ML18152A470

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Cycle 13 Control Rod Performance Test Results.
ML18152A470
Person / Time
Site: Surry  Dominion icon.png
Issue date: 06/13/1996
From:
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18152A471 List:
References
IEB-96-001, IEB-96-1, NUDOCS 9606180367
Download: ML18152A470 (19)
v • d • e


Text

ATTACHMENT VIRGINIA POWER SURRY UNIT 2 CYCLE 13 CONTROL ROD PERFORMANCE TEST RESULTS 9606180367 960613 PDR ADOCK 05000280 G__ _ ____ f:JifL _____ ~--

TABLE OF CONTENTS Section LIST OF TABLES .......................................................................................................................

3 LIST OF FIGURES .....................................................................................................................

3

1.0 INTRODUCTION

............................................................................................................

4 2.0 SURRY 2 CYCLE 13 RODDED FUEL ASSEMBLY OPERATION........................

4 2.1 Rodded Fuel Assembly Operation

..................................................................

4 2 .2 S2C13 Reactor Trips.........................................................................................

7 3.0 CONTROL ROD DROP TIME RES UL TS...................................................................

8 4.0 CONTROL ROD DRAG TEST RESULTS .................................................................

12

5.0 CONCLUSION

..............................................................................................................

18

6.0 REFERENCES

..............................................................................................................

19 Page 2 e LIST OF TABLES Table 2.1 -S2C13 Reactor Trip History.................................................................................

7 Table 3.1 -Surry 2 Cycle 13 Control Rod Drop Time Summary .......................................

8 Table-4.1

-Surry 2 Cycle 13 Control Rod Drag Test Results ...........................................

14 Table 4.2 -Surry Control Rod Drag Test Results to Support the Westinghouse Root Cause Evaluation

............................................................

15 LIST OF FIGURES Figure 2;-1 -Surry 2 Cycle 13 Comparison of BOC and EOG Rodded Fuel Assembly Burn up . .... .. . ... ... .. ... . .. .. .. .. . . . .. . . .. .. . . .. . . .... . . .. .. . . .. . .. . ... ... . . . . .. .... . . .. .. . . . .. .. .. . . . 6 Figure 3.1 -Surry 2 Cycle 13 Comparison of BOC and EOG Control Rod Drop Times to Dash pot.......................................................................................

9 Figure 3.2 -Plot of Control Rod Drop Time versus Fuel Assembly Burnup ...................

10 Figure 3.3 -Plot of Control Rod Drop Time versus Upper Guide Tube Drag Force .....................................................................................................................

11 Figure 4.1 -Upper Guide Tube Drag Data versus Assembly Burnup .............................

16 Figure 4.2 -Dash pot Drag Data versus Assembly Burnup ...............................................

17 Page3 --

e 1.,0 INTRODUCTION The N RC -issued Bulletin 96-01, "Control Rod Insertion Problems," in March 1996 (Reference 6.1 ). This Bulletin identified instances at Wolf Creek and South Texas Unit 1 where several control rods failed to fully insert following a reactor trip. The Bulletin also identified a situation at North Anna where two control rods in the spent fuel pool could not be removed from the fuel assemblies they were temporarily stored in with normal operation of the control rod handling tool (i.e., there was abnormally high drag force in these two assemblies which caused the control rod handling tool to trip on overload).

Bulletin 96-01 requested licensees with Westinghouse plants to perform several actions to ensure that the required shutdown margin is maintained following a reactor trip. Virginia Power's response to Bulletin 96-01 was sent to the NRC on April 8, 1996 (Reference 6.2). Section 4.2 of Virginia Power's response states that Virginia Power will perform control rod drop time tests at the end of all operating cycles during the 1996 calendar year. Section 4.2 also states that control rod drag forces will be -measured and evaluated for all -rodded -fuel assemblies in the -core at the -end of all operating cycles during 1996. Surry Unit 2 completed Cycle 13 on May 3, 1996. The end-of-cycle (EOC) control rod drop time tests were performed shortly after shutdown.

