ML20217H844
| ML20217H844 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/22/1997 |
| From: | Hanshaw B CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20217H831 | List: |
| References | |
| 97-00362, 97-00362-R00, 97-362, 97-362-R, NUDOCS 9708130289 | |
| Download: ML20217H844 (41) | |
Text
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ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 NRC DOCKET NOS. 50-325 AND 50-324 OPERATING LICENSE NOS. DPR-71 AND DPR-62 REQUEST FOR LICENSE AMENDMENTS SPENT FUEL CASK HANDLING DESIGN AND LICENSING BASIS ISSUES ENGINEERING SERVICE REQUEST 97-00362 REVISION 0 l
L IF-300 CASK DESIGN, TESTING, AND INSPECTION
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9708130289 970806 Y PDR ADOCK 05000324 P PDR ,
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Form 1 ENGINEERING SERVICE REQUEST ESR8 9700362 Rev 8 0 WR/JO 8 Other Documente (CR, OEF, etc.)
CR 9701282 LER 97 004
.p p Primary System 8 Primary System Name Muttiple Systems 7105 F16-SPENT FUEL SYSTEM Affected Title Originator / Phone IF 3OO Cask Design, Testing, and Inspection HENSHAW, BRYAN KEITH /850-3147 Plant Customers (Pnnt Name, S6gn, Date) Engineering / Plant Programs (Pnnt Name, Sqn, Date) \
Nuct Fuels g g 5.(risa w dyrA swmM N e5HANs 7/,,/ v7 Regulatory Aff airs ~% Pnsley h hd ?fil fil iv -
Reviews (Pnnt Name, Sgn, Date)
DesignVenficaten 3,j;JQ,,f Md* M/t )
O Other Reviews Required O Records Attached Engineering Dieciplines (Pnnt Name, Sgn, Date)
Mechanical y,g [m ,,f d f/
Civil / Seismic c(,ugwy 'j7Mmte /fLO . I'.
Materials M#fM4 Ft hv7N4/F Response Type CC DOC CHANGE ESR N/A Quality Class Team A Safety-Related Due Date 07-08-97 APPROVALS le a 10CFR 50.59 Safety Review required per 3 NAS Defore Approval / implementation (plant specific procedure)7 NAS Before Closecut Yes O Safety Screen ONLY 3 PNSC Before Approval / implementation a NRC Before implementation @ USOD
. @ Plant General Manager O N/A (Engineering Dispositen Only)
Responelbie Engineer - BRYAN KEITH HENSHAW ,g ,
Responelbie Manager (Pnnt Name, Sign, Dat 4M d8fdM Y)
Plant General Manager (Print Name, Sign. Date Ngh ,
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Procedure Form EGR-NGGC-0005-13 DCM01a2a Oh1/97
,- 'i-Page a cf _
Form 1 - ENGINEERING SERVICE REQUEST ESR # Revs Tnk 9700362 0 IF 300 Cask Design. Testing, and inspection nequest--
-In correspondence to the NRC dated November 16, 1982, CP&L stated that " the cask' redundant lifting yoke and the work basket redundant lifting rig are of redundant design, and the crane on which they are used is single-failure proof.
Therefore a drop with regard to equipment handled by the above_ systems is not
. considered credible." During preparation of a Request.For Additional Information (RAI) ~ to Bulletin 96-02 for the NRC, CP&L discovered that the transfer of the cask from the tilting cradle to the secondary yoke involves a I
non-redundant lift. This lift / transfer occurs on the rail car at the 20' elevation of the reactor building.
This lift was contrary to statements made to the NRC, and a review of NUREG 0612," Control of Heavy Loads at Nuclear Power Plants" revealed that since the yoke did_not meet single failure proof guidelines, a load drop analysis would be required. CP&L reported this situation to the NRC in LER 97-004 as being outside of-design basis, because we do not own a load drop analysis for a cask drop. This analysis would introduce an accident not previously analyzed for at BNP.
NUREG 0612 is a guide for heavy load lifts over fuel / safe shutdown equipment
-which was published after the spent fuel shipping cask yoke was in service.
BNP design basis is that a load drop will,not occur. A review of the spent fuel cask lifting yoke's design margin, past load tests, inspections, and lift history should be performed to determine if a load drop is a credible accident.
If a load drop is not credible, Brunswick is not outside of design basis and the LER should be retracted, i
Procedure: Form EGR-NGGC-0005-1-3 DCMO3 03/14/96
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Page 3 cf _
Form 1 ENIINEERING SERVICE REQUEST
[SR # rie * # Title 9700362- -O- IF 300 Cask Design. Testing and inspection
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'See attached.
, Keywordst. Spent Fuel Shipping Cask, IF-300, Load Drop Analysis, NUREG 0612, l
~ ANSI N14.6, Heavy Loads l_
The purpose of this ESR is to show that a load drop from material failure of l the IF-300 Spent Fuel Shipping Cask Redundant Yoke is not credible while using existing site procedures.
I l
Procedure: Form EGR-NGGC4X)051-3 DCM04 OY28/96
o ESR 97-00362 Page 4 Rev.O List of Effective Pages:
Cover Sheets:
Page 1 2 3
-Rev. 0 0 0- l Design Section:
Page 4 5 6 7 8. 9 10 11 12 13 14 Rev. 0 0 0 0 0 0 0 0 0 0 0 Page 15 16 17 Rev. 0 0- 0 Reviews:
Page 18 19 20 21 22 23 24 25
-Rev. 0 0 0 0 0 0 0 0 Attachment A: Fracture Toughness input -
All pages - Revision 0
=.
, 3 ESR 97-00362 Page5 Rev. 0 ;
Table of Contents
' Cover Sheets 1 List of Effective Pages .4 Table of Contents 5 DESIGN SECTION Problem Identification Problem Statement - 6 ESR Design Specification =
ESR Functional Requirements 6
' References 7 Design inputs - 8
- Assumptions 10 Evaluation 11 Document Update Section Document Update Form 13 FSAR Change Package 14 -
Installation Section N/A Reviews 10CFR50.59 Safety Review - '18 Design Verification Form 25 Attachments Attachment A Fractureloughnessinput '
ESR 97 00362 Page 6 Rev.O DESIGN SECTION:
Problem Identification:
Problem Statement:
In previous correspondence to the NRC, CP&L has always maintained that in the process of transferring spent fuel for shipment, the spent fuel shipping cask could not be inadvertently dropped due to redundancy in all vital portions of the cask hoisting mechanism (Ref. 32). In order to respond to the NRC's Request for Additional Information (RAI) Regarding NRC Bulletin IE %02, a review of the spent fuel shipping cask handling activities was undertaken. This review confirmed that brunswick Nuclear Plant has a single-failure-proof crane and that cask handling activities on the refuel floor and in the spent fuel pool employ the use of redundant rigging. However, it was recognized that during loading and unloading of the spent fuel shipping cask from the railcar inside the reactor building at elevation 20':
a the design of the redundant yoke requires lifting of the cask with only the primary yoke (single lifting device)in order to perform the installation of the secondary yoke and obtain redundancy.
m If this activity resulted in an inadvertent drop of the spent fuel shipping cask, safe shutdown equipment could be damaged.
ESR Design Specification:
l ESR Functional Requirements:
1 This ESR will evaluate the structural integrity of the spent fuel cask primary lifling yoke and determine if a load drop is credible when used in accordance with existing procedures.
t i
ESR 97-00362 Page 7 Rev.O
References:
This list represents the majority of all correspondence and regulatory information relative to Brunswick Plant and spent fuel shipping cask handling. For the most part, the information is in chronological order. A review of this information is necessary to determine design and licensing basis requirements.
- 2. Safety Evaluation of the Brunswick Steam Electric Station, November 1973.
- 3. Supplement No.1 To The Safety Evaluation of BSEP, January 31,1974.
- 4. Branch Technical Position APSCO 9-1 " Overhead llandling Systems For Nuclear Power Plants," April 1975.
- 5. Letter from CP&L to NRC, Jones to Rusche, dated June 18,1976.
- 6. Reg Guide 1.104," Overhead Crane Handling Systems For Nuclear Power Plants," Feb.1976.
