ML18150A293
| ML18150A293 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/13/1986 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Rubenstein L Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18150A294 | List: |
| References | |
| REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR 86-239, TAC-60309, TAC-60310, NUDOCS 8605210335 | |
| Download: ML18150A293 (9) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261
- w. L. STEWART VICE PRESIDENT NUCLEAR OPERATIONS May 13, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Lester S. Rubenstein, Director PWR Project Directorate No. 2 Division of PWR Licensing-A U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RE:
PROPOSED TECHNICAL SPECIFICATION CHANGE TRANSFER CANAL DOOR - HEAVY LOAD se*rial No.
( NO/JBL: acm Docket Nos.
License Nos.86-239 50-280 50-281 DPR-32 DPR-37 In our letter of December 11, 1985, Seri~l No.85-571, we proposed a change to the Surry Technical Specifications to allow movement* of the transfer canal doors over spent fuel if necessary.
In subsequent telephone conversations with the NRC Project Manager and by your letter dated April 10, 1986, you have requested supplemental information to support this Technical Specification change. to this letter provides the information you requested.
Should you have any further questions, please contact us.
~
W. L. Stewart Attachments rl 0605210335 060513 PDR ADOCK 05000280 I p PDR 1
cc:
Dr. J. Nelson Grace Regional Administrator NRC Region II Mr. Chandu P. Patel NRC Surry Project Manager PWR Project Directorate No. 2 Division of PWR Licensing-A NRC Senior.Resident Inspector Surry Power Station Mr. Charles Price Department of Health 109 Governor Street Richmond, Virginia 23219
ATTACHMENT 1 SURRY POWER STATION UNIT NOS. 1 AND 2 RESPONSE TO THE NRC'S REQUEST FOR ADDITIONAL INFORMATION RE:
PROPOSED TECHNICAL SPECIFICATION CHANGE TRANSFER CANAL DOOR - HEAVY LOAD
L
- 1.
NRG REQUEST:
Adequate justification has not been provided for the need to move a transfer canal door over the spent fuel.
Provide this justification, including the plant conditions that would necessitate movement of the.
transfer canal door over the spent fuel, and the intended travel path to be followed.
(SRP 9.1.5 Parts I, II, and III).
RESPONSE
The overhead crane (1-FH-CR-15) which is located on the east end of the spent fuel pool is the only crane in the area with sufficient lifting height to remove the transfer canal door from the spent fuel pool. In order to perform maintenance on either of the transfer canal doors, the affected door must first be moved by the spent fuel pool bridge crane (1-FH-CR-13), traversing the pool to the east end of the pool giving the overhead crane (1-FH-CR-15) access to the door.
This change to the Technical Specifications has been requested because maintenance could be required on a transfer canal door and because the spent fuel pool is near capacity.
This Technical Specification change will allow us to utilize the most direct travel path, thereby eliminating unnecessary shuffling of spent fuel in the spent fuel racks.
- 2.
NRC REQUEST:
In accordance with SRP 9.1.5 Part III.3.c, describe the operating and test procedures, preventive maintenance checks, and inspections on the overhead crane and its systems that will be performed prior to and during the movement of the transfer canal door over the spent fuel.
RESPONSE
Preventive maintenance is performed annually on the fuel handling cranes (1-FH-CR-13 and 15).
The preventive maintenance includes the following:
a)
Clean, inspect, and test the low voltage motors.
b)
Both visual and magnetic particle examination of the crane hooks.
c)
Inspect the wire rope for broken, worn, crushed or corroded wire and inspect the wire end connections.
d)
Overall equipment inspection which includes:
- 1)
Bearings, shafts, gears, and rollers.
- 2)
Brake system.
- 3)
Structural members, bolts and rivets.
e
- 4)
Wire rope, hooks and retainers.
- 5)
Safety devices, including limit switches.
- 6)
Electrical system.
These crane inspections are performed by a qualified contractor.
The inspections and preventive maintenance were last completed on the fuel handling (1-FH-CR-13) and overhead crane (1-FH-CR-15) on December 5, 1985.
Operator qualification is required to operate either crane.
In addition, a procedure is in place which controls the movement of the transfer canal door.
- 3.
NRG REQUEST:
List and describe all interlocks and safety devices that are used to maintain the system in a safe condition when traveling over the spent fue 1.
( SRP 9. 1. 5 Part I II. 3. b)
RESPONSE
As indicated in our "Nine Month Report" on NUREG-0612, Control of Heavy Loads, for Surry Power Station - Units 1 & 2, Response for Section 2.2.4, the motor driven platform and hoists (1-FH-13) are used in the movement of the transfer canal doors.
The motor driven platform and hoist is equipped with several interlocks and safety features to ensure safe operation.
These are identified and briefly described as follows:
a)
Hoist Gear-Unit Stops*
This is an interlock provided in the logic of the controls which stops the upward and downward movement at a
predetermined location.
In addition, there are mechanical backup stops provided to stop the hoist in case of any possible failure of the above interlock.
These are provided for both hoists.
b)
Deadman Pushbutton Switch Controls The motor driven platform and hoists require an operator to depress and hold the button in to continue operation.
Upon release of the button, all movement will cease. -To begin movement again, the button must be, depressed and held in.
c)
Emergency Stop Switch The pendant control station for the hoists is provided with an emergency stop switch which when activated kicks out the breaker.