The control rod drag force measurements were completed on May 14, 1996. This report documents the results of the control rod drop time tests and the control rod drag force measurements performed at the end of Surry 2 Cycle 13 (S2C13). In addition to the requirements to measure and evaluate the control rod drop times and control rod drag forces in rodded fuel assemblies in the S2C13 core, Virginia Power participated in the root cause evaluation effort by allowing Westinghouse to examine ten additional discharged assemblies residing in the Surry spent fuel pool. The results of the drag force tests of the ten additional assemblies are reported here as well. 2.0 SURRY 2 CYCLE 13 RODDED FUEL ASSEMBLY OPERATION

2. 1 Rodded Fuel Assembly Operation Surry 2, Cycle 13, achieved initial criticality on March 19, 1995. The cycle ended operation on May 3, 1996. The cycle length was 13,043 MWD/MTU or 377 effective full power days. There was no coastdown at the end of the cycle. Surry Units 1 and 2 operate with 48 control rods (6 banks with 8 control rods per bank). All rodded fuel assemblies in the S2C13 core were the Surry Improved Fuel (SIF) design which is primarily the Westinghouse 15X15 OFA design. These fuel assemblies have inconel top and bottom grids with five Zircaloy-4 mid-grids.

The inside diameter of the upper guide tube for the SIF design is Page4 e e nominally 0.499 inches. The inside diameter of the guide tube dashpot is 0.455 inches. The dashpot region begins approximately 24 inches above the top plate of the fuel assembly bottom nozzle. With the exception of four of the eight fuel assemblies in C-bank, all rodded fuel assemblies in the S2C13 core were initially once-burned from Surry 2 regions 14A and 14B (initially used in Surry 2 Cycle 12). The four C-bank assemblies in the outer group were initially twice-burned from the Surry 1 Region 13B. The average .. rodded assembly burnup.at the beginning of S2C13 was 24,238 MWD/MTU. The minimum rodded assembly burn up at the beginning of S2C13 was 18,834 MWD/MTU, and the maximum was 34,695 MWD/MTU. The average rodded assembly burnup at the end of S2C13 was 37,960 MWD/MTU. The minimum rodded burnup at the end of S2C13 was 32,850 MWD/MTU, and the maximum was 43,633 MWD/MTU. The maximum burnup increase in any rodded fuel assembly in S2C13 was 16,591 MWD/MTU. Figure 2.1 shows the beginning-of-cycle (BOC) and end-of-cycle (EOG) average fuel assembly burnup for the rodded assemblies in Surry 2, Cycle 13. Page 5 Figure 2.1 Surry 2 Cycle 13 e Comparison of BOC and EOC Rodded Fuel Assembly Burnup {All fuel assembly burnup values are assembly average and represented in MWD/MTU) A B C D E F G H 6W1 3W7 2 25,115 22,909 33,820 35,246 6W5 3 22,485 38063 4H1 4W6 4 34,364 23,737 43,396 39,850 OW2 5 19,079 35,518 5W6 3W3 6W7 2W2 6 24,875 22,982 24,768 24,573 33,328 38,884 41,359 40,215 6WO 4W9 7 22,076 24,724 37,487 39,901 3W4 2W9 8 23,292 24,955 35,373 40,359 6W8 4W3 9 22,553 24,582 38,098 39,722 4WO 3W8 5W9 3W2 10 24,796 23,885 24,330 25,003 33,471 40,166 40,744 40,309 1WO 11 19,609 36,074 4H4 5W3 12 34,181 23,425 43,338 39,429 5W8 13 21,669 37,161 5WO 4W5 14 24,319 22,821 32,850 35,109 15 (Shaded location is maximum rodded assembly burnup) Page 6 J K 6W3 24,698 33,198 5W2 21,755 37,227 3W9 22,758 38,602 5W7 24,889 41,168 4W1 24,287 39,404 1W5 24,898 40,198 3W6 24,404 39,387 5W5 24,409 40,698 4W4 23,471 39,539 6W4 22,428 37,927 4W8 24,685 33,179 L M \llijlll!ll!