- 7. Supplement No. 4 To The Safety Evaluation of BSEP, September 1976.
- 8. Letter from NRC to CP&L, A, Schwencer to J.A. Jones, dated August 26,1977
- 9. General Electric to CP&L letter dated June 9,1981.
- 10. NRC's December 22,1980 letter to Licensee's regarding NUREG 0612.
I 1. Letter from CP&L to NRC, Utley to Ippolito, dated June 22,1981.
- 12. Letter from CP&L to NRC, Utley to Ippolito, dated September 22,1981.
l 13. DRAFT TER, from NRC to CP&L, dated July 9,1982 (prepared by Franklin Research Center).
I 14. Letter from CP&L to NRC, Zimmerman to Vassallo, dated November 16,1982.
- 15. Letter from NRC to CP&L, Vassallo to Utley, dated November 3,1983.
- 16. Letter from CP&L to NRC, Zimmerman to Vassallo, dated February 03,1984.
- 17. Letter from CP&L to NRC, Zimmerman to Vassallo, dated March 20,1984.
1S. Letter from NRC to CP&l , Vassallo to Utley, dated May 18,1984,
- 19. Generic Letter 8511 from NRC (Thompson) to Licensees, dated June 28,1985. (GL 85-11).
- 20. NRC Bulletin 96-02, dated April 11,1996.
- 21. Letter from CP&L (Campbell) to NRC, dated May 10,1996.
- 22. Letter from NRC to CP&L, Trimble to Campbell,(NRC's Request for Additional Information Related to Bulletin 96-02), dated Dec. 06, I996.
- 23. Letter from CP&L to NRC (BSEP 97-0039), Campbell to Gentlemen, dated February 07,1997.
- 24. Letter from CP&L to NRC (BSEP 97-0235 containing LER l-97-004), dated June 05,1997.
- 25. NEDO-10084-4, Vectra IF-300 Shipping Cask Consolidated Safety Analysis Report, March 1995.
- 26. BSEP Spec. 005-011 Rev.1. Appendix D, page D57
- 27. CMAA Spec. 70
- 28. gel-92817D, Operating Instructions IF 300 Irradiated Fuel Transportation System, T.E. Tchan, December,1992.
- 29. ANSI N14.6 - 1978.
- 30. NUREG 0612," Control of fleevy :.oads at Nuclear Power Plants."
- 31. Letter from NRC to CP&L, Taylor to Smith, dated October 9,1996.
- 32. UFSAR Section 9.l.
ESP. 97 00362 PageS Rev.O Design inputs / Commitments:
Basic Functions - ne basic function of the cask lifting yoke is to transfer the weight of the IF-300 cask to the Reactor Building crane main book while maintaining structural integrity during cask transport from the rail car to the refueling Door and spent fuel pool, ne source of the majority of the design inputs is from correspondence between CP&L and the NRC, Other input is obtained from vendors, reg guides, NUREGs, and ANSI standards. A summary of these design inputs and commitments is given below: A ne objective of part 1 of NUREG 0612 was to ensure that all load handling systems at nuclear power plants are designed and operated so that their probability of failure is uniformly small and appropriate for the critical tasks in which they are employed. ne general guidelines are given in Section 5.1.1 of the NUREG (Ref.18).
He section applicable to the cask lifting yoke is 5.1.l(4) which states:
"Special lifting devices should satisfy the guidelines of ANSI N14.61978 " Standard for Special Lifting Devices Ier Shipping Containers Weighing 10,000 pounds (4500 kg) or More for Nuclear Materials."
ne standa:1 should apply to all special lifting devices which carry heavy loads in the areas as defined above. For ooeratine olants certain inspections and load tests may be accepted in lieu of certain material reauirements in the standard, in addition, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device based on characteristics of the crane which will be used ' This is in lieu of the guideline of Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor on only the weight (static load) of the load and intervening components of the special handling device.
- For the purpose of selecting the proper sling, loads imposed by the SSE need not be included in the dynamic loads imposed on the sling or lifting device."
On Dec. 22,1980, the NRC requested verification that special lifting devices comply with the requirements of ANSI N14.61978. Where these standards, as supplemented by NUREG 0612, Section 5.1.l(4) were not met, the NRC requested a description of proposed alternatives and a demonstration of equivalency in terms ofload-handling reliability (Ref.10).
In letters to the NRC on June 22,1981 and on November 16,1982 (Ref. Ii & 14), CP&L stated that the " cask redundant lifting yoke and the work basket redundant lifting tig are of redundant design, and the crane on which they are used is single failure-proof. Derefore, a load drop with regard to equipment hanoled by the above systems is not considered credible."
For a short time period, during transfer of the cask from the tilting cradle to the secondary yoke on the 20' of the Reactor Building, this redundancy does not exist. In addition, this activity is being performed in the proximity of equipment needed for safe shutdown. Since a single lifting device is provided instead of a dual lifting device, there is an increase in the probability of a load drop and a potential unreviewed safety question exists.
\ .
ESR 97-00362 Page 9 Rev.O In 1980, the NRC required assurance that Brunswick heavy loads handling practices either ensured that the probability of a load drop was made extremely low through enhanced design features, or that the consequences of a failure were acceptable. CP&L responded that due to redundancy of the yokes, a load drop was not credible. The NRC approved our cask handling activity based on this response. Since during the short time period of time when the redundant lifting rig is bdng installed the cask lift is not redundant, BNP has operated outside oflicensing basis.'
We reported to the NRC that due to redundancy in the yoke, a load drop was not credible. The function of the yoke is to prevent a load drop by maintaining its structural integrity, flowever, redundancy is not the only acceptable method of proving that a load drop is not credible. For example, ANSI N14.6, Section 6,"Special Lifting Devices for Critical Lifts" allows increased stress design factors to be used where redundancy is not incorporated. Also, NUREG 0612, Section 5.1.l(4) states that for operating plants certain inspections and load tests may be accepted in lieu of certain material requirements in the standard. Based on an analysis of .
Brunswick specific design, past load tests, and inspection, a load drop at Brunswick is not considered a credible event. Herefore, since the design basis function of preventing a load drop is not compromised, and since redundancy is not the only method for proving that a load drop is not credible, BNP did not operate outside of design basis.8
' According to letter to CP&L (Sherwood Smith) from the NRC (James Taylor), dated October 9,1996,"De liccasing basis for a plant originally consists of that set ofinformation upon which the commission, m issuing an initial operating license, based its comprehensive determination that the design, construction, and proposed operation of the facility satisfied the Commission's requirements and provided reasonable assurance of adequate protection to public health and safety and common defense and security The licensing basis evolves and is modified throughout a plant's licensing term as a result of the Commission's continuing regulatory activities, as well as the activities of the licet,see."
- In letter to CP&L from the NRC dated October 9,1996, design bases was defined as,"that information which identifies the specific functions to be performed by the structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design..."
nV
ESR 97-00362 Page 10 Rev.O Design inputs / Commitments (continued):
ANSI N14.6 Input: ANSI N14.6 requires a safety factor of three with respect to yield strength and five with respect to ultimate strength for redundant lifting devices. These safety factors are to be doubled for non.
redundant lifling devices. Although ANSI N14.6 requires only consideration of static loads, NUREG-0612 supplements the ANSI standard by requesting that dynamic loads also be considered. It should be noted that when using materials with yield strengths in excess of 80% of their ultimate strene.ths, the stress safety factors discussed above do not apply per ANSI N14.6. Design shall then be bawd on the material's fracture toughness, and the designer shall establish the criteria.
De safety factors for each component of the primary yoke are given below (Ref. 9) 'Ihe safety factors are based on static loading. Based upon guidance given in CMAA Spec. 70, dynamic loading for a crane with maximum hoisting speed of 3 fpm, such as the reactor building crane, would be 1.5%. Per Ref.18, the NRC considers our dynamic loads small and allows us to disregard them.
