ML20202F635
| ML20202F635 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 12/02/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20202F628 | List: |
| References | |
| NUDOCS 9712090152 | |
| Download: ML20202F635 (5) | |
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NUCLEAR REGULATORY COMMISSION k..... /g WASHINGTON. D.C. 30656 0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.190 TO FACILITY OPERATING LICENSE NO. DPR-71 AND AMEN 0 MENT NO.221 TO FACILITY OPERATING LICENSE NO. DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT. UNITS 1 AND 2 DOCKET NOS. 50 326 AND 50-324
1.0 INTRODUCTION
By letter dated August 6. 1997. Carolina Power and Light Company (CP&L)(the licensee), submitted a license amendment request pursuant to 10 CFR 50.90 for approval of changes to the U) dated Final Safety Analysis Report (UFSAR).
The proposed changes relate to tie cask handling operations involving lifting Vectra IF-300 spent fuel shipping casks at the Brunswick Steam Electric Plant (BSEP). Units 1 and 2.
The requested UFSAR changes are due to a discrepancy between the licensing basis and site procedures regarding the handling and control of casks. The site procedures require casks to be lifted over a short section of the load path without the single failure proof lifting system and with the cask valve Jox covers removed.
Past correspondence to the NRC indicates that the same lift is made with single failure proof capability and with the valve box covers installed.
Through Licensee Event Report Number (LER)97-004, dated June 5. 1997. CP&L informed the NRC of the discrepancy between the site procedures and the UFSAR regarding the movement of a spent fuel shipping cask.
In the LER. CP&L indicated that the cask handling operations were performed under conditions that are outside the design basis and involve an unreviewed safety question (US0).
In the August 6. 1997, submittal. CP&L provided a detailed analysis pursuant to 10 CFR 50.59 regarding a postulated cask drop accident to support its conclusion that the cask handling operation does involve a US0 but is not outside the design basis of the plant.
2.0 BACKGROUND
In a letter dated November 16. 1982. CP&L informed the NRC that BSEP's reactor building crane design satisfied the requirements of NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants."
In that letter and other correspondence provided to the NRC. dated June 18, 1976. and June 22. 1981. CP&L indicated that BSEP uses a lifting device consisting of a yoke (primary yoke) and a lifting basket (a secondary yoke) which are of " redundant" design to lift spent-fuel shipping casks. CP&L also indicated that these components are used in conjunction with a single failure proof crane during transfer of the casks.
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l 2-Based on these capabilities, CP&L concluded that-a load drop of a cask is not credible.
18, 1984, NRC Technical Evaluation Report / Safety Evaluation, dated May approved CP&L's load handling operations, and acknowledged that BSEP's s)e:ial
-lifting devices satisfy the single failure proof guidelines as provided ay~
NUREG 0612 and the requirements of ANSI N14.6 1978, " Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds or More for that a load drop of the cask is not credible, pted the licensee's positia Nuclear Materials." Therefore, the staff acce r
NRC Bulletin (NRCB) 96-02, dated April 11, 1996, requests that all licensees e
review their plans and capabilities for handling heavy loads while the reactor is at power in accordance with existing regulatory guidelines.
The bulletin-also requests that licensees determine whether their current heavy loads-hanoling activities are within their current licensing basis and, if not.
submit a license amendment request.
In addition, a Request for Additional Information (RAI), dated December 6, 1996, regarding further evaluation of.
load handling activities was issued to BSEP.
This evaluation was to determine if a cask tipping-cver hazard exists while dry casks are being moved by plant cranes.
BSEP responded by letters dated May-10. 1996, and February 7, 1997, to NRCB-96-02 and the associated RAi, respectively.
In both responses, BSEP indicated that the cask drop and cask tipping over hazards are not credible at the Brunswick plant due to the redundant design of the special lifting device, the single failure proof upgrade of the crane, and BSEP's compliance with not credible. gain, the staff accepted BSEP's conclusion that a cask drop is NUREG-0612.
A BSEP's recent_ discovery that there is a discrepancy between its procedures and existing analysis is contrary to information )reviously provided to the NRC that indicated that single failure proof capa)ility is used throughout the entire cask handling operation.
3.0 EVALUATION 3.1 Analysis of Postulated Cask Drop Accident The cask lift occurs at the 20' elevation (ground level) of the Reactor Building at the lower level of the equipment hatch with a lift height of approximately 7' above the rail car. As the cask is raised to the vertical
. position from a horizontal position and lifted into the cask lifting basket, redundancy in the design of the yoke does not exist.
Therefore, the yoke is not single failure proof in accordance with Section 5.1.6 of NUREG 0612. Once the cask-is in the lifting basket, the " redundant" yoke is used to lift the i
assembly to the 117' elevation. All the components of the yoke including the cross members, arms, and J-hooks are made of ASTM A514 material exce)t for the
-yoke pin which is made of ASTM 4340 steel. The rateo load of the yoce is 140.000 lbs. and the Vectra IF-300 cask loaded with spent fuel weighs
. approximately 140.000 lbs.
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NUREG-0612 Section 5.1.6. " Single Failure Proof Handling Syste: ' provides the alternative of upgrading an existing crane in lieu of complying ith certain recommendations of NUREG-0554 " Single Failure Proof Cranes for haclear Power Plants." to achieve improved reliability in a load handling system.