til.@§(;}:@f OW6 19,139 35,467 3W5 23,182 39,065 5W4 23,340 39,317 OW5 18,834 35,335 3H9 34,248 43,323 N 5W1 22,368 37,780 6W2 21,494 36,862 p R 4W7 24,447 32,940 4W2 23,034 35,289 6W6 24,838 33,262 Assembly ID BOCBumup EOC Burnup h e 2. 2 S2C13 Reactor Trips -There were six reactor trips during the operation of S2C13. With the exception of the control rod in core location M 10, there were no occurrences of anomalous control rod insertion behavior noted during any of the reactor trips. The rod position indication system for location M 10 was faulty giving erroneous indication of rod position after the control rod was inserted into the core. Subsequent rod drop tests proved that the control rod in location M10 functioned properly and the problem existed with the rod position indication system. There were no operational problems associated with the rod with the position indicator for core location M10 subsequent to the issuance of NRG Bulletin 96-01, and maintenance was performed on the system components and control rod drive mechanism for M10 during the refueling outage following Cycle 13. The reactor trips for S2C13 are summarized in the following table: Table 2.1 S2C13 Reactor Trip History Date Description Reference May 11, 1995 Manual reactor trip due to control bank B, group 2 control 6.3 rods dropping into the core. May 21, 1995 Manual reactor trip due to control bank A, group 2 control 6.3 rods dropping into the core. June 14, 1995 Automatic reactor trip due to main transformer generator 6.4 differential lockout. November 7, Automatic reactor trip on loss of 2-RC-P-1 C. 6.5 1995 February 23, 1996 Reactor manually tripped from low power for maintenance 6.6 outage. May 3, 1996 Reactor manually tripped from low power for refueling none outage. Page?

7 e 3 .. 0 CONTROL ROD DROP TIME RESULTS Control rod assembly drop times were measured shortly after Surry 2 Cycle 13 shut down for refueling.

The tests were performed at hot, full flow conditions.

All control rod drop times were within the limit of 2.4 seconds as specified in Section 3.12.C.2 of the Surry Technical Specifications (Reference 6.7). Full insertion of the control rods into the fuel assemblies was determined by reviewing the timing test strip charts for evidence of recoil. Recoil occurs when the spring pack on the control rod hub compresses when the control rod impacts the fuel assembly upper end fitting. When the spring pack decompresses, the control rod lifts off the fuel assembly upper end fitting. This bouncing motion is referred to as recoil. Control rod recoil was observed for all 48 control rods during the test. The following table compares the control rod drop times that were measured at the beginning of Cycle 13 to the drop times that were measured at the end of Cycle 13. The average increase in control rod drop time is 0.04 seconds. The greatest drop time increase in single rodded location was 0.1 seconds. Table 3.1 Surry 2 Cycle 13 Control Rod Drop Time Summary BOC Average Drop Time, seconds 1.28 Minimum Drop Time, seconds 1.22 Maximum Drop Time, seconds 1.40 Average Rodded Assembly Burnup 24,238 MWD/MTU Minimum Rodded Assembly Burnup 18,834 MWD/MTU Maximum Rodded Assembly Burnup 34,695 MWD/MTU EOC 1.32 1.25 1.45 37,892 MWD/MTU 32,807 MWD/MTU 43,587 MWD/MTU The core map on the following page (Figure 3.1) identifies the individual control rod bank position with the corresponding BOC and EOG control rod drop times. Figure 3.2 is a plot of control rod drop time versus fuel assembly burnup and Figure 3.3 is a plot of control rod drop time versus drag force in the assembly's upper guide tube. The upper guide tube is the guide tube region that is above the guide tube's dashpot region. The dashpot region extends approximately 24 inches above the fuel assembly's bottom nozzle. Page 8 A B 2 3 4 5 A 6 1.32 1.35 7 D 8 1.31 1.33 9 A 10 1.31 1.34 11 12 13 14 15 e Figure 3.1 Surry 2 Cycle 13 e Comparison of BOC and EOC Control Rod Drop Times to Dash pot (All drop times are represented in seconds) C D E F G H J K L M A D A 1.27 1.26 1.27 1.35 1.28 1.30 SA SA 1.30 1.26 1.32 1.27 C B B C 1.28 1.27 1.25 1.28 1.31 1.30 1.31 1.29 SB SB 1.28 1.22 1.33 1.27 B D C D B 1.27 1.40 1.31 1.31 1.26 1.37 1.45 1.34 1.36 1.31 SA SB SB 1.26 1.30 1.28 1.28 1.36 1.35 C C 1.30 1.28 1.33 1.30 SA SB SB 1.27 1.28 1.30 1.28 1.33 1.36 B D C D B 1.27 1.32 1.30 1.31 1.27 1.33 1.36 1.35 1.34 1.31 SB SB 1.27 1.26 1.34 1.29 C B B C 1.27 1.28 1.25 1.27 1.31 1.31 1.30 1.30 SA SA 1.28 1.28 1.30 1.27 A D A 1.26 1.26 1.31 1.32 1.29 1.31 Page 9 N SA 1.27 1.25 SA 1.27 1.28 p R A 1.28 1.30 D 1.23 1.26 A 1.27 1.28 Bank ID BOC Time EOC Time 1.50 1.45 1.40 u Cl) 1.35 Ill ..; 0 C. .c Ill cu C 0 -Cl) 1.30 E j:: C. 0 ... C c( (.) 1.25 (.) 0:: 1.20 1.15 1.10 0 -e FIGURE3.2 RCCA DROP TIME VERSUS FUEL ASSEMBLY BURNUP 5000 Surry RCCA Drop Time vs. Assembly Burnup ---. * .. -----. * * '"' ** i. . * .. *** * .. * -----* * *~ * * ***** *** ... * .... u ** ** * *** * ----*

