ML20151K380
| ML20151K380 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/25/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151K367 | List: |
| References | |
| NUDOCS 8808030173 | |
| Download: ML20151K380 (4) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 113 TO FACILITY OPERATING LICENSE N0. DPR-3 YANKEE ATOMIC ELECTRIC COMPANY-YANKEE NUCLEAR PCKER STATION INTRODUCTION By a latter dated January 15, 1988, Yankee Atomic Electric Company requested a modification to the Yankee-Rowe Nuclear Power Station's Technical Specifica-tions.
By letter of March 31,198P, the licensee submitted a "Cask Drop Analysis" to suoport the above request.
The current Yankee-Rowe Technical Specification (T/S) 3/4.9.7 limits the loads traveling over the Spent Fuel Pit (SFP) to 900 pounds, except for six identified items.
The proposed modification T/S 3/4.9.7 would add four identified items to the exceptions provided for in the Limiting Condition for Operation (T/S 3.9.7), and include these items as appropriate in the Surveillance Requirements (T/S 4.9.7.1) and Bases.
By letter dated May 19, 1988, the licensee submitted Supplement No. 1 to the January 15 letter. In Supplement No.1 an additional modification was proposed to add the same four identified items to another section of the Surveillance Requirements (T/S 4.9.7.2).
The Yankee-Rowe SFP has an accumulated inventory of nonfuel components, pre-dominantly control rods and follower sections.
The storage space that these components are occupying is needed for additional spent fuel storage beginning with the 1988 refueling outage.
To remove the control rods from the SFP, the licensee has stated that it will be necessary for 35-ton shipping casks, volume reduction equipment weighing approximately 10,000 pounds, and shipping cask liners with gross weight less than 2,000 pounds to travel over the northern end of the SFP.
Also, to perform non-destructive testing of spent fuel assemblies, it will be necessary to periodically lift test equipment weighing less than 2,000 pounds into the northern end of the SFP.
Each of the above items exceed the 900-pound T/S limit.
The Surveillance Requirements of T/S 4.9.7.1 provide that the excepted loads may travel over the SFP only in accordance with approved written precedures; and that the excepted loads shall be prevented from travelina over fuel assemblies in the SFP by administrative control and, when appropriate, by the steel framing at the southern edge of the SFP superstructure roof opening.
The Surveillance Requirements of T/S.4.9.7.2 provide that the excepted loads not be permitted to be moved over the SFP unless the spent fuel in the SFP has decayed for a specified number of days.
The Bases provide that the restric-tion of movement of the excepted loads ensures that these items cannot be dropped on spent fuel; and that dropping any one of the excepted loads from its maximum height will not result in loss of integrity of the fuel pit floor.
8808030173 080725 DR ADOCK 050 9
2 E{ALUATION Standard Review Plan (SRP) Section 9.1.5, Overhead Heavv Load Handling Systens (OHLHS), provides guidance for this safety review involving loads weighing rore than one fuel assembly and its associated handling device in accordance wit 5 General Design Criteria (GDC) 2, 4, 5 and 61.
NUREG-0612, "Control of Heavy Loads at Muclear Power Plants", is referred to as providing specific criteria for meeting the relevant requirements of GDC 2, 4 and 61.
The review examined the systen design with respect to potential system malfunctions or failures that could pre-vent safe shutdown of the reactor, or cause an unacceptable release of radio-activity, a criticality accident or the inability to cool the fuel in the reactor vessel or spent fuel storage pit.
The staff revi~ewed the analyses of the poten-tial consequences that could result from dropping of a heavy load at any point along its path of travel.
In a previous staff review of the control of heavy loads for the Yankee-Rowe Nuclear Power Station (letter from J. Zwolinski, NRC, to J. Kay, Yankee Atomic Electric Company, dated February 19, 1985), the staff concluded that the guide-lines of Sections 5.1.1 and 5.3 of NUREG-0612 had been satisfied.
The safety evaluation addressed Section 5.1.1 ouidelines on safe load paths, load hardlino procedures, crane operator training, special lifting devices, lifting devices (not specially designed), cranes (inspection, testing, and maintenance), ard crane design; and Section 5.3 quidelines to be used in the interim neriod until measures at operating plants are upgraded to satisfy the guideline, of Section 5.1.