Safety Factor Safety Factor Yield Ultimate Component (w/ respect to vield) (w/ respect to ultimate) Material Strength Strength l Cross Member 7.92 9.24 A514 90 ksi 105 ksi l Arms 20.52 23.94 A514 90 ksi 105 ksi l - J-hooks 3.63 4.24 A514 90 ksi i
105 ksi Yoke Pin 6.21 " "
4340 Steel 125 kal Worst Case Welds 6.18 " " " "
" Not Available Since the Cross Members, Arms, and J-hooks are all made of ASTM A514 material which has a yield strength in excess of 80% of ultimata strength, the evaluation will base acceptability of these components on the material's fracture toughness. He Yoke Pin and Worst Case Welds will be evaluated based on safety margin and past load tests since the ultimate strength is not available and a safety factor of 10 with respect to ultimate strength cannot be demonstrated. Since ANSI N14.6-1978 did not exist at the time the yoke was fabricated, the option given in NUREG-0612,5.1.l(4) which states "For operating plants certain inspections and load tests may be accepted in lieu of certain material requirements in the standard" will be used.
Assumptions:
- 1. Ilased on review of references I through 10, it will be assumed that the December 22,1980 letter from the NRC to Licensees regarding NUREG 0612 constitutes the comprehensive guide to heavy loads handling at nuclear power plants and supersedes previous guidance,
- 2. Dye penetrant and'or magnetic particle tests are performed per OPM-CRN503 on an annual basis. Although these tests should detect all surface cracks, it is assumed that cracks of up to 1/16" in length may be missed due to the cleaning process. See Attachment A for excerpts from procedures and input from CP&L NDE Level 111 which give ensurance that cracks of 1/16" or greater in length will be detected in PM routes. Site procedures for NDE (NDEP-301 and NDEP 201) require test rejection for any indications determined to be cracks or linear indications. Per these procedures, linear indications are those indications in which length is more than three times the width. A characteristic of fatigue cracks is that depth of crack is less than or equal to length of crack. It will also be assumed that since inspections will detect cracks greater than or equal to 1/16" long, we will also detect any cracks greater than 1/16" in depth.
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ESR 97-00362 Page1I Rev.0 Evaluation:
Analysis of IF 300 Yoke (Cross Member. Ilooks. and J-hooks) based on Material Fracture Touchness Of the three components mentioned above, the J-hooks had by far the smallest safety factor and are the critical components. He weight of a fully loaded IF 300 cask is 140,000#. Although this is a constant loading, we will consider dynamic loading from crane uplift and the unlikely event of an carthquake as load variations.
Combining all into one load case, we find the maximum possible loading to be as follows:
Maximum Loads (Design Basis Earthquake):
Cask Weight 140,000#
Dynamic Loading (1.5%) 2,100#
Seismic Loading (.345g) 48.300#
190,400#
Note: Cask Weight is obtained from Vectra Consolidated Safety Analysis Report. Dynamic Loading is based on .5% of static load for each foot per minute of crane hook speed (per CMAA Spec. 70). Reactor Building crane maximum hoisting speed is 3 fpm. For seismic loading, the wire rope will be considered as rigid in the vertical direction. BSEP Spec. 005-011, Appendix C, Page D57 gives a peak vertical acceleration of.23g under DBE conditions. Seven percent damping was used for cranes as specified on Page 18 of the specification. Assuming the crane girders act in a flexible manner, the maximum vertical acceleration is taken as 1.5 times the peak acceleration. Herefore, seismic loading will be .345g.
From original G.E. calculations, the J-hooks have a safety factor of 3.63 with respect to yield strength and static loading. The minimum yield strength in the material (ASTM A514)is 90 ksi. Herefore, the maximum actual stress during a scismic event based on loading given above would be 90 ksi x (190,400# / (140,000# x 3.63)] = 33.7I ksi.
Equation Used: yield strength x [DBE loading / (S.F. x cask weight)) = stress This analysis is based on fracture mechanics. Basic inputs are 1) the maximum crack size that may go undetected during NDE and 2) the applied stress. Using these inputs, a stress intensity factor may be calculated. If the calculated stress intensity factor is less than the critical stress intensity factor for the material, then an undetected crack smaller than the size postulated will not grow in an unstable manner. Our acceptance criteria shall be that stresses during the design basis earthquake shall not cause a crack of 1/16" to grow in an unstable manner.
A maximum undetected crack size of 1/16" will be postulated with an applied stress of 33.71 ksi. Per Attachment A, the calculated stress intensity factor of the 1/16" crack (16.7 ksiVinch) is much less than the critical stress intensity factor for ASTM A514 steel (150-177 kslVinch). Derefore, an undetected crack of 1/16" will not grow in the yoke in an unstable manner.
Conclusion:
Therefore, based on fracture toughness of material, a load drop of the IF-300 spent fuel shipping cask due to material failure of the cross member, arms, or J-hooks is not a credible accident.
Analysis ofIF-300 Yoke (Worst Case Welds and Yoke Pin)
As stated earlier, ANSI N14.6-1978 did not exist at the time the yoke was fabricated, and it is appropriate to use the option given in NUREG-0612,5.1.l(4) which states "For operating plants certain inspections and load tests may be accepted in lieu of certain material requirements in the standard." his analysis will be based on existing safety factors with respect to allowable yield stress, past load tests, and periodic inspection program.
The purpose of safety factors in design of stmetural members is to compensate for factors such as the variation in quality of materials and fabrication, possible overloads, secondary stresses due to errors introduced by design assumptions, and approximations in calculation procedures.
ESR 97-00362 Page 12 Rev.O ne worst case welds have a safety factor or 618 under static loading with respect to yield strength. Original calculations do not specify material type or ultimate strength of weld material. A safety factor of 6.18 represents a load capacity of 6.18 x cask weight = 865,200#. A load test of 1.5'(W + impact) was performed independently on the primary and secondary yokes by the vendor prior to use (Ref. 28, page A-4). CP&L performed additional load tests. According to Reference 17 "the Spent Fuel Cask Redundant Yoke Assembly was originally tested to 200% of rated load. In 1980, the Spent Fuel Cask Redundant Yoke Assembly was modified and retested to 200*6 of rated load." ne secondary yoke would not have carried any lud in this test configuration (Ref 9). Herefore, the primary yoke itself has been tested to 200% of rated load. Rated load is a minimum of 140,000#, which is the weight of a fully loaded IF 300 cask. Since the load tests were successful, it can be concluded that there are no material defects in the primary yoke ( that afTect integrity of the primary yoke when subjected to loads less than or equal to 200% of rated capacity). In addition, an inspection program which is in compliance with standards set forth in ANSI N14.6 ensures that periodic dye penetrant and/or magnetic particle testing are performed on critical areas of the primary yoke. These actions have ensured that the yoke's ability to withstand loading of 200% of rated load throughout its service life. To summarize, the welds are designed to withstand a loading of 865,200# before yielding, and past load tests and l periodic inspections give confidence that the welds will support a load of at least 280,000# without any observable effects.
He highest anticipated load the yoke will ever experience is 190,400 # in a seismic event. Since the yoke has been load tested to 280,000 #, the probability of load exceeding the capacity of the yoke is negligible.
The previous 200% load test, analysis of yoke under site specific maximum postulated accident loads, and an inspection program which complies to ANSI N14.6 (Ref.17) ensures that a cask drop is not a credible event.
It must also be considered that the evolution is of short duration and is required to place the cask lifting yoke into its redundant configuration.
Conclusion:
He maximum load under design basis accident conditions is much less than both design and load tested capacities. Derefore, a failure of the primary yoke's worst case welds or 6" pin, and a resultant drop of the IF 300 spent fuel shipping cask is not credible.
Overall
Conclusion:
The objective of part 1 of NUREG 0612 is to ensure that all load handling systems at nuclear power plants are designed and operated so tht their probability of failure is uniformly small and appropriate for the critical tasks in which they are employed. Although BNP was outside its licensing basis, the objective of NUREG-0612 was met because a load drop is not credible.
Note: Altemate calculations using AISC Code:
De primary lifting yoke was originally analyzed based on static load with safety factors being based on material minimum yield strength (Ref. 9). The limiting component was the Lhook with a safety factor of 3.63. AISC normal allowables are based on .6 X yield strength. His equates to an allowable load of 304,920#. Under design bases loading conditions, the hook would see a load of 190,400# Since 190,400 is much less than normal AISC allowable load, even under design basis (earthquake) conditions, the yoke will maintain structural integrity.