  • 10000 15000 20000 25000 30000 35000 40000 45000 Assembly Burnup, MWD/MTU Page 10

-Figure 3.3 e Control Rod Drop Time versus Upper Guide Tube Drag Force RCCA Drop Time versus Upper Guide Tube Drag Force 1.50 . -1.45 1.40 -* -~ * * * -. . .----* * * * * *** aA ** * * *** * * * . .. .. -. -----1.30 * * * * * * * * *~ * . -1.25 1.20 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 Upper Guide Tube Drag Fore, lbf Page 11

  • e 4 .. 0 CONTROL ROD DRAG TEST RES UL TS Westinghouse personnel measured the drag force of control rods in the rodded fuel assemblies from Surry 2, Cycle 13 (S2C13). The tests were performed in the spent fuel pool following the offload of the S2C13 core using the actual rodded fuel assemblies from S2C13 with their respective control rods. The acceptance criteria for the drag tests were adopted from the Westinghouse Fuel Handling Specification F-5.1 (F-Spec 5.1 ). F-Spec 5.1 indicates that the maximum drag load of a control rod in a fuel assembly should not exceed 100 pounds in the dashpot region or 40 pounds in the upper guide tube region (region above the dashpot).

All 48 rodded fuel assemblies from S2C13 had less than 100 pounds of control rod drag force in the dashpot region. Two fuel assemblies exceeded the F-. Spec limit for control rod drag force in the upper guide tube region. An upper guide tube drag force of 42.5 pounds was measured in fuel assembly 3W2 (40,309 MWD/MTU).

An upper guide tube drag force of 46.5 pounds was measured in fuel assembly 3W3 (38,884 MWD/MTU).

While there appears to be a gradual increasing trend of control rod drop time* with increasing upper guide tube drag force (see Figure 3.3), upper guide tube drag forces extrapolated to beyond 50 pounds would still result in control rod drop times well within the Technical Specifications limit. The control rod drop times for these two assemblies were 1.35 seconds and 1.37 seconds, respectively, which is similar to drop times measured in assemblies with upper guide tube drag forces as low as 15 pounds. Recoil was observed for all control rods during the drop time tests performed at the end of the cycle. Therefore, the control rods in assemblies 3W2 and 3W3 both fully inserted and were within the Technical Specifications limit for drop time. A Fuel Anomaly Form was issued in accordance with Nuclear Analysis and Fuel Implementing Procedure NAF-202 documenting that the drag force exceeded the acceptance criteria.

