ML20236C699

Safety Evaluation Supporting Issuance of License SNM-1945. Amend 4 to Indemnity Agreement B-102 Encl

ML20236C699
Person / Time
Site:	07003013
Issue date:	07/27/1987
From:	Mccaughey D NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20236C665	List: ML20236C661 ML20236C685 ML20236C699 ML20236C728 ML20236C746
References
NUDOCS 8707300162
Download: ML20236C699 (10)

Text

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JUL 2 71987 l

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' DOCKET NO: 70-3013 APPLICANT: Commonwealth Edison Company (CECO)  ;

p FACILITY: Braidwood Station, Unit 2 L

SUBJECT:

SAFETY EVALUATION REPORT, RE REVISED LICENSE APPLICATION FOR A MATERIALS LICENSE I. INTRODUCTION On June 8,1984, CECO applied for a license to receive, possess, inspect, and store unirradiated nuclear fuel assemblies. Additional information was provided by Ceco in supplements dated September 18 and November 19, 1984. On May 6, 1987, CECO submitted a revised application but did not incorporate the September supplement.

In December 1975, the NRC imsued CPPR-133 for Braidwood Station, Unit 2, which is located in northeastern Illinois, 3 miles southwest of the Kankakee River, 20 miles south-southwest of the Town of Joliet, and 60 miles southwest of Chicago, Illinois. Braidwood-Station, Unit 2, is a pressurized water reactor and will use fuel supplied by Westinghouse Electric Corporation for the initial core loading. The materials license will allow early receipt of fuel for the purpose of inspection and preparation of the fuel for reactor loading. The license will  !

not authorize core loading. ]

1 The uranium is in the form of U-235 enriched uranium oxide ceramic pellets. A fuel rod consists of pellets encapsulated (clad) in Zircaloy-4 tubing which is seal-welded at both ends. A fuel assembly contains fuel rods in 264 positions in a 17 x 17 square array. Other positions are taken by incore instrumentation and guide tubes. Each fuel assembly weighs about 1,358 pounds. The apr'licant requests authorization to possess 198 assemblies, which includes 5 replacement assemblies, for the initial core load.

II. SCOPE OF REVIEW The staff reviewed and discussed the CECO application with the NRR Project Manager, the Resident Inspector, and the applicant's staff. The Interim Security Plan was reviewed and approved by the NRC's Division of Safeguards. j i

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III. POSSESSION LIMITS

, = The' applicant .has requested authorization to receive 2,200 kg U-235 '~as low enriched uranium' oxide. The U-235 enrichment will range from 2.1 w/o U-235 +

Lto 3.55 w/o U-235. '.To-accommodate this request, the staff recommends the; Y .following license condition:

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Naterial. Form Quantity u-Uranium enriched U02 in reactor 2,200 kg U-235 up to 3.55 w/o.in fuel assemblies (198 assemblies) the U-235 isotope

'IV AUTHORIZED ACTIVITIES ,

TheLapplicant requests authorization to receive, possess, inspect, and store ,

fuel' bundles at Braidwood Station, Unit 2. The applicant has also requested

. authority' to package such fuel bundles for delivery to a carrier. This latter authority is covered under a general license in 10 CFR 71 and hence will not be authorized in this license.- To authorize the other activities, the staff recommends the following license conditionsi The licensee is authorized to receive, possess, use, and transfer uranium in fuel bundles in accordance with statements, representations, and conditions in the revised license application dated May 6, 1987, and Response 7 in the supplement dated September 18, 1984.

The authorized place of use is the Braidwood Station, Unit 2, 20 miles south-southwest of the Town of Joliet', Illinois.

V. ORGANIZATION AND ADMINISTRATIVE CONTROLS

The Station Manager exercises overall managerial and supervisory responsibility for the safe operation of the-plant and its equipment. He is responsible for compliance with the Station's NRC licenses, government regulations, and the

-applicant's quality assurance program.-

!A. Radiation Control The Station Radiation / Chemistry Supervisor is responsible for the Station's radiation protection and environmental inonitoring programs. During unusual or abnonnal. radiological conditions, he has access to the Station Manager on matters relating'to radiation protection.