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Sheet 1 of 1
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ESR 97 00362 Page 14 Rev.0 Log No. Form ORCl 04.1-1 BSEP FSAR Change Review and Approval Form
Subject:
Method of Soent Fuel Shippina Cask Handlina Unit ,JA2 (Briet Description of Change)
Initiating Document: ESR 97-00362 Onclude Rev. #1 Basis for Change:
The existing UFSAR, section 9.1.4.2.2 is vague with respect to spent fuel shipping cask transfer from the tilting cradle to the primary yoke. Although it is stated that the cask is uprighted with only the primary yoke, it is unclear as to whether an actual lift occurs during transfer. Clarification was added to this section. Also, Section 9.1.4.2.3.2, Safety Analysis states that redundancy is employed in all vital portions of the cask hoisting mechanism. This is true in the vicinity of fuel, but not true during cask transfer from tilting cradle to secondary yoke on the 20' elevation of the Reactor Building. An exception has been added to cover this short period of time.
Affected Sections, Tables,9.1.4.2.2
~
Figures: (attach mark-ups) 9.1.4.2.3.2 Effective Date: 8/15/97 TARGET ACTUAL Safety Review Package Attached: y_gg No t
l (if 'No", Provide Justification Bdow)
Initiator: F _ M MA/ A N N h / K ms/Cf / 7/2//D (Pnnti 1 -
Organization Date App oval / /
(Sign) U Organization ' Qate '
NOTE: Attach additional sheets as necessary to assure completeness.
NOTE: Applicable review / approvals to be determined by the Supervisor, Licensing or designee.
NOTE: Use current updated FSAR pages for mark-up.
^ ' g S g 9 7 o C 3r.2 MMSla L" protection: allowable spent fuel shippin conditions: Shielding; and continuity of decay heat removal capacit for all credible cask accident events. The maximum weight of the loade cask must not exceed 75 tons. The 75 ton maximum cask weight was established by the spent fuel pool structural analysis.
9.1.4.2 2 Spent Fuel Shipping Cask Handling Saent fuel casks are delivered to the plant on railcars or flat bed trailers w11ch enter the Reactor Building through an airlock at Elevation 20.0 ft. The casks are transported in a horizontal position. The railway car or tr 'ler is
- cositioned under the equipment access _ hatch to_facilitata handling After pp.OP inspect.ing-ana removal or any crasrr structure ena tie downs, the ask is y:
raised to a vertical position using a single lifting yoke attached to the
-Reactor Building crane. In the event the cask must be left standing for any M^""g extended period, tie down points and cables are provided to prevent possible F*1' MW tipping during a seismic event. 7ENT.
The single lifting yoke is replaced with a double lifting yoke which provides redundant lifting capacity (see Section 9.1.4.2.3). The cask i -
NW throuah -
the acu_ioment access hatrh to clear ths refullina floor /s non miw ElevatRin 117.3 f t (Figure 1.2.2-6) aniFplaced in tne deconla7flination ar?a. adjacent to the hatchway.
l cask is lifted using the double yoke and placed in the fuel pool.After In cleaning. inspection, and i traversing the path from access hatch to its resting position in the pool, the
) cask at no time As a backup to operator action, passes limitin over any other part of the fuel pool.
travel over spent fuel, g stops in the crane control system prevent unwantedas recommende Facility Design Basis."
After the cask is loaded the cask closure head is set in place on the cask.
The redundant lifting yoke is then attached and the cask raised to permit placement of some of the head bolts or nuts. As cask raising continues. the cask is-hosed down and the radiation dose rate monitored. Tne cask is then l moved to the decontamination area and surfaces are decontaminated. A water jet System is available for cleaning the casks. as well as the dryer-separator and reactor well areas. After decontamination is completed, the remaining head bolts or nuts are installed and tightened. g surface contamination and tightness. The cask is then checked pyg for g(A1tKurg he redundant yokt is aaaili u ed to_ return the_ cask to tha t" nsoort vehicle.
iter cask ~ temperature ano p essure navelermonitoreo and equii1DTO --
determined, tie down bolts are installed and shiccing placards attached. As WM(g discussed in Section 9.1.4.2.3, the loaded fuel cask does not exceed the rated crane capacity.
9.1.4.2.3 Spent Fuel Shipping Cask Hoisting 9.1.4.2.3.1 Description The crane was designed so that its critical components and functions provide redundant capabilities (Figure 9.1.41) during handling and moving of the spent fuel cask. A description of these components and functions follows:
9.1 4 2 Mendment No. 8 1
CSR 97- o0%2 BSEP 1 & 2 E'EV. G)
UPDATED FSAR FM C (fo e) 1.imi t switches - 1.imit dwitches are provided to control the horizontal movement of the trolley and of the bridge. The switches are located so that movement of the loaded crane over the spent fuel pool is prevented. Limit switches are also provided to indicate overloads and critical elevations of the block assembly.
f) Pendant control - For additional operator control during handling and movement of the spent fuel pool cask, pendant control (operated from the refueling floor, Elevation 117.3 f t) is provided. The pendant controls include all functions which have been incorporated in the cab controls.
9.1.4.2.3.2 Safety Analysis It is extremely improbable that th3 spent fuel cask could be inadvertently or otherwise because: dropped during the process of transferring spent fuel for shipmen t
- g MA1m
~ -
- GM W h
a) Redundancy is employed in all vital portions of the cask hoisting l "
mech,anism.
pcg l
b) Conservative design margins have been used for the cask's related handling equipment (crane, rigging, hooks, etc.).
c) Periodic nondestructive equipment test and inspection procedures are practiced.
d) Qualified operators are used and operating and administrative procedures are enforced.
The crane has a Seismic I design. Provisions have also been provided to prevent the bridge and trolley f rom jumping their respective track during a seismic event by means of an upkick lug which provides a positive mechanical engagement of the crane to the rail (Figure 9.1.4-3). The effect of a seismically induced maximum lif t swinging load was required by the performance specification. No provision was specified, however, for the consideration of an operational induced pendulum and swinging load effeet in the design of the trolley.
The Whiting Corporation has calculated, with respect to their Farley installation, that forces induced on the bridge and trolley due to pendular motion transmitted through hoist cables account for a very insignificant portion of the stresses produced in the crane structure during a seismic event. United Engineers and Constructors, Inc. , by independent calculation concurs with this assessment in that the calculated f requency of pendulum induced forces is approximately 0.024 Ha. This is well below the minimum crane lateral f requency (1.5 Hz). Based on the above, the crane does comply with the intent of the requirement.
The crane has a capacity of 125 tons with a 5 ton auxiliary hoist and serves the Reactor Building in addition to moving the spent fuel cask. All parts of the crane are designed to resist dead, live, wind, and seismic loads; overloads; and the forces produced by impact, thrust, and the rated breakdown torque of motors. The maximum hoist speed is three fpm.
9.1.4-4 j
ESR 97 00362 Page 17 Rev.0 BSEP I A2 UPDATED l'SAR Cil ANGES Replace mark up #1 of Section 9.l.4.2.2 with the following:
Afler inspection and removal of any crash structure and tie downs. the cask is raised to a vertical position using the prhnary yoke (non redundant yoke), in the event the cesk must be left standing for any extended perk >d. tie down points and cables are provided to prevent possible tipping duririg a seismic event.
The primary yoke is used to lift the cask from the tilting cradle and place it in the secondary yoke which is also on the rail car. Once the secondary yoke is engaged, the lifting device has redundant lifting capacity (see section 9.l.4.2.3). The cask is then raised through the equiprnent he'ch to the refueling floor.
Replace mark up #2 of Section 9.I.4.2.2 with the following:
The redundant yoke is again used to return the cask to the uansport vehicle. Steps involved in the loading process are reversed for the unloading process.
