ML13330B395
| ML13330B395 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/23/1988 |
| From: | Medford M SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8809300039 | |
| Download: ML13330B395 (35) | |
Text
Southem California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 M. 0. MEDFORD TELEPHONE MANAGER OF NUCLEAR ENGINEERING (818) 302-1749 AND LICENSING September 23, 1988 U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Spent Fuel Transshipment San Onofre Nuclear Generating Station Unit 1 In order to alleviate the accumulation of spent fuel in the spent fuel pool at San Onofre Unit 1, SCE has proposed a method for shipment of the spent fuel to San Onofre Units 2 and 3. On August 12, 1988, representatives from SCE met with members of the NRC staff to discuss the NRC reviewer's questions regarding the transshipment of spent fuel.
These questions were provided to SCE by letter dated August 11, 1988. During the meeting, verbal responses were provided to the NRC questions. Detailed responses to the questions are provided in Enclosure 1. In addition, during the meeting, five open items remained which required additional information that is provided herein. Those items are resolved as follows.
- 1. The NRC reviewer requested that the loads on the walls and the basemat should be provided for the postulated drops of the spent fuel cask. This information is provided as Enclosure 2.
- 2. It was requested that secondary missiles resulting from the spent fuel cask falling into walls should be addressed. This should include the possible damage to equipment and spent fuel.
With regards to the affect on spent fuel, criticality should be addressed. The response to this concern is provided in the response to Question 5 in Enclosure 1.
- 3. The NRC requested SCE's plans for repair of the spent fuel pool liner.
SCE's plans for repair of the spent fuel pool liner leakage are provided in the response to Question 9 in Enclosure 1.
8809300039 8SO923 PDR ADOCK 05000206 P
Document Control Desk
-2
- 4. The NRC reviewers also requested that SCE provide the plans for the 5 year load testing of the spent fuel cask lift rig. The spent fuel cask and its appurtenances are the property of Pacific Nuclear Systems and are under lease to SCE for the transshipment of spent fuel during the Cycle X refueling outage. Certification of the cask and its appurtenances including the lift rig is provided to SCE for verification of use of the equipment to ship spent fuel.
As part of this verification, it is determined that the lift rig is certified to ANSI N14.6. Correspondence on the load test and inspection of the lift rig is provided as.
- 5. The NRC reviewer requested that SCE provide the status of the procedure and training of individuals that will be used to perform the fuel shipment. The procedure under which the fuel movement will be done, S0123-X-9, Transshipment of Spent Fuel Using the IF-300 Cask, is completed, approved and in place for use. SCE personnel involved in the transshipment have received classroom training, and performed the evolutions under the instruction of the cask vendor representative during a dry run at Units 2 and 3. Key aspects, such as head removal and installation, were repeated for maximum training of personnel.
Due to the length of time that has elapsed since this training, the classroom training and the dry run will be repeated prior to the start of the transshipment program.
At the conclusion of the meeting, it was indicated that the NRC staff would proceed with the review and development of the safety evaluation. If there are any additional concerns regarding this issue or this information, please let me know.
Very truly yours, Enclosures cc: 3. B. Martin, Regional Administrator, NRC Region V F. R. Huey, NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3 Question 1 The plan view (Figure 1) needs to be supplemented by one or more vertical sections through the Decontamination Pad, Cask Laydown Area, Upender Area, Spent Fuel Pool and New Fuel Area, and the Turbine Deck Areas.
Response
Drawing numbers 568133, 568134, 568138, 568140, 568141, 568144 and 568148 were provided to Mr. Rinaldi of the NRC during the August 12, 1988 meeting. SCE's June 10, 1988 submittal indicates there is a waterproof membrane surrounding the basemat of the spent fuel pool and extends up to grade elevation. The drawings show the membrane only extends to Elevation 12'.
This will not affect the conclusions in the June 10, 1988 submittal since the leak chase system is pumped down and the water is always maintained below Elevation 12' and the postulated cask drop into the pool will not amplify the leakage because a plate is provided to prevent liner perforation.
Question 2 You identify a Category B seismic event and a 0.67g seismic event. Define these events.
Response
At San Onofre Unit 1, Seismic Category A refers to the maximum free-field horizontal ground motion acceleration for the Design Basis Earthquake (DBE) of 0.67g, Housner response spectra, increased 10% in the period range 0.07 second to 0.25 second and for a maximum free-field vertical ground motion acceleration for the DBE of 2/3 the horizontal Housner response spectra increased 10% in the period range from 0.05 second to 0.15 second. Seismic Category B is the equivalent static loading of 0.20g horizontal and 0.13g vertical applied simultaneously or the requirements of the current edition of the Uniform Building Code at the time of final design, whichever is greater.
Seismic Category B does not refer to the OBE. An OBE corresponding to the Seismic Category A criteria does not exist for San Onofre Unit 1.