In this evaluation the staff determined that the design of the yard area crane would be consistent with the provisions of guidelines contained ir NUREG-061E when modifications to conform with the Crane Manufacturers Association of America (CMAA) have been conpleted and that upon revision of the procedure for inspection and maintenance of the yard crane the guidelines of NUREG-0F'.2 will be met.
The licensee has described the safe load path, the yard area crane (addressed in the above-cited NRC letter dated Jaunary 19, 1985) to be used for this operation, plant procedures, the cask lifting device, an.d a cask drop analysis.
Yankee Atomic stated that the design of the yard area crane complies with the guidelines of CMAA Specification No. 70, and Chapter 2-1 of ANSI B 30.2-1976.
The yard area crane has a capacity rating of 75 tons and has been used in con-juretion with 24 fuel shipments with a 75-ton spent fuel shipping cask in lifts nearly identical to those proposed for shipping control rods.
The safety factors for this 35-ton lift, considering the crane and the main boist rope of the crane, are greater than 10, which is con:istent with the defense-in-depth approach recommended in Section 5.1. of NUREG-0612.
Redundant limit switches have been installed to provide an additional margin of safety.
Plant procedures have been revised to include the recommendations of NUREG-0612, Section 5.1.1(2) and previously identified commitments, including inspections, maintenance, and opera-tional testing.
A cask lift procedure will be prepared prior to this task.
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. : Plant' procedures requiro that the crane operators be trained arf rualified in accordance with Chapter ?.3 of ANSI B 30.2-1976.
The cask lifting fixture con-i forms to the requirements of NUREG-0612, _ Section 5.1. l(4).
Yankee Atomic also performed an analysis which concludes that there is _ ample margin against pit leakage from a hypotheticel cask drop into the SFP.
4 The licensee stated that this proposed change to.T/S 4.9.7.2 is necessary-to maintain the validity of the conclusions accepted by the steff in the previous safety evaluation referred to by the licensee and described below.
The licensee's submittal referred to the staff's previcus safety evaluation (letter from G. Lear, NRR to G. Papanic, Jr., Yankee Atomic Electric Company dated May 20,1986) of the licensee's request to amend the T/S to allow move-ment of SFP Building batches over the SFP (letters dated March 28 and May 3, 1984, and-Pay 7, 1985).
In that safety evaluation the staff referred to (1)
NUREG-0612, which states that no potential for criticality appears to exist as a result of a heavy load drop in a PWR spent fuel pool containing only spent fuel; and (2) the-licensee's criticality analysis and assumptions.
The maxi-mum K was found to be 0.8932.
The staff found this assumption acceptable and t8$bcriticality will not occur in the SFP as a result of the heavy load drop.
In the previous analysis the staff requested that the licensee demonstrate that a heavy load drop on irradiated fuel in the SFP would not exceed the design basis criterion in SRP Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents".
The licensee cited two previous analyses - one assumed that 721 newly discharged assemblies in the upper tier were damaged after a 90-day decay period and the second analysis assumed damage of 76 newly discharged assemblies in the lower tier after a 45-day decay
^ period.
The calculated offsite doses were 100 rem and 75 rem, respectively.
The staff performed an analysis of the licensee's submittal and determined that the Technical Specifications proposed therein would assure that offsite doses would not exceed the dose guideline in SRP Section 15.7.4 in the event of a heavy load drop.
The staff has reviewed the information provided in the licensee's submittals and in the staff's previous evaluations and concludes, based on this information, that the reouested bodification to Technical Specification Section 3/4.9.7 of the Yankee-Rowe Technical Specifications meets.the guidance of SRP Section 9.1.5 regarding conformance to the relevant requirements of GDC 2, 4, and 61, and is censistent with the Yankee-Rowe Technical Specifications and, therefore, is acceptable.
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4 ENVIRONMENTAL CONSIDERATION A fictice of Issuance of an E'nvironr. ental Assessment and Finding of No Significant Impact relating to the proposed amendment for the Yankee Nuclecr Power Station was Published in the Federal Reaister on July 23, 1988.
CONCLUSION The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will rot be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Conmission's regulations and the issuance of this amend. rent will not be inimical to the conenon defense and security or to the health and safety o' the public.
ACXNOWLEDGEMENT Principal Contributors:
Amira Gill Charles Nichols Frank Rinaldi Morton Fairtile Dated:
July 25, 1988 m.
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