These two fuel assemblies are not scheduled for reuse. However, these two assemblies will be restricted from use in a core location that requires an insert component should they be scheduled for use in the core in the future. Ten additional assemblies were tested to support Westinghouse's root cause evaluation of the incomplete control rod insertion problem. These ten assemblies were selected based on their burnup (ranging from 45.6 GWD/MTU to 56.1 GWD/MTU), date of fabrication, and design. Five of these ten assemblies exceeded either the upper guide tube drag force criterion of 40 pounds or both criteria for dashpot drag force (100 pounds) and upper guide tube drag force. Assemblies 5H1, 3H8, 4G1, and OV2 exceeded both criteria.

Assembly OV7 exceeded only the upper guide tube drag force criterion.

Both 5H1 and 3H8 were rodded assemblies in their last cycle of operation (Surry 1, Cycle 13, core locations FOB and H10, respectively).

Both OV2 and OV7 were rodded assemblies in their last cycle of operation (Surry 2, Cycle 12, core locations FOB and GO?, respectively).

Assembly 4G1 was not rodded Page 12

.. e e . during its last cycle of operation which was Surry 1 Cycle 12. The Surry dummy control rod (used to measure the drag forces in the ten additional Surry assemblies to support the Westinghouse root cause investigation) fully inserted into these five fuel assemblies prior to performing the drag tests. The fact that the dummy control rod fully inserted into each of these assemblies in the spent fuel pool, in spite of the drag forces exceeding one or both of the acceptance criteria, is evidence that the control rods would have fully inserted into these four assemblies (5H1, 3H8, 4G1, and OV2) during cycle operation.

Fuel Anomaly S296-7 documents that the drag forces in these five assemblies exceeded the acceptance criteria.

None of these five fuel assemblies are scheduled for reuse. However, these assemblies will be restricted from use in a core location that requires an insert component should they be scheduled for use in the core in the future. Table 4.1 lists the control rod drag test results for the rodded assemblies from Surry 2 Cycle 13. Table 4.2 lists the control rod drag test results for the ten assemblies that will be used to support the Westinghouse root cause evaluation.

Figures 4.1 and 4.2 are plots of the upper guide tube drag force versus assembly burnup and dashpot drag force versus assembly burnup (respectively) measured at Surry. In general, the control rod drag forces measured in the rodded fuel assemblies from S2C13 are consistent with control rod drag force data measured at other plants. The drag force data obtained from the ten additional Surry discharged assemblies tend to be higher but are consistent with the trend of increasing drag force with increasing assembly burnup. Page 13

-Table4.1 e Surry 2 Cycle 13 RCCA Drag Test Results Assembly Burnup Core Dashpot Drag Guide Tube Drag Drop Time Fuel Assembly ID (MWD/MTU)