Question 3 Discuss the basis for the selection of the 10'-6" and 2'-8" heights for the postulated cask drop without the impact limiter. The 2'-8" is close to the 2'-3" cask lift height identified in your response to Question 6. Please clarify your assumptions. Also, expand your explanation of the values of 0.09" and 1.92" on page 3.
-2
Response
Postulated cask drops which were evaluated included drops from the height of 10'-6" on the portions of the north turbine deck extension and the decontamination pad and 2'-8" high drop on the decontamination pad. The selection of the drop heights was based on:
- a. The top of the horizontal beam of the gantry crane leg is 10' above the elevation of the north extension deck and the decontamination pad. An additional 6 inches is assumed to clear the cask platform when lifting the cask off the crane leg.
Thus, the total height of the lifted cask above the north extension deck and the decontamination pad is 10'-6".
During this lift, the impact limiter is used.
- b. The lower portion of the impact limiter is 2-foot high. The cask will be lowered above the impact limiter to maintain a 3" clearance. At that time the impact limiter will be removed. The analysis was performed to determine the maximum acceptable drop on the decontamination pad without the impact limiter. However, to reduce the impact on the decontamination pad slab, the height is limited to 2'-3" by using the crane height limit switch when the impact limiter is not used.
The maximum acceptable drop was determined to be 2'8". In the event of this postulated 2'-8" drop, the cask was calculated to penetrate 0.09" into the concrete slab. At this height, the cask would perforate a 1.92" concrete element. The decontamination slab is 9" thick with a 9" thick concrete topping. The penetration and perforation calculations were performed with the modified Petry and Ballistic Research Laboratory formulas, respectively.
Ouestion 4 In your response to Questions 6 and 7 you address the potential impacts of the cask on the walls and components within the Decontamination Area and Spent Pool walls, as a result of the postulated cask tipping after a 4', 2', 3', and 42-7", height cask drops.
Please address the bases for the various selected drop heights and provide results that include forces and reactions related to the various postulated impacts and the respective stress levels as related to FSAR allowables.
Response
The postulated drops were evaluated at the heights of 10'-6" over the 4' section of the impact limiter, 4'-6" over the 2' section of the impact limiter, 2'-3" over the decontamination pad without the impact limiter, and 40'-3 3/4" in the spent fuel pool cask handling area. The postulated drop heights on the decontamination pad and the north turbine deck extension are described in more detail in Item 3 above and on pages 9-12 in the April 28, 1988 submittal.
-3 The cask impact forces on the decontamination pad, north turbine deck extension and the spent fuel pool cask handling area are provided in.
Question 5 Figure 4 shows a 3-D view of the impact limiter. However, it doesn't convey dimensional values of the limiter. Additional information is needed to fully describe the impact limiter and the various cask impact scenarios identified in your previous responses. Also, in the second paragraph of item 2 for the response to Question 6, you state conclusions with respect to portions of a postulated damaged wall falling in the new fuel areas. However, you do not provide analytical results supporting your conclusion that the new fuel racks will not be affected.
Response
The dimensions of the impact limiter are shown in Figure 1.
The new fuel racks will not be impacted in the event of the postulated drops or tipping due to the physical location of the new fuel racks and the masonry wall.
The racks are approximately 20 feet away from the wall.
In the worst case, the cask would strike the wall at a height of about 10 feet, the wall would be locally damaged and potential secondary missiles (i.e., concrete blocks) would be created. Since the blocks could travel approximately 10 feet and the separation between the racks and wall is 20 feet, the new fuel will not be affected.
The other critical direction of cask tipping is toward the spent fuel pool.
The closest point to the spent fuel racks that the cask could strike the northern reinforced masonry wall of the decontamination pad is at a height of about 10' above the floor elevation. The corner of the spent fuel rack is horizontally 10' away from this point and 27' below the operating floor level (Elev. 42').
Therefore, the spent fuel is far enough away to not be adversely affected by potential wall fragments.
Question 6 Your responses toQuestion 7 and 10 address the postulated cask drop in the spent fuel pool.
More information with regard to the details of this analysis are required to allow the staff to reach the same conclusions. This information should include: material properties, modeling, computational models, analysis procedures and results for the evaluation of the pool walls, slab, and liners. Also, you should address the drop orientations considered and the controlling drop orientation.
-4
Response
The north, south and east walls of the cask laydown area of the pool are 4 feet thick. The west wall that separates the cask laydown area from the upender is 2'-6" thick. The base mat is 4'-9" thick. The stainless steel liner thickness is 11 ga below Elevation 4'-0" and 16 ga above Elevation 4'-0".
The 2 1/4" thick liner protector plate is placed on top of the basemat liner. The concrete strength is 4500 psi and reinforcement details are shown in the drawings provided in response to Question 1. Also see the response to Question 4.