Replace mark up #3 of Section 9.l.4.2.3.2 with the following:
a) Redundancy is employed in all vital portions of the cask hoisting mechanism except during lifts required for installation and removal of the redundant yoke
E $R 'M co%2-ATTACHMENT B F'V o Guideline for 10 CFR 50.59 Safety Evaluations FAM M ATTACHMENT 1 <
Page 1 of 5 l 10 CFR 50.69 Safeiy Evaluation Screen Page 1 of ,,,,,,
ACTWi100 ESA 974D88L IP 300 Cent Desten, Teetna. And '--_ .. . -
4Ev. p 1.
DOES THE ACTMTV REQUIRE A CHANGE TO THE OPERATWO L8 CENSE OR TECHe# CAL SPECIPICATKM7[ ] YES (slNO Seele: No thenpos to the opereeng Neenes er technical oposincauene are needed to support ,-.A 2 of Nde acWvity.
Note: N Yes, and the seepe g( the eoWytty is tunited to e Technical Specinsettenfberating Lisono6ng change, then tempeste Secuen M et thee form, and per plant procedure. N the ecolo of the nesvar le het IWalled to e Tech. Soon er OL change. in edenen to prosessing a T . Spec. er OL chenpo roguest, conunue the screentas presses. N No, conunue the corooning procese
- 3. IS THE ACTMTV M SOUNDED SY A PREVIOUSLY PERPORIND 10 CPR f8 80 SAFETV EVALUAT:088Y []YES l e) NO Evaluellen No, Note: N Yes, stech a copy or provide document number fur roertevel cepehtetty of the pnytevoey portonned it CPR M.90 Safety Evetustion and composte Boston 8 et thee form. N No, consnue the screening process
- 3. DOES THE ACTMTV IAAME CHANGES To THE PACILITY AS DESCRWED M THE SART [jYSS [a]NO Seeee: The activity dose not phyescelly eller any componente denenbod in the F&AR List SAR llome/Secueno toviewed: 9 i
- 4. DOES THE ACTMTV thAME CHANGES TO PROCEDURES AS DESCMSED 'N THE SART [a ] YES (JNO f Seeee: A taough the F GAR conecoy nosentes the cook handhng actweten the beste for acceptatWhty g,ven in the NRC e sekty onelysee se the rodurvient nature of the hfhng yoke Since we do not rneintain reduno.incy el ea times, we have enered key eseumphone the NRC made in issuing the SER Grunswick cask handhng procedures are in confhet with the MRCe besse for mou6ng the SER List SAR llome/Secuena reviewed: 9.1
- s DOES THE ACTMTV 18fVOLVE A TEST Oft EXPERatAENT NOT DESCRSED M THE SART l]YES [ n) NO teone: Spent het cask hendung to deserthed h the SAIL No leets or esportments se alenned by 10CPnse.t* ere involved, i
List SAR teome4eettene reviewed: 9.1 NOTE: W any guessen 3 through 6 le enewsrod YES, then mort Secuen el Not AppNcebte (WA) and templete Unrovtewed telety Queeson en - r otherweee cometete Section es. ,
S. DECIPLINE PRMT NASAE WONATURE ist OSR: WA f i Date:
Other OSR: # i Date:
Other OSR: I- i Date: i and OSR: WA / i Date: ,
Ansch ademones ohosee N esaded.
.10 CFR 50,59 Rev.5 Page 13 of 37 ,
OAl 109 - Rev.9 Page 1 of 8
cst"M-oo367.
red V o ATTACHMENT B Guideline for 10 CFR 50.59 Safety Evaluations ATTACHMENT 1 Page 2 of 5 10 CFR 50.59 Safety Evaluation Unreviewed Safety Question Determination Form Page _ of _
- 1. May The ProhM;ny Of Occurrence Of An Accident Previously Evaluated in The SAR Be increased?
Basis: A spent fuel shipping cask drop accident is not evaluated in the FSAR.
- 2. May The Consequences Of An *ccident Previously Evaluated in The LAh Be increased?
Baelst An accident invoMng a spent fuel shipping cask drop is not evaluated - ]YES in t e FSAR. [x ] NO
- 3. May The Possibility Of An Accident Of A Different Type Than Any Previously Evaluated in The SAR Be Created?
[] YES [x] NO Basis: As determined by ESR 97 00362, a cask drop accident is not credible using the existing procedures for spent fuel shipping cask handling. Although redundancy is not maintained during a '
short time pen'od during transfer of the cask from the tilting cradle to the secondary yoke, the design of the primary yoke, previous load tests, and a thorough inspecton program maintain the probability of an accident extremely small, as was the case with the redundant yoke.
- 4. May The Pro;;eul;ti Of Occurrence Of A Malfunction Of Equipment important To Refety Previously Evaluated in The SAR Be incrossed?
x]YES NO Basis: Failure of rigging of equipment used for the spent fuel shipping cask was n[ot conskie[r credible since it was to be redundant at all times. Although a cask drop accident is stin ist considered to be a credible accident, since the transfer between the tilting cradle and secondary yoke involves a lift with a single lifting device, there is an increased probability of occurrence of a malfunction of equipment important to safety.
10 CFR 50.59 Rev. 5 Page 14 of 37
- 0Al-109 Rev 9 Page 2 of 8 eV h "N
r ATTACHMENT B e,ge 9 i.co3c,2 l Guideline for 10 CFR 50.59 Safety Evaluations gg,y o ATTACHMENT 1 *#
Page 3 of 5 10 CFR 60.59 Safety Evaluation I Unreviewed Safety Question Determination Form Page,_,,of _
8.
May The Consequences of A Malfunction Of Equipment important To Safety Previously Evaluated in The SAR Be Increased?
[ ] YES [x No Basis: Since a cask load drop was originally not censidered to be credible, no consequences)of a malfunction of equipment important to safety were previously evaluated in the SAR. Since a load i drop is still not considered credible, the consequences of a malfunction of equipment important to !
safety previously evaluated in the SAR has not been increased.
- 4. May The Possibility Of A Malfunction Of Lquipment important To Safety Of A Different Type Than Any Previoush Evaluated in The SAR Be Created?
[ ] YES [x)NO Basis: Since a cask drop is still not considered a credible event, the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR has not been created.
~
- 7. is The Margin Of Safety As Defined in The Bases Of Any Technical Specification Reduced? Note: The basis may be discussed in the SAR.
[ ] YES [x)NO Basis: The margin of safety for the spent fuel cask lifting yoke is not described in the bases of any technical specifications, therefore the margin of safety has not been reduced.
10 CFR 50.59 Rev. 5 Page 15 of 37 OAl-109 Rev. 9 Page 3 of 8
Es2 9;-03 6 Rcv. 0 ATTACHMENT B pgj d Guldeline for 10 CFR 50.59 Safety Evaluations ATTACHMENT 1 Page 4 of 5 10 CFR 50.59 Safety Evaluation Unreviewed Safety Question Determination Form Page of l l
If the answer to any of the questions in this evaluation is yes, then a potential Unteviewed Safety Question exists for the activity as proposed. Mark question 8 yes and sign this form.
8, 18 THERE A POTENTIAL USQ INVOLVED 7 [x ] YE8* []NO l
'if YES, PNSC review is required.
REFERENCES:
FSAR 9.1, Ch.15, SER 9.1.2,9.1.4,9.1.5 TER C5506-340/341, l
NEDO 10084 4 (VECTRA CSAR), TS 3 / 4.11.2.1 d REVIEWERS: Discipline Print Name s6gnature 1st OSR: C.It/lLlrTR /8 vn) fr mi/!nu d 'Z A M /d % )oe Date: 0~7MN1 Other QSR: / / Date:
Other QSR: / / Date:
2ND QSR: 9TRVCTVR A L /OLEMhrt s RATENDRA / M). 21 Date: [// fl 9't.,,,
Attach additional sheet for other QSRs if needed.
r 10 CFR 50.59 Rev.5 Page 16 of 37 OAl-109 Rev.9 Page 4 of 8 Y
8, aasc.2.