Location Force (lbf) Force (lbf) (sec) Design 3H6 43633 M04 31.0 12.5 1.29 SIF 4H1 43396 D04 27.5 6.0 1.31 SIF 4H4 43338 D12 26.0 7.5 1.31 SIF 3H9 43323 M12 20.0 3.0 1.30 SIF 6W7 41359 F06 42.5 27.5 1.45 SIF 5W7 41168 K06 51.5 36.0 1.36 SIF 5W9 40744 F10 47.5 15.0 1.36 SIF 5W5 40698 K10 31.0 6.0 1.34 SIF 2W9 40359 FOB 58.5 31.0 1.33 SIF 3W2 40309 H10 60.0 42.5 1.35 SIF 2W2 40215 H06 50.0 26.0 1.34 SIF 1W5 40198 K08 40.0 16.0 1.30 SIF 3W8 40166 D10 45.0 26.5 1.33 SIF 4W9 39901 G07 38.5 22.5 1.36 SIF 4W6 39850 F04 44.0 24.0 1.30 SIF 4W3 39722 G09 50.0 25.0 1.33 SIF 4W4 39539 K12 26.0 2.5 1.30 SIF 5W3 39429 F12 41.5 28.5 1.31 SIF 4W1 39404 J07 47.5 24.0 1.35 SIF 3W6 39387 J09 50.0 26.0 1.36 SIF 5W4 39317 M10 30.0 5.0 1.31 SIF 3W5 39065 M06 32.5 14.0 1.31 SIF 3W3 38884 D06 60.0 46.5 1.37 SIF 3W9 38602 K04 27.5 7.5 1.31 SIF 6W8 38098 C09 35.0 16.0 1.28 SIF 6W5 38063 G03 32.5 12.5 1.32 SIF 6W4 37927 J13 27.5 4.0 1.27 SIF 5W1 37780 N07 25.0 2.5 1.25 SIF 6WO 37487 C07 30.0 9.0 1.28 SIF 5W2 37227 J03 21.0 2.5 1.27 SIF 5W8 37161 G13 31.0 8.5 1.30 SIF 6W2 36862 N09 25.0 4.0 1.28 SIF 1WO 36074 E11 53.5 34.0 1.34 SIF OW2 35518 E05 49.0 30.0 1.33 SIF OW6 35467 L05 32.5 10.0 1.27 SIF 3W4 35373 B08 25.0 4.0 1.33 SIF OW5 35335 L11 30.0 4.0 1.29 SIF 4W2 35289 P08 25.0 7.5 1.26 SIF 3W7 35246 H02 26.0 4.0 1.28 SIF 4W5 35109 H14 25.0 6.0 1.29 SIF 6W1 33820 F02 40.0 22.5 1.35 SIF 4WO 33471 B10 31.5 11.5 1.34 SIF 5W6 33328 B06 36.0 20.0 1.35 SIF 6W6 33262 P10 25.0 4.0 1.28 SIF 6W3 33198 K02 24.0 1.5 1.30 SIF 4W8 33179 K14 20.0 1.0 1.31 SIF 4W7 32940 P06 28.5 9.0 1.30 SIF 5WO 32850 F14 25.0 10.0 1.32 SIF Page 14 7 r Assembly ID 5H1 3H8 1GO 4G1 OF6 OF3 3J1 OV7 OV2 OJ9 OUM e e Table 4.2 Surry RCCA Drag Test Results to Support the Westinghouse Root Cause Evaluation Burnup Dashpot Drag Guide Tube Drag (MWD/MTU)

Force (lbf) Force (lbf) 56138 101.3 70.0 55866 112.5 77.5 53967 67.5 33.5 53885 130.0 53.5 49364 40.0 15.0 49246 25.0 7.5 46189 42.5 17.5 46160 98.5 49.0 46146 100.5 55.0 45643 61.5 25.0 0 15.0 1.0 Page 15 Fuel Assembly Design SIF SIF SIF SIF STD STD SIF SIF SIF SIF STD r ---100.0 90.0 80.0 70.0 .... .0 G) 60.0 u .. 0 LL. en ns .. C C1) 50.0 .0 :::s I-C1) :2 :::s (!) 40.0 .. C1) C. C. ::::, 30.0 20.0 10.0 0.0 " 0 Figure 4.1 Upper Guide Tube Drag Data versus Assembly Burnup o S2C13 Rodded Assemblies for NRCB 96-01 Compliance x Additional Surry Data to Support Westinghouse Root Cause Evaluation 0 0 -0 0 --0 0 0 0 -tl @o 0 0 a ODO ::J 0 tl ~o ' fb 0 El CD 0 10000 20000 30000 40000 Assembly Burnup, MWD/MTU Page 16 X l(-X X X X X X X 0 0 X 0 0 50000 60000 150.0 140.0 130.0 120.0 110.0 100.0 ... 90.0 :9 G) 0 80.0 u.. C) I! C 70.0 -0 Q. .c UI cu 60.0 C 50.0 40.0 30.0 20.0 10.0 0.0 . ( Figure 4.2 Dashpot Drag Data versus Assembly Burnup D S2C13 Rodded Assemblies for NRCB 96-01 Compliance x Additional Data to Support Westinghouse Root Cause Evaluation

...,..a D __ D D --DD ~jl D -tr D D D £I n.D~ D D .... --CD DD *c -D -* X 'II! X X -X X X

  • D 8 X -0 5000 10000 15000 20000 25000 30000 35000 40000 45000 50000 55000 60000 Assembly Burnup, MWD/MTU Page 17
  • e

5.0 CONCLUSION

Based on the evaluation of the data presented in this report, the control rods in Surry Unit 2 Cycle 13 operated as designed without any incident of incomplete control rod insertion.