- Accordingly, upgrades to the crane and associated lifting devices can be achieved through an increase in the factors of safety, and through redundancy-or duality in certain active components.-
Because ANSI N14.6. Section 6 allows increased stress factors to be used to demonstrate reliability where redundancy is not an inherent part of the design of the yoke, the licensee analyzed the failure potential of the components of the yoke in accordance with the standard. ANSI N14.6 requires that special lifting devires be designed to a safety factor of three with respect to yield strength, and five with respect to the ultimate strength of the material.
NUREG-0612. Section 5.1.6(1)(a) requires that these factors be doubled for special lifting devices that are used to handle heavy loads where an upgraded crane is involved. With respect to the yield and ultimate strengths of the components of the yoke, the following compares the licensee's findingt with regard to the nonredundant yoke to the requirements of NUREG-0612.
Yield NUREG Ul c ate NUREG strenath
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/ ANSI Cross members 7.92 6
9.24 10 Arms-20.52 23.94
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J hooks 3.63 4.24
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Yoke pin 6.21
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Worst case welds 6.18.
- The ultimate strength cannot be demonstrated or is not available.
Based on the above, the safety factors of the cross members and arms approximate or exceed that required by the standard. With regard to the J-hooks, the licensee used stress analysis to demonstrate that the J-hooks would not fail under design basis earthquake conditions.
The licensee found that the cross members, arms, and J-hooks of the non-redundant yoke have yield strengths in excess of 80% of the ultimate strength of the material. As a result, the licensee based its acceptability of these components on the material fracture toughness as is allowed by ANSI N14.6.
Section 4.2.1.1.
Based on the fracture toughness analysis of these.
components, the licensee found that unstable cracking propagation throughout the material due to maximum loading of the components will not occur. Thus, the licensee concluded that a cask drop due to a failure of the components of the yoke is not credible.
Since the safety factors regarding the ultimate strength of the yoke pin and worst case' welds could not be demonstrated, the licensee evaluated these-components based on safety margins and inspections as is permitted by NUREG-0612. Section 5.1.1(4). Based on-this evaluation. the licensee found that prior load tests (original and post modification tests) had been performed on the yoke at 200% of its rated load of 140.000 lbs (i.e., at 280.000 lbs.).
Since ANSI N14.6. Section 6.2 requires that after fabrication and prior to
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4 initial use, special lifting devices will be load-tested at 150% of the rated-load, BSEP's prior load tests on the yoke exceeded the requirements of the
- standard, in addition to the above analyses, the licensee performed a walk-down of the systems located within the load
)ath and verified that, due to the availability of redundant safe slutdown systems, shutdown of the plant can be achieved following a dropped cask.
Accordingly, the licensee concluded that the potential damage that could result from a cask drop would not preclude the operation of sufficient equipment needed to achieve safe shutdown.
Based on the analyses discussed above, the licensee reviewed its crane design, the strength of materials for the components of the lifting yoke, prior load tests, the load lifting history,luded that a cask drop is not credible. previous inspe and unloading processes and conc These findings, as supported by the licensee's analysis, are acceptable.
3.2 Changes to UFSAR Sections 9.1.4,2.2 and 9.1.4.2.3.2 The licensee proposed to change UFSAR Sections 9.1.4.2.2, " Spent Fuel Shipping Cask Handling, and 9.1.4.2.3.2, " Spent Fuel Shipping Cask Hoisting." The changes are intended to make the licensing basis consistent with the site procedures governing the l, d handling operations. The changes clearly indicate when the non-redundant and redundant yokes are used to lift the cask.
Changing the UFSAR to more accurately reflect actual plant conditions during the transfer of the cask will hel) to assure compliance with and o)eration within the plant-specific design Jasis. Therefore, the proposed c1anges to the UFSAR are acceptable to the staff.
4.0 Results Of Staff Review Based on the above discussions, the staff finds that the licensee has appropriately proposed changes to the UFSAR to make its load handling o)eration consistent with BSEP's design basis.
In support of the proposed c1anges, the licensee has evaluated the potential for a cask drop accident using the non-redundant yoke.
Through analysis, the-licensee has demonstrated that the components of the non-redundant yoke should not fail.
In addition, the licensee has reviewed the results of prior nondestructive examinations and inspections of the worst case welds and finds that no prior material defects had been observed. Also, the lift history of the system, and the use of qualified operators and adequate load handling and inspection procedures further assure the licensee that the crane and hfting device will not experience failure without prior detection. Based on these findings. and on the fact that lifting the cask using the non-redundant yoke is confined to a relatively short section of the load path, the staff concludes that failure of the yoke is not credible.
In addition, the staff concludes that BSEP has been operating outside the licensing bases for the facility because its load handling o)erations do not conform with information provided to the NRC in CP&L's Novem)er 16, 1982, letter.
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5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendments. The State official had no coments.
6.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change in the installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20.
The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a propused finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (62 FR 48897).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above.
that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposr.d manner. (2) such activities will be conducted in compliance with tt? Coonission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
Brian E. Thomas Date: December 2, 1997
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