confirm the location of damaged drives and gives added assurance that drives which passed the " acceptance criteria" are satisfactory for operation.

Once the criteria was developed and verified, it was then used to inspect the remaining drives on the Unit 1 vessel head. In strict adherance to the critaria inspection requirements, six (6) additional drives were removed and shipped to DPSC because of some abnormal indications being found. The indications found during the inspr: tion were abnormal markings on the male coupling and/or the upper leadscrew extension in the area where the torque taker fits. These indications were not associated with the dry trip. All other drives passed the criteria inspection completely.

The tests and inspections discussed in the preceding Parts I through IV do provide assurance that the drivec left on the Unit 1 vessel head are completely functional for operation of the reactor.

A-5

1 SL*MMARY OF DRIVES 4 REMOVED FROM REACTOR VESSEL HEAD

1. prives That Were Dry Tripped Drive Nozzle Core Drive Nozzle Core S/N No. Location S/N No. Location 6 2 G-7 26* 5 K-7 7 8 L. 8 28 7 H-10 10 4 K-9 53* 3 G-9 11 1 H-8 61 9 H-6 17* 6 F 40 22 D-8

' 107 63 B-10 1

• Drives found to be improperly coupled upon disassembly.

f II. Drives (Shim Safety) With No Apparent Dry Trip Damage Driva S/N Nozzle No. Core Location 12 15 E-9 23 39 C-9 44 11 F-10 t 60 59 H-14 III. APSR Drives Damaged By Coupling-Uncoupling Drive S/N Nozzle No. Core Location X-64 30 D-6 X-66 37 F-4 X-68 35 N-6 X-70 36 L-4 IV. Drives (Shim Safety) With Abnormal Marks on Leadscrew Upper Extension

-(Not Attributed to Dry Trip)

Drive S/N Nozzle No. Core Location 2 29 M-5 25* 58 B-8 30 46 D-4 34*- 20 K-5 54*: 19 11-7 55 64 F-14

• Also had abnormal' marked male' couplings.

,. . l o

L A-6 l

V .

i-

. p e 4

VI. Condition of Drives (Shim Safety) Found To Be Damaged By Dry Trio Serial No.

Number with tangs sheared off. - 5 7,11,28,53,61 Number with tangs bent -4 6,10,17,76 Number with support ring brinelling only - 40,107 d

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A-10

DUKE I - CRG! INSPECTION g

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