Rev. o ATTACHMENT B Nr 22 Guideline for 10 CFR 50.59 Safety Evaluations ATTACHMENT 1 Page 5 of 5 10 CFR 50.59 Safety Evaluation PNSC Unreviewed Safety Question Determination Form Page _ of _
l Activity identification CAPS 97-01281 Al #1 Rev 0 l DETERMINATION:
Soent fuel shinoina activities conducted at the Brunswick Plant included a condition that was unanalyzed and an unreviewed safety auestion exists.
BASIS:
Brunswick Plant performs a non sinalgt-failure oroof lift of the spent f.pel shloolna cask on the 20' elevation of the Reactor Buildina. Althouah the non redundant liftina device has had a 200% load test and is sublected to yearly insoections per ANSI N14.6. the probability of a failure is considered to be hiaher than if a redundant voke were used.
' For this reason. an unreviewed safety auestion exists.
ACTION TAKEN:
Site procedures for soent fuel shiocina. handlina. and receivina have been olaced on administrative hold.
n i
PNSC Chairman
~
v \]
Date: I[
10 CFR 50.59 Rev.5 Page 17 of 37 OAl-109 Rev.9 Page 5 of 8
' l csra. 9 7.o0362 ATTACHMENT C #
Page 1 of 2 FA6c 23 Item Classification Form DOCUMENT NO. ESR 97-00362 REV.NO. 0 Y.E.6 MQ
- 1. Does this item (including temporary changes to the facility or [x ) []
3rocedures as described in the UFSAR) require a revision to the JFSAR? (See Section 5.0 of 0Al 109 for instruction).
- 2. Does this item involve a change to the Off Site Dose Calculation Manual? (if yes, see Section 5.0 of 0Al 109 for instruction.) [] [x ]
- 3. Does this item constitute a change to the Process Control Program? (if yes, see Section 5.0 of 0Al-109 forinstruction.) [] [x]
- 4. Does this item involve a major change to a Radwaste Treatment System? (if yes, see Section 5.0 of Wil-109 for instruction.) [] [x]
- 5. Does this item involve a change to the Technical Specification Equi 3 ment List? If yes, process the change in accordance with
[] [x]
ORC -02.6. For fu(rther instruction see Section 5.0 of 0Al 109.)
- 6. Does this item impact the NPDES Permit or a "significant environmentalimpact"? Use the guidance contained in Attachment A to make th(is determination, if yes, see Section []
5.0[x]
of 0Al.109 for instruction.)
- 7. Does this item involve a change to a previously accepted:
a.
b.
Quality Assurance Program Secunty Plan including Training, Qualification, and
() [x]
Contingenc lans x
- c. Emergency lan? )? lx:
(If yes see Section 5.0 of 0Al-109 for instruct;on and complete a 10 CFh 50.54 evaluation using Attachment D.)
- 8. Does the item involve a change to the following Fire Protection Program Documents: OPLP-01; OPLP-01.1; [} [x]
OPLP-01.2; or DPLP-01.57 (If yes, see Section 5.0 of 0Al-109 for instruction.)
i 0M 109 Rev 9 Page 6 of 8
1
[">R.97 c0%2 MGV. o I ATTACHMENT C FA c c - 2M Page 2 of 2 i item Classification Form DOCUMENT NO. ESR 97_QO362 REV. NO. O f 1
l Y_lui NQ
! 1. Does this change require a revision to the improved Technical [ ]-
ification, includin associated bases or the Technical [x]
S R uirements Manua
-(If yes, see Section 5.0 of 0Al-109 for instruction.)
References:
Identify specific references used for any "Yes" answer.
_ OFSAR 9 l. 4 OAl-109 Rev. 9 Page 7 of 8
e: 5/t 9 % cc3c2 N V. O PAcc 2C ATTACHMENT 2 Sheet 1 of 1 Record of Lead Review Design ESR 97 00362 Revision 0 The signature below of the Lead Reviewer records that:
the review indicated below has been performed by the Lead Reviewer; appropriate reviews were performed and errors / deficiencies (for all reviews performed) have been resolved and these records are included in the design package; the review was performed in accordance with EGR NGGC 0003.
E Design Verification Review O Engineering Review O Owner Review e Design Review a Alternato Calculation u Qualification Testing O Special Engineering Review O YES /A Other Records are attached.
.D.d hws/YA' Lead Reviewer (print / sign) 6 vu.
Discipline
_/Askr Date item No. Deficiency Resolution /Date Revise the " Response" section to 1
provide a summary of the results of the
/dve# ' '!##
evaluation.
Revise the " Design inputs" section to 2 provide more clearly the design requirements baing met. (N(d.
2/ujTE/
Revise the " Evaluation" section to 3 clearly show how the design requirements are being met.
[gretto. 2[u/py Provide a basis for the 1/16" crack 4 assumption, M u ##V'4#d- fj FORM EGH NGGC oooJ 2 o
tTr 97-oc362 Ay w*uco7.<g I
V 5C I i
Attachment A Material Fracture Toughness input 1
, 07/16/97 06:54 C919 546 6005 CPkt-FPDD/F0D fd 002 m w,. ara
^ risw.vewr ll EVALUATION OF THE INTEGRITY OF A CASK LIFTING YOKE IN THE ecv v PRESENCE OF AN UNDETECTED CRACK N2
Purpose:
Calculate the value of the stress intensity factor for aN undetected crack in the lifting yoke and compare to the critical stress intensity factor for the material ofinterest.
References:
- 1. Figure V-76, Cask Lifting Yoke.
- 2. Harsom, John and Stanley Rolfe, Fracture and Fatigue Control in Structures. Second Edition, Prentico liall, Inc., Englewood Cliffs, New Jersey,07632.
- 3. ASTM Specification, A514 82a, Standard Speelfication forliigh Yield-Strength, Quenched and Tempered Alloy Steel Plate, Suitable for Welding.
- 4. FAX from Keith Henshaw on July 9,1997 to Ocorge Montgomery.
Calculations:
A drawing of the lifting yoke is drown in Reference 1.
From Reference 2 the formula for the stress intensity factor in a single edge crack model is l
K, = 1.12o Jn a 1
where cr is the value of the stress and u is the crack depth. The largest undetected crack from a dye penetrant test is assumed to be 1/16 inch and fmm Reference 4 the maximum I stress in the yoke i KSt. Therefore, the stress intensity factor is calculated as-2& MU j KS!4 inch. 3D g -7/Ik The fracture toughness of ASTM A514 was not available in Reference 3 but the fracture tougimess of ASTM A517 is available on page 93 of Reference 2 and is given as 150-177 KSl4 inch. Since ASTM A514 and ASTM A517 are similar in composition and mechanical propertics, the fracture toughness of A514 will be taken as that of A517.
Since the stress intensity factor of the 1/16 inch crack is less than the critical stress intensity factor, an undetected crack of 1/16 inch will not grow in the yoke in an unstable manner, i
Prepared by:
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- a a' Nuclear Enginernes O(ptttment N 'f7 003(. T.
Brunsw,ick Engine: ring Support Section A7m(NA<u7A Fax Sheet PO 80s10429 TEM d Southport, NC 284614429 PA6r 3 fas 910/457 3444 To d e b[ - -
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A ETC 9'7 cawy y ^Qg%on X"
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v1 v toe r<a A Component Safety Factor OW) Safety Factor (W Cross Member 2.64 7 92 Arm 6.84 . ' U. , .
Yoke Pin 2.07 6.2 i Worst Case Welds 2.06 6.18 h ,,g, Io '2 l 1.41 I
a 4
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, P%:4 G ENERAL ELECittC COMPAM
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PROPRETARY NORMATEN c 74.5 in. - -->
> =* - 0.75in.TYP -
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t 38.5 in, : 36.5 in. y e - 13 an. .e-ALL MATERIAL TO DE ASTM AS14. TYPE E OR TYPE P ey (MIN)
- 58 hsi
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FIGURE V.76. CASK LIFTING YOKE
ns 9, . o u,.