Question 7 In your response to Question 12, you describe the requirements for the turbine deck load bearing test. The loads address the previous mode of fuel movement (air pallet).
State why the new load should not be the 105 tons resulting from the new load (70-tons) increased by the dynamic factor.
Response
The turbine deck load bearing test was performed to verify the continued adequacy of the turbine post-tensioned concrete deck and the supporting members for spent fuel movement utilizing the air pallet system. The air pallet system loads the concrete deck and the supporting beams. When using the gantry crane for the spent fuel movement, no loads are transmitted to the concrete deck and the deck supporting beams. The loads are transmitted from the crane directly to the crane rail girders and then to the columns.
The girders and the columns stresses were verified to be within allowable limits as shown in Table 3 of the June 10, 1988 submittal.
Therefore, no new load test is being introduced or required for the proposed transshipment method.
Ouestion 8 In your response to Questions 9 and 10, you address several staff concerns.
We request the following clarifications:
o State the criteria that established the value of the 100 ton load used on the analysis of the gantry crane.
o The seismic accelerations considered for Seismic Category B loads (0.13g and 0.2g) are considerably different then the 0.67g identified in your response to Question 5. Address these differences.
o The 20 kips load on the crane identified in Item (c) fails to state if its direction was considered. Also, the rope capacity in item (e) is identified as 9 tons while the load on the crane is established as 20 kips. Discuss these staff concerns.
-5 o
Your response to Question 10 describes the structural components of the sliding roof. State if it is considered as a Seismic Category I structure and if it has been designed for the related FSAR requirements.
Response
- a. The 100 ton load was used because multi-element casks on the order of 70 to 100 tons were considered during the initial studies of transshipment alternatives. We are now using the GE IF-300 70 ton cask. However, modifications were designed with allowances to support 100 ton cask.
- b. As indicated in the response to Question 2, the Seismic Category A is.67g horizontal and.44g vertical and Seismic Category B is.2g horizontal and.13g vertical.
- c. The two cable restraints are designed for cask motion in the East-West direction. The 20 kips is based on the 100 ton cask.
Since the actual cask to be used is 70 tons, the specific design of the cables was limited to 9 tons each.
- d. The Fuel Storage Building was evaluated for the.67g modified Housner spectrum as part of the Seismic Reevaluation Program for SEP. This is the Seismic Category A criteria for SONGS 1. This is documented in SCE's submittals dated September 30, 1982 and December 20, 1982 and the NRC's SERs dated November 7, 1985 and July 11, 1986.
Question 9 Your response to Question 17 identifies the daily discharge, from the wells for the leak chase system of the spent fuel pool, as 10 gallons per day. From these results, it appears that the liner is not providing adequate leakage protection. Describe your plans for repair of the liner.
Response
The spent fuel pool has been evaluated to determine the extent of the liner leakage and determine possible repair. The liner is made of 11 gauge stainless steel with the 16 gauge stainless steel above the elevation 4'-0".
The leakage through the liner has not caused any structural damage to the spent fuel pool walls. Structural damage will not occur in the future as long as the monitor well which receives the leakage is pumped down.
It has been determined that there are approximately six leaks in the upender area of the spent fuel pool.
The leaks are located in vertical welds near the weir gate and in welds at the junction of the 11 gauge and 16 gauge steel plates.
The inspection of the weld leaks indicates three possible failure modes could have occurred:
(1) there appears to be a lack of complete fusion in the welds; (2) a low cycle fatigue mechanism due to the dewatering of the upender area for refueling; and (3) sulfur induced stress corrosion cracking of the welds.
-6 It also appears that there is leakage in the spent fuel pool or the cask handling area. Following the draining down of the upender area, leakage into the monitoring well occurs at 20 to 30 gallons per week.
SCE is currently looking into revising the water chemistry of the spent fuel pool to lower the sulphate limit. This would reduce the possibility of sulfur induced stress corrosion cracking. SCE is also evaluating repair of the welds by welding or epoxy. It is expected this would occur at the earliest after the Cycle 10 refueling and transshipment outage.
Question 10 Your responses provided in your submittal dated June 10, 1988, reworded some of the original questions provided by the staff. Although they include most of the requested information, they appear to have missed our Question 12 which said "State if you plan to provide a technical specification equivalent to Item 4.13, Attachment 1, for the proposed transport mode."
Please address this staff request.
Response
There are no plans to initiate a Technical Specification equivalent to Item 4.13, Attachment 1, for the proposed transport mode.
Since the proposed transshipment will no longer load the post-tensioned concrete deck, the above Technical Specification is being deleted as part of Amendment Application No. 148 dated April 28, 1988.