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q Commonwealth Edison Company, SER 3 JUL 2 71987 The Station Health Physicist is responsible for. daily health physics activities.

He reviews health physics surveys and dose, internal deposition, meteorological, and environmental data. As assigned, he participates in the health physics aspects of emergency planning and environmental monitoring activities and assumes the duties of the Radiation / Chemistry Supervisor.

B. Nuclear Criticality Safety The Fuel Handling Foreman is responsible for developing and implementing procedures involving the control and handling of nuclear fuel.

The Operating Engineer is responsible for the operation of the mechanical and electrical equipment and certain common plant systems, such as radioactive ,

waste processing and fuel handling. He is also responsible for authorizing functional acceptance tests to be conducted by operation and/or technical staff personnel.

The Technical Staff Supervisor provides technical support for plant operations and adequacy of Station procedures. He also has the responsibility and authority for implementation of the onsite review function.

C. Technical Qualifications The minimum technical qualifications for the Station Manager cre in accordance with Section 4.2.1, " Plant Manager," of ANSI N18.1-1971.

The minimum technical qualifications for the Radiation / Chemistry Supervisor are in accordance with Sections 4.4.3 and 4.4.4, " Radiochemistry" and " Radiation Protection," of ANSI N18.1-1971. These qualifications may alternately be met by technical personnel reporting to the Radiation / Chemistry Supervisor. The  ;

minimum technical qualifications for the Radiation / Chemistry Supervisor or the  !

Station Health Physicist are in accordance with the requirements for the Radiation Protection Manager of Regulatory Guide 1.8, September 1975.

The minimum technical qualifications for the Operating Engineer are in accordance with Section 4.3.2, " Supervisors Not Requiring AEC Licenses," of ANSI N18.1-1971. The minimum technical qualifications for the Fuel Handling Foreman are in accordance with Section 4.3.1, " Supervisors Requiring AEC Licenses," of ANSI N18.1-1971. The minimum technical qualifications for the Technical Staff Supervisor are in accordance with Section 4.6.1, " Engineer in Charge," of ANSI N18.1-1971.

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0 Commonwealth Edison Company, SER 4  ;

D. Training Prior to receipt of fuel onsite, radiation safety personnel will be trained in i radiation safety, as outlined in Section 13.2.1.7 of the Byron /Braidwood FSAR, and in CECO radiation protection procedures related to fuel assembly handling.

Fuel handling personnel will receive training in proper fuel handling procedures, '

including the related health and safety aspects of the activfties, prior to receipt of fuel onsite. l 7

E. Procedures The Fuel Handling Foreman develops and implements procedures involving the control and handling of nuclear fuel. Procedures, and changes thereto, for the control and handling of nuclear fuel shall be reviewed and approved by an 2 0)erating Engineer and the Technical Staff Supervisor. At least one individual ,

w1o reviews and approves fuel handling procedures will hold a Senior Reactor Operator License for Braidwood Station.

VI. NUCLEAR CRITICALITY SAFETY 1 A. General Fuel assemblies may be handled and stored in shipping containers in the receiving area, the new fuel storage vault, and the spent fuel storage pool in the Fuel Handling Building.

t Each fuel rod consists of slightly enriched uranium dioxide ceramic pellets contained in Zircaloy-4 tubing. The fuel pellets are right circular cylinders with a nominal diameter of 0.3088 inches. The Zircaloy-4 cladding has a nominal ,

l thickness of 0.0225 inches and a nominal outside diameter of 0.360 inches. Each fuel assembly contains 264 fuel rods positioned in a 17 x 17 array with a rod pitch of 0.496 inches. The remaining 25 positions in the fuel assembly contain incore instrumentation and guide thimbles. Calculations performed by the NRC staff assumed fuel assemblies containing uranium enriched to 3.55 w/o U-235 for l the new fuel storage vault and to 3.22 w/o U-235 for assemblies in the spent l fuel storage pool. These are the respective maximum enrichments requested by l

the applicant for the storage vault / pool. To be conservative, each assembly was assumed to have 289 fuel rods in a 17 x 17 array.