This shuts off all power to the hoist and will stop all movement.
d)
Mechanical Rail Stops There are mechanical stops provided on all rails at both ends which prevent the platform from derailing.
Procedures are in place which provide for preventive maintenance and periodic inspection of these systems.
NDT procedures exist for testing of the hooks as well.
In addition, to operate the motor driven platform and hoists, an operator must have had appropriate training.
- 4.
NRC REQUEST:
The evaluation in the December 11, 1985 letter states that there will be damage to the spent fuel pool, but does not address the potential for leakage.
Discuss the extent of damage that the spent fuel pool may incur from a dropped transfer canal door, and justify that sufficient borated make-up water is available to maintain the spent fuel pool water level.
RESPONSE
Load case C, the inclined drop over a leak test channel, is the most limiting with respect to damage to the spent fuel pool liner.
Case C is postulated to occur over a leak test channel on the pool floor, as shown in Figure 5-2 of NES Report No. 83A1040.
It was found that the liner plate will deform a maximum of O.132 inches.
At this deflection, the door will contact the surrounding concrete and the remainder of the impact energy will be transferred directly to the concrete.
The liner itself will yield along the edge of the leak test channel, and will rotate a maximum of 12 degrees at the yield plane.
However, because of the high ductility of the stainless steel, no fracturing occurs.
Therefore, no leakage occurs, and borated makeup water is not required to maintain spent fuel pool water level.
- 5.
NRG REQUEST:
The evaluation in the December 11, 1985 letter states that there would be a radioactivity release due to damage to a control rod assembly, but that it would be less than the analysis performed in the UFSAR
- 14. 4. 1. 3, "Fuel Handling Accident in the Spent Fuel Pool." NUREG-0612 in Section 5.1 provides evaluation criteria for use in assessing alternative approaches for the control of heavy loads; Item I states the following:
"Releases of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce doses that are well within 10 CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than 1/ 4 of Part 100 limits);"
Show that the radioactive release referred to in the letter is within this NUREG-0612 guideline.
(SRP 9.1.5 Part II)
RESPONSE
Dose calculations for the release of radioactive material involving accidental dropping of a postulated heavy load were not performed because the drop analysis showed that no spent fuel would be damaged.
The door drop may damage a control rod assembly (not spent fuel) as a result of the deformation of 2.42 inches at the top of the spent fuel rack cell (load case D).
A more detailed discussion of this analysis can be found in response to NRC Request 6.
While a dose calculation for damage to a control rod was not performed, it can be concluded that a radioactivity release due to damage to a control rod assembly is considerably less than the release due to a damaged spent fuel assembly.
Surry UFSAR 14. 4.1. 3, Fuel Handling Accident in the Spent Fuel Pool, indicates that a damaged spent fuel assembly in the spent fuel pool would not result in an excessive radiation exposure at the site boundary.
The whole body and thyroid doses are below the guidelines of 10 CFR 100 with the conservative assumption that all 204 fuel rods in an assembly fail.
The potential releases from failure of 204 fuel rods is much greater than the potential radioactivity releases from a damaged control rod assembly since there is no fissionable material in a control rod.
Therefore, releases from a damaged control rod assembly would be well within the NUREG-0612 guideline of 25% of the 10 CFR 100 limits.
- 6.
NRC REQUEST:
The evaluation in the December 11, 1985 letter states that the analysis considered a vertical drop on a spent fuel rack that results in 2.42 inches of permanent axial deformation, and that the impacted cell would dislodge from the individual cell support legs.
The analysis does not consider the effects of a dropped transfer canal door other than on the direct axial centerline of a fuel rack.
Show that the vertical drop along the fuel rack axial centerline is the worst case condition with regard to fuel rack damage, radioactivity release, and maintaining subcriticality within the spent fuel pool (Keff = 0.95)
RESPONSE
The evaluation, in fact, considered the vertical drop on a single cell, regardless of its location.
Because of the geometry and construction features of the rack, each cell acts independently of the others for the vertical impact. Further, it was assumed that only a third of the cross-sectional area of the cell will be effective in resisting the impact from the door. This is conservative because the minimum dimension of the door is 7 inches and during impact, the flares at the top of the cell will fold, providing a larger contact area.
The evaluation predicted a local crushing of 2.42 inches at the top of the cell, not axially distributed along the cell.
This case is the most conservative scenario, since the impact forces are distributed over a very small area of a cell. Any other scenario, e.g. tipping of the door after initial impact, would result in contact forces which would be distributed over a large number of cells.
Even under the conservative assumption of the analysis, the overall rack integrity is maintained, and the cells remain within the top and bottom grids.
The dislodgement of the impacted cell at its base results only in vertical motion of the cell. The centerline distance between cells remains the same in the active fuel area.
It is therefore concluded that the overall rack integrity is maintained, and subcriticality is not compromised.
- 7.
NRG REQUEST:
Please provide a copy of the report entitled, "The Transfer Canal Door Drop Analysis for Surry Units 1 and 2, Revision 2 11 referenced in the safety evaluation.
RESPONSE
A copy of the report is provided as Attachment 2 to this letter.
ATTACHMENT 2 SURRY POWER STATION UNIT NOS. 1 AND 2 NUCLEAR ENERGY SERVICES, INC.
TRANSFER CANAL DOOR DROP ANALYSIS REVISION 2