ACL:0186n CASK HANDLING AREA OF THE SPENT FUEL POOL WALL MOMENTS DUE TO CASK IMPACT Supports Mid Span HORIZONTAL MOMENT Impact load 1042 (k-in) 2410 (k-in)
Ultimate capacity 1128 (k-in) 2883 (k-in)
VERTICAL MOMENT Impact load 3157 (k-in) 997 (k-in)
Ultimate capacity 9280 (k-in) 1128 (k-in)
Wall is rectangular plate supported on three sides.
IMPACT ON BASEMAT Impact Force 162,303 (kips)
Basemat Capacity*
174,195 (kips)
- Based on soil bearing
70 Ton Cask Impact Loading Member Concrete Structural Impact Location Slab Steel Load Capacity Load Remarks Decon Pad Cantilever 87 psi (shear) 68 psi Between 76.5 psi (bending) 68 psi Beams Modified 428.6 (kips) 362 (kips)
Load Capacity W24 x 94 is based on Beams plastic capacity W8 x 40 344 (kips) 274 (kips)
Column (B7-9)
North West of 176 psi (bending) 68 ksi Turbine Crane Deck Extension W18 x 45 35.2 ksi (bending) 9.1 ksi (bending)
MC18 x 42 35.2 ksi (bending) 11.5 ksi (bending) 22.4 ksi (shear) 4 ksi (shear)
E E4 LL
GENERAL ELECTRIC NUCLEAR TECHNOLOGY & FUEL DIVISION GENERAL ELECTRIC COMPANY
- MORRIS OPERATION
- 7555 EAST COLLINS ROAD
- MORRIS, ILLINOIS 60450 (815) 942-5590 January 18, 1988 cc: G. T. Borst T. E. Ingels C. E. King Mr. Thomas W. Raidy S. P. Schmid Fuel Engineer T. E. Tehan Southern California Edison R. L. Shingleton - PNS P.O. Box 128 San Clemente, CA 92672
Dear Mr. Raidy:
The purpose of this letter.is to certify that Pacific Nuclear Systems, Inc.
Cask IF-301 has been inspected, tested, and is in compliance with Certificate of.Compliance 9001, Rev. 22 dated July 16, 1987. Annual testing of IF-301 was complete on January 13, 1988. Reinspection of IF-301 shall be scheduled prior to February 1, 1989.
The General Electric Company has a lease contract with Pacific Nuclear Systems, Inc. to use the IF-300 shipping casks (IF-301, IF-302 and IF-304).
One condition of the lease agreement is that the General Electric Company is responsible for performing periodic inspections while the lease is in effect.
We are an approved vendor of Pacific Nuclear Systems, Inc.
See attached copy of a letter from E. T. Brooks (PNS) to G. Smith (NPPD) dated August 7, 1987.
All IF-300 Casks continue to have Quality Assurance program approval under USNRC Approval Number 4, with an expiration date of May 31, 1990. Acceptance of the QA plan (NEDO-20776, Rev. 8) by the Commission implies that all activities important to safety concerning shipping Casks IF-301, IF-302 and IF-304 applicable to the design, fabrication, inspection, testing, purchase, use, maintenance, repair and appropriate levels of management are contained in quality assurance/quality control manuals.
This letter also certifies that the IF-301 Standard (Long) Yoke has been inspected and is in compliance with ANSI N14.6-1978. Inspection was complete on January 15, 1988. The yoke shall be scheduled for reinspection prior to February 1, 1989.
This letter also certifies that the IF-300 Offset Trunnions #9 and #10 have been inspected and are in compliance with ANSI N14.6-1978. Inspection was complete on December 15, 1987. Offset Trunnions #9 and #10 shall be scheduled for reinspection prior to January 1, 1989.
Please report any unusual event or occurrence that results in damage to the shipping casks, the IF-301 yoke or the trunnions. We shall evaluate action needed to maintain compliance to Certificate of Compliance 9001.
Mr. S. P. Schmid and Mr. T. E. Tehan are delegated as Level II Field Inspectors for this equipment.
Sincerel
- yours, L. L. Denio LLD:tp Supervisor, Quality Assurance & Safeguards
NUELE/5 e
A Phons NuwCapan.p February 8, 1988 Filer MM QA-STD.0 Southern California Edison Company San Onofre Nuclear Generating Station Railroad Tracks North of Unit #1 Gate San Clemente, CA 92672 ATTENTION:
S.W, Stilwagen, D-3B
REFERENCE:
NuPac WO# 3221 S.C.E. PO# SM127002 Modification of IF-300, M19-43 Yoke/
Outrigger System NuPac Dwg. #3221-100 CERTIFICATION Nuclear Packaging, Inc. certifies that all materials, processes and manufactured items provided for the modification and testing of the IF-300 M19-43 Yoke/Outrigger System assembly were per formed in accordance with the referenced S.C.E. contract.
All testing was performed in accordance with the provisions of ANSI N14.6.
All records pertaining to this contract are retained in the Nuclear Packaging, Inc. Quality Data Files.