B. Shipping Container Storage i Fuel assemblies will be received in Westinghouse Models RCC-1 or RCC-3 shipping containers (package) presently licensed under NRC Certificate of Compliance No. 5450. Certificate of Compliance No. 5450 authorizes Fissile Class I shipments of 250 packages, each containing 2 assemblies. Because there is no limit associated with the number of undamaged Fis:ile Class I packages, no 1, criticality safety controls need be established for the container storage area.  !

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I C. ' New Fuel-Storage Vault I

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The new fuel will be stored in racks in a reinforced concrete vault. The

' storage racks are composed of cells. fastened together in a 2.x 22 module which l tis attached to a high-strength carbon-steel support structure. There are three

such modules in'the vault. .The individual' cells are on 21-inch centers.- Each cell consists of 1/8-inch-thick Type 304 stainless: steel square cylinders with-9.00-inch inside dimensions. The. lattice spacing between the modules are

, greater than 55 inches edge-to-edge. The applicant reported.that the k for-the3-modulearraywouldbelessthan0.98withfuelofamaximumpine$ffchment

'o_f 4.00 w/o:U-235 under optimum water moderation conditions. The calculations

. were performed using KEN 0, a Monte Carlo. code, and a 123 energy group cross-section set which was generated from a basic GAM-THERMOS library using the NITAWL routine in the AMPX code package. The staff calculated the array assuming a maximum .

enrichment of 3.55 w/o U-235 and used a-27 group cross-section set which is found in the SCALE program, along with KENO. The staff has determined the k f for the array to be about 0.85. The k isindependentofthedegreeofwatefboderation withinandbetweenassemblies6ffthe degree of concrete reflection surrounding the array.

.D. Spent Fuel Pool Storage

. Thel spent fuel storage racks consist of square stainless-steel tubes of 1/8-inch thick Type 304 stainless steel. The tubes are held at 14 inches center-to-center by 1/8-inch thick Type 304 stainless steel plates. . The pool has a total capacity of 1,050 fuel assemblies. The applicant has indicated that, prior to the initial- core loading, the pool will be used for dry storage of fuel with up to 3.22 w/o U-235 enrichment in only a checkerboard pattern or wet storage of initial core fuel with up to 3.22 w/o U-235 enrichment in borated water of at least 2,000 ppm boron concentration. The applicant reported that the k for the checkerboard array would be less than 0.98 with fuel of a maximum e8bfchment of 3.20 w/o U-235 and for the wet storage array less than 0.95 with fuel of a maximum enrichment of 4.00 w/o U-235. The applicant's calculations were performed using PDQ-7, a four-group diffusion theory code. Cross-sections for these calculations were generated with NUMICE,.a version of the LEOPARD code. Selected i cases were verified by the applicant using KEN 0 and the GAM-THERMOS cross-section library. The staff calculated both arrays using KEN 0 and assumed a maximum enrichment of 3.22 w/o U-235. The 27 group cross-section set that is found in the SCALE program was used for the staff's calculations. The staff has determined that the k 0.83andO'$k,forthe~checkerboardarrayandthewetstoragearraytobeabout respectively. No credit was taken for the presence of the boron in any calculation, and the k is independent of the degree of water moderation  !

within and between assemblies *$f the degree of concrete reflection surrounding the array.

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AE l Commonwealth Edison Company, SER 6 E. Handling of Fuel Assemblies To ensure that fuel assemblies outside of storage remain safely.suberitical at all times, the applicant has committed to having no more than one fuel assembly 1 out of a shipping container or a storage' rack at a time.

The plastic dust wrapper on each fuel assembly in the vault or the pool must be removed from the fuel assembly or must be open at the bottom so that water will.

not collect in the. wrapper. If the storage array were to become flooded, the dust wrappers filled with water, and then the pool or vault drained, the fuel assemblies could be well-moderated and effectively coupled to other well-moderated j fuel assemblies because the isolating water between the fuel assemblies had j drained away. The staff evaluated the condition of full density water within the fuel assembly and low density water between fuel assemblies. ~There is not enough steel in the storage racks to assure that the array remains subcritical under this condition. The revised application fails to state what type of. controls i will'be placed on the dust wrappers. Therefore, the staff recommends the

, following condition:

The dust wrapper on each fuel assembly stored in the vault or the pool shall be removed from the fuel assembly or shall be,open at the bottom so that water will not collect in the wrapper.