The data in this report support the conclusion presented in Attachment 2 of Virginia Power's response to NRC Bulletin 96-01 which states that the control rods in Surry Units 1 and 2 remain operable in light of current information and that operation of S1C14 and S2C14 to their respective design burn up limits is acceptable.

This finding is based on the following:

  • Except for problems with the rod position indication system associated with the control rod in core location M10, there were no incomplete control rod insertion problems similar to the events described in NRG Bulletin 96-01 following the six reactor trips during the operation of Surry 2 Cycle 13. There were no operational problems associated with the rod with the position indicator for core location M 10 subsequent to the issuance of-N RC Bulletin 96-01, and maintenance was -performed on the system components and control rod drive mechanism for M10 during the refueling outage following Cycle 13.
  • As determined by control rod drop time tests performed at the end of Surry 2 Cycle 13, all 48 control rods in the Surry 2 Cycle 13 core were within the Technical Specifications limit for drop time to the dashpot.
  • As determined by control rod drop time tests performed at the end of Surry 2 Cycle 13, control rod recoil was observed for each of the 48 control rods in the core. The presence of recoil verifies that the control rod fully inserted when tripped into the core.
  • Drag tests of the Surry 2 Cycle 13 rodded assemblies were performed in the spent fuel pool following the offload of the Cycle 13 core. Two assemblies (assemblies 3W2 and 3W3) exceeded the drag force acceptance criterion of 40 pounds in the upper guide tube by 2.5 pounds and 6.5 pounds, respectively.

As expected, there appears to be a gradual increasing trend of control rod drop time with increasing upper guide tube drag force. However, when the upper guide tube drag force data are extrapolated to 50 pounds (and beyond), the resulting control rod drop times are still well within the Technical Specifications limit. These two fuel assemblies are not scheduled for reuse. However, these two assemblies will be restricted from use in a core location that requires an insert component should they be scheduled for use in the core in the future.

  • All drag test results of the Surry 2 Cycle 13 rodded assemblies with the exception of 3W2 and 3W3 cited above were within the acceptance criteria of 40 pounds in the upper guide tube and 100 pounds in the dash pot. Page 18
  • Five of the ten additional discharged Surry fuel assemblies that were tested to support the Westinghouse root cause evaluation of the incomplete control rod insertion problem exceeded either the upper guide tube drag force criterion of 40 pounds or both criteria of 40 pounds in the upper guide tube and 100 pounds in the dashpot. Four of these five assemblies were rodded assemblies during their last cycle of operation.

However, the fact that the control rod used to test assembly drag forces fully inserted into the test assemblies in the spent fuel pool prior to performing the drag test is evidence that a control rod would also have fully inserted into the assemblies during cycle operation.-

These five fuel assemblies are not scheduled for reuse. However, these assemblies will be restricted from use in a core location that requires an insert component should they be scheduled for use in the core in the future.

6.0 REFERENCES

6.1 NRC Bulletin 96-01, "Control Rod Insertion Problems," March 8, 1996. 6.2 Letter from Mr. James P. O'Hanlon (Virginia Power) to the NRC, Serial No.96-135, entitled "Virginia Electric and Power Company, North Anna and Surry Power Stations Units 1 and 2, NRC Bulletin 96-01 Control Rod Insertion Problems," Dated April 8, 1996. 6.3 Letter from Mr. M. L. Bowling (Virginia Power) to the NRC, Serial No.95-289, entitled "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Monthly Operating Report," Dated June 12, 1995. 6.4 Letter from Mr. M. L. Bowling (Virginia Power) to the NRC, Serial No.95-353, entitled "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Monthly Operating Report," Dated July 13, 1995. 6.5 Letter from Mr. R. F. Saunders (Virginia Power) to the NRC, Serial No.95-628, entitled "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Monthly Operating Report," Dated December 12, 1995. 6.6 Letter from Mr. M. L. Bowling (Virginia Power) to the NRC, Serial No.96-125, entitled "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Monthly Operating Report," Dated March 13, 1996. 6.7 Surry Technical Specifications Section 3.12, "Control Rod Assemblies and Power Distribution Limits." Page 19

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