Very truly yours, NUCLEAR PACKAGING, INC.
Joe R. Olivadoti Quality Assurance Director JRO/man Nuotear Pactkging, Inc.
100 Sou nr 336t1 Stueel Federal Wa) Washington 96003 (206) 874-2235 Telx. 152687 PNSI UD
TVIFIE eo NIJCLEfIIR February 19, 1988 SYSTENI5 Stephen W. Stilwagen Southern California Edison Company P. 0. Box 128 San Clemente, CA 92672
SUBJECT:
TEST AND INSPECTION CERTIFICATE
Reference:
SCE P.O. No. 8M127002; NuPac W.O. No. 3221, IF-300 Yoke and Related Lifting Equipment
Dear Steve:
Pacific Nuclear
- Systems, Inc. hereby certifies that all materiala, processes and manufactured items used in
- 1) the modification and testing of the IF-300 Yoke System and, 2) the fabrication and testing of the Hook Extension Assembly were performed in strict accordance with the requirements contained in SCE P.O. SM127002 and the PNSI Proposal No. SFSD-8701, Rev 01.
Testing of the following items was successfully performed in accordance with the applicable requirements of ANSI N14.6.
The test procedure and acceptance requirements are contained in the NuPac Procedure LOT-69, Rev 0 "Load Testing Procedure for the IF 300 Lifting Yoke With Modifications and Misc. Components".
- 1.
Modified IF-300 Yoke M19-43, S/N 302 (3221-110-Al)
- 2.
Hook Extension (3221-150-Al)
- 3.
Pin (3221-150-A2)
- d.
Rook Extension (150csa09)
- 5.
Sister Hook Adapter (829C5309)
- 6.
Pin (192B4346)
- 7.
Outrigger Stands (2) (3221-140-A1)
Pacifle Nuclear Systems, Inc. 10c Soulth 336*n $reet rederal Way. Washingtcn 98003 206) 874-2235 Telex 152667 PNSI UD
Stephen W. Stilwagen Page 2 February 19, 1988 All QA records pertaining to NuPac W.O. 3221 have been retained in the Nuclear Packaging,1 Inc.
Quality Data Files -and are available for review.
Sincerely, E~d r T. Brooks Corporate Quality Assurance Director cc:
T. E. Tehan W. C. Wheadon
-NU[EETR e#
A Pau ic NasCleat CompanV February 23, 1988 File: MM QA-STD.0 Southern California Edison Company.
San Onofre Nuclear Generating Station Railroad Tracks North of Unit #1 Gate San Clemente, CA 92672 ATTENTION:
S. W. Stilwagen, D-3B FEB29988
REFERENCE:
- 8M127002 Modification of IF-300, M4 Yoke/Outrigger System NuPac Dwg. #3221-100
-ERTIFICATION Rev. 1 Nuclear Packaging, Inc. certifies that all materials, processes and manufactured items provided for the modification and testina of the IF-300 M19-43 Yoke/Outrigger System assembly were performed in accordance with the referenced S.C.E. contract.
All testing was performed in accordance with the provisions of ANSI N14.6.
An inspection record copy of LOT-69, Load Testing Procedure for the IF-300 Lifting Yoke with Modifications & Misc.
Components is attached for your information.
All records pertaining to this contract are retained in the Nuclear Packaging, Inc. Quality Data Files.
Very truly yours, NUCLEAR PACKAGING, INC.
Joe R. Olivadoti Quality Assurance Director JRO/man Attachment Nuclear Packaging. Inc.
1H.
shngton 9003 (KC61 8742235 Telex. 1526b
- JUD
J IN10LE R(
aAi 1O 01pIIIJ(RUNE6 SPONIECTIONNmwc~~n LOAD TESTING PROCEDURE AID G.
C 0FOR THE IF-300 LIFTING YOKE WITH MODIFICATIONS & MISC. COMPONENTS LT-69 FO INION O
FORMI'vll0N OLY REVISION:
DATE:
-a ESSENTIA RELAT NUPAC DOCUENS The following related NuPac document(s) contain operations or information essential to performance of instructions herein and must be issued in conjunction with this document:
1.,%/'o--a
- 2.
- 3.
- 4.
- 5.
6..
Date Date Prepared By Prgam Manager Qnginring (Licensed Products) other Engineering (Matl Handlinga Ergineering (Analysis)
Engineering (Process Sys)
Q Manufacturing R
ase (Documentation Control)
Nuclear Packaging, Inc.
1010 50, 3316th Street Federa! Way. Washington 98003 (206) 874 2235 Tetex: 152667 PNSI UD
Procedure No.QL....-.69 Preparation Date 1-28-88 TITLE. IF-300 LIFTING YOKE WITH MODIFICATIONS & MISC. COMPONENT RECORD OF REVISIONS REV DESCRIPTION PAGE(S)
DATE SIGNATURE 0
Original release ALL LIST OF EFFECTIVE PAGES Page No.