+rmcment h" K4"V. O
~ #
Cp&L To: Keith lienshaw From: Ed Black Date: July 16,1997
Subject:
NDE I imitations for Lifting Devices In response to your question "Do existing site NDE Procedures e.sure detection of cracks 1/16" in length and greater?" A conservative dimension of 1/16 inch in length should be ,
used. I arrived at this dimension using the following facts / assumptions:
l The NDE procedures we use for liquid penetrant (PT) and magnetic particle (MT) inspections of lifting devices have been demonstratcJ to find flaws that are smaller than 1/16 inch in length or diameter. This fact is based on experience as well as review of previous reports, (copies of reports can be provided if needed). Ilowever, due to the difficulty experienced with characterizing a 1/16 inch indication as a crack versus a pore, the 1/16 inch dimension should be used.
The reason for the 1/16 (0.0625) inch threshold is the difficulty experienced with characterizing rounded versus linear indications (porosity versus crack). Using the definition of a linear indication where the length must exceed the width by a factor of 3, one would have to accurately measure an indication width of 0.020 inch in the field to determine the 3 to I ratio. This is difficult to perform with a six inch scale graduated in 1/64 (0.015) inch increments. It can and has been done, but it is difficult.
Most Code acceptance criteria has recognized this difficulty by treating indications with dimensions of 1/16 inch or less as non-relevant (no requirement to evaluate).
In conclusion, the recommended 1/16 inch threshold of crack detection is driven by the difficulty experienced with characterizing 1/16 inch indications, not by the process used. This limitation is imposed by the tools used in the field, not by the physical ability of the inspector.
CP&L NDE Inspectors are re-qualified on a tri annual basis to ensure proficiency and must undergo an annual test to ensure vision acuity.
If you have further questions, call me at 3639, d7sb Ed Ifiack, CP&L NDE Level III
)0.,30-1981 9,3GPM FRCH 65* p.1 \
se rg i. , i nw c.
ow \
NDEP. 01, Rev. 20 l LIQUID PENETRANT EXAMINATION (VISIBLE DYE, SOLVENT REMOVAB ATTACHMENT B LIOUID PENETRANT EXAMINATION ASME Boller & Pressure Vessel Code, Section Ill, Division i 1974 Edition with Addenda through Winter.1976 Aniyity: Examination of welds, areas of welds prepared for repair and repaired welds.
Accent 2nce criteria: Division 1; NB 5350, NC 5350, ND 5350 NE 5350*
NF 5350, NO 5350.
a.
Unless otherwise specified, the following relevam indications are unacceptable:
- 1. Any cracks or linear indications
- 2. Rounded indications with dimensions greater than 3/16 inch:
3.
Four or more rounded Indications in a line separated by 1/16 inch or less, edge-to edge; 4.
Ten or more rounded indications in any 6 square inches of surface with the major dimension of this area not to exceed 6 inches with the area taken in the most w g unfavorable location relative to the indications being evaluated.
- b. Indications with major dimensions greater than 1/16 inch shall be ccasidered televant.
Holc: Llacar indications are those indications in which the length is more than three times the width.
Rounded indications are circular or elliptical with lergth less than three times the width.
Relevant indications are those which result from mechanical discontinuities.
- For examination to Article NE 5350, (b) above does not apply.
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Page 22 of 47 or o<*ek
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10-30-19a1 9:37PH FRCri p, p 3 3,w w *
- m e NDEP 301, Rev.12 MAGNETIC PARTICLE EXAMINATION (DRY POWDER, PRODS AND YOKE)
O A*ITACIIMENT A MAGNETIC PARTICT F EXAMINATION ASME Boller & Pressure Vessel Code,Section III, Division 1 1974 Editbn with All Addenda 1977 Edition with'All Addands 19g0 Edition with All Addenda 1983 Edition with All Addenda 1986 Edition with No Addenda Activirv: Examination of pressure retaining material and product fonn (i.e. " base material).
Accentance criteria: NB 2545, NC-2545, ND 2545, NE 2545 1
l
- a. Only indications with major dimensions greater than 1/16 in. shall be considered relevant.
- b. 'Ihe following relevant indications are unacceptable:
I
- 1. Any linear indications greater than 1/16 in. long for materials less than 5/8 in, thick, greater than 1/8 in. long for materials from 5/8 in. thick to under 2 in, thick and 3/16 in. long for materials 2 in thick and greater;
- 2.
- Rounded indications with dimensions greater than 1/8 in. for thickness less than 5/8 in, and greater than 3/16 in. for thicknesses 5/8 in. and greater;
- 3. Four or more indications in a line separated by 1/16 in. or less edge to edge:
- 4. Ten or more indications in any 6 square b:hes of trea whose major dimension is no more than 6 inches with the dimensions taken in the most unfavorable location relative to the indir9 ions being evaluated.
Natt: Relevant indications are those which result from unacceptable mechanten) discontiDuilles. Linear indications are those indications in which the length is more than three times the width. Rounded indicetions are circular or elliptical with length less than three. times the width.
O Page 18 of 37
10 2
,.,7-1981 4:40Pti FRCri p, j
- cu ,,--ue A%c4,vgyy <Ar nv. o NDEP 301, Rev.12 NF9 MAGNETIC PARTICLE EXAMINATION (DRY POWDER, PRODS AND YOKE)
P .
ATT4CHMENT I MAGNETIC PARTICLE EXAMINATION ANSI B30.2.0 Overhead and Gantry Cranes l
Activitv: Magnetic particle examination of crane hooks.
^~~~* critaria: Hooks with any indkations detennined to be cracks shall be rejected.
F e
Page 31 of 37
Y[2595aE e,r er.-,W d;rygeur w" NDEP 201, Ov3d LIQUID PENETRANT EXAMINATION (VISIBLE DYE, SOLVENT REMOVABLE)
ATTACllMENT J &
LIOUID PENETRANT EXAMINATION ASME/ ANSI B30.2 " Overhead and Gantry Cranes
- i including ASME/ ANSI B30.2a 1991 Addenda Anivirv: Liquid Penetrant Examination of Overhead and Gantry Crane Hooks I Are***ne, Critstia: Hooks with any liquid penetrant Indications determined to be cracks shall be l
rejected.
LIOUID PENETRANT EXAMINATION ASME/ ANSI B30.10 - 1987 with Addenda ASME/ ANSI B30.10.c 1992 liooks m
Activirv:
Liquid penetrant exammatio.- 'Ihooks (except noted above in " Overhead and Gantry Cranes")
Aecap**=ce criteria:
Hooks with any liquid penetrant indications determined to be cracks, nicks or gouges shall be rejected.
m Page 30 of 47
, o
- NAS INDEPENDENT SAFETY REVIEW ROUTING TRAVELER Page 1 TRAVELER #: 197073 FILE NUMBER: 7105 DOCUMENT DATE: 7/22/97 DOCUMENT TYPE: ESR DOCUMENT NUMBER: 9700036200 i
SUBJECT:
IF 300 Casks Design, testing and inspection.
REVIEW DISCIPLINFS:
DISCIPLINE NAME QSR INITIAL DATE Nuclear Power Plant Operations C. L. Schacher 4Y 7/22/97 Nuclear Engineering C. L. Schacher Yes M 7/22/97 Chemistry and Radiochemistry J. D. Ferguson Yes /[
Radiological Safety J.D.Ferguson Yes Mechanical Engineering C. L. Schacher Yes 7/22/97 Seismic C. L. Schacher Yes 7/22/97 ISRE ADMINISTRATIVE REVIEW: c,1, n.4.J.,. M / 7/22/97 INITIAL DATE SPECIAL REVIEW INSTRUCTIONS:
Independent Safety Review (ISR) screening questions:
- 1. Does the 10CFR50.59 address changes to the facility which affect the design, function or method of performing the function of a Structure, System or Component (SSC) described in the SAFETY ANALYSIS REPORT (SAR) either by text, drawing or other information relied upon by the NRC7 Yes[] No [/ ]
- 2. Does the 10CFR50.59 address changes to (1) procedures that are outlined, summarized or completely, described in the SAR or (2) anything described in the SAR that defines or describes activities or controls over functions, plant configuration, tasks, reviews, or tests?