Rev Pace No.
Rev PeNo.
Rev LOT-69, Rev 0 1/28/88 1.0 SCOPE The purpose of this proof test is to verify the structural integrity of the load carrying components and the functioning of the IF-300 YOKE outrigger operating system.
The load testing shall be performed after paint has been removed, by sandblasting, from all welds of the following components:
(a)
YOKE (outriggers may be painted prior to testing)
(b)
HOOK EXTENSION/S (c)
SISTER HOOK ADAPTER Aluminum and stainless steel components are not to be painted.
A copy of test weight scale tickets and load cell calibration documents to be attatched to the last page of this test agenda.
2.0 REFERENCE DOCUMENTS 2.1 YOKE LIFTING SYSTEM ASSEMBLY Dwg. 3221-100 IF-300 M19-43 S/N 302 2.2 HOOK EXTENSION fr-wg. 159C5309 23 SISTER HOOK Dwg. 829C5309 2.4 ASME Code, Section Ill. Division 1. Subsection NB. Article NB-5000 2.5 ASME Code,Section V Article 7 3.0 COMPONENTS TO BE TESTED 3.1.
Proof test modified IF-300 YOKE M 19-43 S/N 302 W/HOOK EXTENSION ASSEMBLY 3.2.
Proof te HOOK EXTENSION ASSEMBLY I59C5309 W/SISTER S'
SISTER HOOK ADAPTER 829El28 3.3.
Operationally test IF-300 YOKE M19-43 S/N 302 OUTRIGGERS 3.4.
Proof test IF-300 YOKE M19-43 S/N 302 OUTRIGGERS AND Po. 3 STANDS 3.5.
Operational test IF-300 YOKE M 19-43 S/N 302 LIFTING PIN SHIFTING MECHANISM Dasib l of 1
LOT-69, Rev 0 1/28/88 4.0. TEST WEIGHTS All proof tests shall be made using test loadss equal to 150 percent of component-rated capacity as follows:
For TEST A............Test load - 70 1 2000 x 1.5 - 210,000 LBS For TEST B.............Test load - 70 x 2000 x 1.5 + 6000 - 216.000 LBS For TEST C............. No test load required.
For TEST D............ Test load - 10,000 x 1.5 - 15,000 LBS For TEST E............. No test load required.
For TEST F............. No test load required.
All testing shall be witnessed by a Nuclear Packaging Inc. representative including Nuclear Packaging Inc. Q.A.
LOT-69. Rev. 0 1/28/88
5.0 TESTS
5.1 TEST A - PROOF TEST OF MODIFIED IF-300 YOKE S/N 302 W/HOOK EXTENSION ASSEMBLY: Prior to starting test, magnetic particle inspect per Ref. 2.4 & 2.5 all IF-300 YOKE welds for cracks or other type defects. Suspend IF-300 YOKE and HOOK EXTENSION as shown in FIGURE A. A calibrated load cell is required in the load line. A copy of the load cell calilbration documents shall be attatched at the end of this test agenda. Using the crane apply a test load of 210,000 pounds.
Hold the test load for 10 minutes. Remove the test load and visually check for permanent deformation of unit components. Magnetic particle inspect per Ref. 2.4 & 2.5 all welds of IF 300 YOKE for cracks.
Visually inspect all welds of HOOK EXTENSION.
YOKE pre-test mag partical inspection complete, no defec.
Set-up of components completed...........
Actual test load(recod) o0x) 4 unds Assembly suspended for 10 minutes............
Inspection for permanent deformation completed, no deformation..
f/
Invpection of all YOKE welds completed, no defect...............
Inspecation of all HOOK EXTENSION welds completed, no defects........
Test and inispection conpleted.......................
9o641...
COMMENTS NUCLEAR PA GING INC. ENGINEER NUCLEAR PACKAGIN INC. Q.A.
DATE Page 3 of 10
LOT-69, Rev. 0 1/28/88 5.2 TEST B-Proof test HOOK EXTENSION ASSEMBLY 159C5309 W/SISTER HOOK ADAPTER 829El28:
Prior to starting test. visually inspect all welds of HOOK EXTENSION ASSEMBLY 159C5309 W/SISTER HOOK ADAPTER 829E128 for cracks or other type defects. Assemble and suspend components as shown in FIGURE B. A calibrated load cell is required in load line. A copy of the load cell calilbration documents shall be attatched at the end of this test agenda. Using the crane apply a test load of 216,000 pounds.
Hold test load for 10 minutes. Remove the test load and visually inspect for permanent deformation of components. Visually inspect all welds for cracks.
Pre-test visual weld inspection completed, no defects....
Set-up of components complete Actual test Ioad(record) z2Ao14
- co Z/A/lounds Assembly suspended for 10 minutes..........................
Inspection for permanent deformation completed, no deformation..
Inspection of all welds completed, no defects..................
Test and inspection com pleted..................................
COMMENTS:
NU EAR PACK IING INC. ENGINEER DATE NUCLEAR CK DATE Page 4 of 10
LOT-69, Rev. 0 1/28/88 5.3 TEST.C - OPERATIONAL TEST OF IF-300 YOKE OUTRIGGERS:
Suspend IF-300 YOKE by lifting pin (HOOK EXTENSION NOT REQUIRED).
Using control valves and hoses with 90 psig air pressure, raise and lower outriggers, as shown in Figure C. ten (10) times. Inspect to assure that outriggers move freely without binding. HIE PRESSURIZED AIR IS TO BE FILTERED TO REMOVE OIL.
MOISTURE. AND CONTAMINANTS.
DO NOT LUBRICATE AIR AFTER FILTERING.
Pre-operational inspection complete, no defects..............
Set-up of components completed................................
CYCLES I 2 3 4 5 6 7 8 9 10 Cycle Outriggers ten (10) times..... S N
2 a R a B
A Record supply air pressure entering control valve 901 psig Record air pressure required to raise outrigger arms.2eI
/psig Record approx. time required to raise outrigger arms m9.
sec Record approx. time required to lower outrigger arms sec Operation is smooth and nonbinding Test and inspection com pleted.........................
B0.................................
COMMENTS-1 Jo rneojs df cL.ec.k iu, pre 5ure re u.e d.Air rr ed rej v d
rerew reqc Joe-Torn eai 0u -PAcE NUCL AR PACKAGING INC. ENGINEER DATE NUCLEAR ACKINGING I C. Q.A.
DATE Pae 5 of 10
LOT-69, Rev. 0 1/28/88 5.4 TEST.D - PROOF TEST OF IF-300 YOKE OUTRIGGERS:
SUSPEND IF-300 YOKE by lifting pin (HOOK EXTENSION NOT REQUIRED).
From four (4) eye pad cables suspend test weight of 15,000 LBS.,as shown in Figure D. Raise outriggers. Set unit on outriggers and outrigger stands. Outrigger stands are to be mounted sufficiently high to keep test load suspended. Hold load for 10 minutes. Remove load and visually inspect for permanent deformation of components.
Visually inspect all welds for cracks. Lift IF-300 YOKE from outrigger stands and lower outriggers. Raise and lower outriggers five (5) times checking to assure outriggers raise and lower smoothly without binding. NOTE: PRESSURIZED AIR IS TO BE FILTERED TO REMOVE OIL. MOISTURE. AND CONTAMINANTS.
DO NOT LUBRICATE AIR AFTER FILTERING.
Set-up of com ponents com pleted.............................................
Actual test weight(record) 1.SpoO 4/
?pounds Assembly suspended for 10 minutes.................
Inspection for permanent deformation completed, no deformation.
Inspection of all welds completed, no defects................
Pre-operational inspection complete. no defects..........
CYCLES 1 2 3 4 5 Cycle outriggers five (5) times.......................................
i R E
5/g Rccord supply air pressure entering control valve s i Record air pressure required to raise outrigger arms 6 6ee psig Record approx. time required to raise outrigger arms
.QL Record approx. time required to lower outrigger arms sec Operation is smooth and nonbinding...................................................
Test and inspection completed.........................................
6e (Continued on page 7)
Dave At If
LOT-69, Rev. 0 1/28/88 COMMENTS-*-10
~
d reauJ.(t?
pree~O(c NUCLEAR PACKA GINP INC. ENGINEER bATE NUCLEAR PA(ZAGING IN Q-A.
iE1
LOT-69, Rev. 0 1/28/88 5.5 TEST E - OPERATIONAL TEST OF IF-300 LIFTING PIN SHIFTING MECHANISM Remove load from LIFTING PIN. Using control valve and air as described in TEST C. disengage and engage LIFTING PIN ten (10) times. Check to assure that LIFTING PIN is engaging and disengaging smoothly and completely. NOTE: PRESSURIZED AIR IS TO BE FILTERED TO REMOVE OIL MOISTURE. AND CONTAMINANTS.
DO NOT LUBRICATE AIR AFTER FILTERING.
Pre-operational inspection complete, no defects..........
Set-up of components complete......................
CYCLES 1 2 3 4 5 6 7 8 9 10 Cycle LIFTING PIN ten (10) times.X&j iB
/
Record supply air pressure entering control valve 3
J
/
Record air pressure required to disengage pin c
s 2
Record air pressure required to engage pin psig Record time required to disengage p 4
econds a c.
Record time required to engage pin
-seconds -7ec.
Operation is smooth and non binding..............
.o...........................
Test and inspection completed...............