Yes [ ] No [./)
NAS INDEPENDENT SAFETY REVIEW ROUTING TRAVELER Page 2 ,
- 3. Does the 10CFR50.59 create a test or experiment which could degrade the margins of safety during normal operations or anticipated transients or degrade the adequacy of stmetures, sys-
- tens or components to prevent accidents or mitigate accident conditions describeci in the SAR?
j Yes() No [/)
l 4. Does the 10Cl R50.59 identify a change to the Technical Specifications, operating license, or l an Unreviewed Safety Question (USQ)? Yes [/ ) No ( )
. NOIE: If the answer to all of the above four questions is "No", an ISR is not required. Skip the remaining questions and document a basis for not conducting an ISR in the Review ,
Comments, if the answer to question 4 is "Yes", complete the ISR prior to implementation.
r
- 5. Does the 10CFR50.59 Safety Evaluation adequately address all potential changes to the ,
facility, identified by the ISR reviewer (s), which could affect the design, function or method of
- relied upon by the NRC)? In answering this question *No" the ISR reviewer (s) have dete mined that the proposed change, without additional evaluation, could adversely affect the ability of the facility to operate without undue risk to the public health and safety or adversely affect nuclear safety. Yes [/] No [ ]
- 6. Does the 10CFR50.59 Safety Evaluation adequately address all potential changes, identified by the ISR reviewer (s), which could affect (1) procedures that are outlined, summarized or
- describes activities or controls over functions, plant configuration, tasks, reviews, or tests? In i answering this question "No" the ISR reviewer (s) have determined that the proposed change, j without additional evaluation, could adversely . affect the ability of the facility to operate i without undue risk to the public health and safety or adversely affect nuclear safety. ,
Yes [/] No [ ]
4 7. Does the 10CFR50.59 Safety Evaluation adequately address all potential changes, identified by the ISR reviewer (s), which could degrade the margins of safety during normal operations or anticipated transients or degrade the adequacy of structures, systems or components to prevent accidents or mitigate accident conditions described in the SAR7 In answering this question "No" the ISR reviewer (s) have . determined that the proposed change, without additional
< evaluation, could adversely affect the ability of the facility to operate without undue risk to the public health and safety or adversely affect nuclear safety. Yes [/] No [ ]
- 8. Does t'ie 10CFR50.59 Safety Evaluation provide an adequate basis to independently conclude the potential changes do not constitute an Unreviewed Safety Question (USQ)?
- Yes [/] No [ ]
1 l NOTE: If the answer to any of the above four questions (5 through 8) is "No", document a basis in the Review Comments. Notify NAS management and individuals involved with the change so that resolution can be c'ompleted prior to implementation.
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NAS INDEPENDENT SAFETY REVIEW ROUTING TRAVELER PageJ REVIEW COMMENTS: See ESR 97 0036200.
Note: This NAS Independent Safety Review (ISR) Routing Traveler la a QA record and will be filed separately by the Nuclear Assessenent Section (NAS).
The 10CFR50.59 Safety Evaluation for the proposed changes is adequate.
_ There is an Unreviewed Safety Question which has been addressed. The proposed changes reference the ;
applicable Codes, Starulards, and Regulations. There are no deleterious effects identified.
There is a potential for sorne increased risk, but it is non<redible. There are no nuclear safety 3
concerns created by this evaluation. No further ISR action is reconunended. '
FOLLOW UP ITEM h None t
i t
1 NAS INDEPENDENT SAFETY REVIEW ROUTING TP.AVELER PageI l
TRAVELER # 197073 FILE NUMBER: 7105 DOCUMENT DATE: 7/22/97 DOCUMENT TYPE: ESR DOCUMENT NUMBER: 9700036200 SUIMECT: IF 300 Casks Des!gn, testing and inspection.
REVIEW DISCIPLINES:
DISCIPLINE NAME QSR INITIAL DATE Nuclear Power Plant Operations C. L. Schacher M 7/22/97 Nuclear Engineering C. L. Schacher Yes M 7/22/97 7/22/97 Chemistry and Radiochemistry J. D. Ferguson _
Yes /[
Radiological Safety J. B. Ferguson Yes Mechanical Engineering C. L. Schacher Yes 7/22/97 Seismic C. L. Schacher Yes W 7/22/97 ISRE ADMINISTRATIVE REVIEW: c.1, SJ.J .,. M 7/22/97 INITIAL DATE SPECIAL REVIEW INSTRUCTIONS:
Independent Safety Review (ISR) screening questions:
- 1. Does the 10CFR50.59 address changes to the facility which affect the design, function or method of performing the function of a Structure, System or Component (SSC) described in the SAFETY ANALYSIS REPORT (SAR) cither by text, drawing or other information relied upon by the NRC7 Yes [] No [/ )
- 2. Does the 10CFR50.59 address changes to (1) procedures that are outlined, summarized or completely, described in the SAR or (2) anything described in the SAR that defines or describes activities or controls over functions, plant configuration, tasks, reviews, or tests?
Yes [ ] No (/]
NAS INDEPENDENT SAFETY RlWIEW ROUTING TRAVELER Page 2
- 3. Does the 10CFR50.59 create a test or experiment which could degrade the margins of safety during normal operations or anticipated transients or degrade the adequacy of structures, sys.
tems or components to prevent accidents or rnitigate accident conditions described in the SAR?
Yes[] No (/]
- 4. Does the 10CFR50.59 identify a change to the Technical Specifications, operating license, or an Unreviewed Safety Question (USQ)? Yes [/ ] No ()
NOTE: If the answer to all of the above four questions is "No", an ISR is not required. Skip the remaining questions and document a basis for not conducting an ISR in the Review Comments. If the answer to question 4 is "Yes", complete the ISR prior to implementation.
l 5. Does the 10CFR$0.59 Safety Evaluation adequately address all potential changes to the l
facility, identified by the ISR reviewer (s), which could affect the design, function or method of performing the function of a Structure, System or Component (SSC) described in the SAR (and relied upon by the NRC)? In answering this question "No" the ISR reviewer (s) have determined that the proposed change, without additional evaluation, could adversely affect the ability of the facility to operate without undue risk to the public health and safety or adversely affect nuclear safety. Yes [/] No [ ]
- 6. Does the 10CFR50.59 Safety Evaluation adequately address all potential changes, identified by the ISR reviewer (s), which could affect (1) procedures that are outlined, summarized or completely, described in the SAR or (2) anything described in the SAR that defines or describes activities or controls over functions, plant configuration, tasks, reviews, or tests? In answering this question "No" the ISR reviewer (s) have determined that the proposed change, without additional evaluation, could adversely affect the ability of the facility to operate without undue risk to the public health and safety or adversely affect nuclear safety. ,
Yes[/] No [ ]
- 7. Does the 10CFR50.59 Safety Evaluation adequately address all potential changes, identified by the ISR reviewer (s), which could degrade the margins of safety during normal operations or anticipated transients or degrade the adequacy of structures, systems or components to prevent accidents or mitigate accident conditions described in the SAR? In answering this question "No" the ISR reviewer (s) have determined that the proposed change, without additional evaluation, could adversely affect the ability of the facility to operate without undue risk to the public health and safety or adversely affect nuclear safety. Yes [/] No [ ]
- 8. Does the 10CFR50.59 Safety Evaluation provide an adequate basis to independently conclude the potential changes do not constitute an Unreviewed Safety Question (USQ)?
Yes [/] No [ ]
NOTE: If the answer to any of the above four questions (5 through 8) is "No', document a basis in the Review Comments. Notify NAS management and individuals involved with the change so that resolution can be c'ompleted prior to implementation.
. __ J
..r NAS INDEPENDENT SAFETY REVIEW ROLTI'ING TRAVELER Page 3 REVIEW COMMENTS: See ESR 97 0036200.
Note: This NAS Independent Safety Review (ISR) Routing Traveler is a QA record and will be filed separately by the Nuclear Assessment Section (NAS).
The 10CFR50.59 Safety Evaluation for the proposed changes is adequate. There is an Unreviewed Safety Question which has been addressed. The proposed changes reference the applicable Codes, Standards, and Regulations. There are no deleterious effects identified.
There is a potential for some increased risk, but it is non-credit'e. There are no nuclear safety concerns created by this evaluation. No further ISR action is tes ,rnmended.
FOLLOW-UP ITEM #: None l
.