CYOMMENTS tma frd..d.A nkr"eA V4LLL CLEAR P CKAGING INC. ENGINEER DATE NUCLE AR PACKAtING INC. Q.A.
DATE
LOT-69, Rev. 0 1/28/88 5.6 TEST F - OPERATIONAL TEST OF IF-300 LIFTING PIN LOCK Remove load from LIFTING PIN. Engage LIFTING PIN. Using control valve and air as described in TEST C. disengage and engage PIN LOCK ten (10) times. Check to assure that lock is engaging and disengaging smoothly and completely. NOTE: PRESSURIZED AIR IS TO BE FILTERED TO REMOVE OIL. MOISTURE. AND CONTAMINANTS.
DO NOT LUBRICATE AIR AFTER FILTERING.
Pre-operational inspection complete, no defects.....
Set-up of components completed..........................
CYCLES 1 2 3 4 5 6 7 8 9 10 Cycle PIN LOCK ten (10) times....
R M f18 21 ft S Record supply air pressure entering control valve 6?-'o psig '
Record air pressure required to disengage pin (7
+
Record approximate time required to disengage pin
/
sec Record approximate time required to engage pin 1-2sec Operation is smooth and non binding.......
Test and inspection completed COMMENTS NUCLEAR PACKAGING INC. ENGINEER DATE NUCLE PACKAGI INC. Q.A.
DATE Cromant 10
LOT-69, Rev. 0 1/28/88 6.0 INSPECTION VERIFICATION AND ACCEPTANCE The above tests were performed as described or as modified in comments and are certified to be complete and accurate:
NUCLEAR PACKAGING INC. ENGINEER DATE NUCLEAR PACKAGING INC. Q.ATE ATTACH TEST WEIGHT SCALE TICKETS AND LOAD CELL CALIBRATION CERTIFICATE/S AS APPLICABLE.
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March 09, 1988
[IFI[.
Ref: ETB 5000.03098 Mr. Thomas Raidy Southern California Edison Conmpany San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, California 92672 Mr Raidy:
Attached is a Product Quality Certification (PQC) that addresses the annual inspection of the Pacific Nuclear Systems IF-304 short yoke. Annual inspection in accordance with the requirements of ANSI N14.6-1978, was completed on 3 March 1988 at the GE-Morris Operation in Morris, Illinois., As per ANSI N14.6 1978, the test and certification will be valid for a maximum of one year or until 3 March 1989.
Supporting documentation for the tests is maintained at the GE-Morris facility.
Sincerely, Ed T. Brooks Di r, Corporate QA Pacific Nuclear Systems, Inc. 1010 South 336th Streei Federal way. Washington 98003 (206; 874.2235 Tele\\ 152667 PNSi UD
jkfRCV 5Y'Facif iC NUC!ear System; 3-i-88 7
- 02AM
- CCITT 324 120557424014 GENERAL*,ELECTRIC NUCLEAR ENERGY BUSINESS OPERATIONS PRODUCT QUALITY CERTIFICATION CU gowliPQ.f C
ROOUTP~Dc?
m'm MPL Wt.
Pacific Nuclear Syste-ms IF-304 Short Yoke N/A PAA/DAWING NO, RY AI REV OO-Q~
O.ANTI l'f THIS IS 70 CERTIFY THAT THE PRODUCTS IDENTIFiED HEREIN YiAVE BEEN IWANUFACTURED UNDER A CONTROLLED QUJALITY ASSURANCE! PROGRAM ANDO ARE IN CONFORMANCE WITH THE PROCIAEMENT QUALITY REQUIREMENTS INCLUDIN: APPL!CABLE CODES, STAND~ARDS AND SPECIFICATIONS AS IDENTIFIED IN THE ABSOVE-REFERENCED DOCUMENTS, UNLESS NOTED BELOW. ANY SUPPORTING DOCUMENTATiON S EITHER ATTACHED, OR WILL BE FlORWARDED OR RETAINED IN ACCORDANCE WITH CONTRACTUAL REQUIREMENTS.
SIGNED:_______________
VATE' March 9, 1988 TITLE' -Supervisor, QA &Safeguards ORGAN!ZATiON:
GE-Morris Operation NONCONPORMANCES FROM PROCUREMENT QUALITY REQUIREMENTS7 NONE REMAPI(S/EQUIPMENT SERIAL NUMBERS:
THE PURPOSE OF THIS.CERTIFICATION IS TO DOCUMENT THAT THE IF-304 SHORT YOKE HAS BEEN !%SPECTED AND TESTED TO THE ANNUAL REQUIREMENTS OF GEI-928218 (5/86)0 "1F-300 IRRADIATED FUEL SHIPPING CASK - MAINTrENANCE INSTRUCTIONS-" THEREFORE, THE YOKE IS IN COMPLIANCE WITH THE REQUIREMENTS OF ANSI t114.6-1978.