F. Exemption The applicant requests an exemption from the monitoring requirements of 10.CFR 70.24 as provided in 70.24(d). Tne applicant's reason for requesting the exemption.is that the procedures and storage facilities provide assurance that inadvertent criticality cannot occur during receipt, handling, and storage of nuclear fuel assemblies at the Braidwood Station.

The applicant's reason for exemption is valid and good cause exists for the exemption. The storage racks provide physical protection to ensure suberiti-cality. The procedural controls provide reasonable assurance that nuclear criticality will not occur during fuel handling, and monitoring is not needed.

Even if the procedural controls were violated, optimum conditions of neutron moderation, physical spacing, and neutron reflection would be required for assemblies to be in a critical situation.

l The procedural controls, considering the limited activities and material handling methods, are deemed adequate-to grant the exemption. This exemption is autho-rized by law, will not endanger life or property or the common defense and I security, and is otherwise in the public interest. Therefore, the following )

license condition is recommended: l The licensee is hereby exempt from the requirements of 10 CFR 70.24, insofar as this requirement applies to materials possessed under this i license, j 2

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.y Commonwealth. Edison Compa'ny, SER 7 JtJL 2 71987 VII' . . RADIATION SAFETY The applicant's radiation protection program includes assignment-of respon-

'sibility for radiation protection surveys, personnel monitoring. devices, instrument-calibration, Braidwood. Station Radiation. Protection Procedures,

p. and Ceco Radiation Protection Standards. The primary hazard from encapsu-

. lated low-enriched uranium is low-level radiation. The applicant's program, combined;with 10-CFR.Part 20 requirements, is adequate to protect the health and safety of the public.-

VIII. ENVIRONMENTAL PROTECTION The NRC staff. prepared.an Environmental Assessment for the proposed activities at Braidwood Station, Units 1 and 2. Based on the Assessment, a Finding of No Significant Impact was made pursuant to 10 CFR 51. The Finding was published iin the Federal Register on September 5, 1986.

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IX. FIRE SAFETY The applicant proposed to provide fire protection by combination of fire dampers in the fire wall, portable fire extinguishers and manual hose stations, and an  ;

automatic fire detection system. The fire dampers are fusible link-type which

.will be closed when the temperature exceeds 165'F. The automatic fire detection system consists of ionization and ultraviolet detectors which alarm and annunci-ate. locally and in the control room. Considering that all licensed material is encapsulated, fire protection is adequate.

X. PHYSICAL PROTECTION The Division of Safeguards, NMSS, determined that the CECO security plan meets the requirements of 10 CFR 73.67 and recommends the following license condition:

The licensee shall maintain and fully implement all provisions of the P

Commission approved ant to the authority of 10 Physical CFR 70.32Security.The (e)lan, including approved changes Physical made pursu-Security Plan consists of Revision 1 to the " Security Plan for Special Nuclear Materials. Security for Commonwealth Edison Company, Braidwood Station,"

submitted by letter dated April 19, 1985. The Physical Security Plan shall be withheld from public disclosure pursuant to 10 CFR 2.790(d).

.XI. CONCLUSION

'A. After reviewing the application and its supplement, the staff finds that:

1. CECO meets the requirements of the Atomic Energy Act, as amended, and the Commission's regulations;

Commonwealth Edison Company, SER 8

2. Issuance of the license would not be inimical to the common defense and security; and

3. Issuance of the license would not constitute an unreasonable risk to the health and safety of the public. ..

B. With the recommended license conditions, the NRC staff finds that:

1. CECO is qualified by reasons of training and experience to use the material for the purpose requested in accordance with regulations in 10 CFR Part 70.

2. CECO's proposed equipment, facilities, and procedures are adequate to protect health and minimize danger to life or property.

XII. RECOMMENDATION The staff recommends issuance of the special nuclear material license provided the conditions identified above are incorporated into the license.

Original Signed Byt David A. McCaughey Uranium Fuel Section Fuel Cycle Safety Branch Division of Fuel Cycle. Medical, Academic, and Commercial Use Safety prisinal Signed By:

Approved by: M tal y tt' Leland C. Rbuse, Chief Fuel Cycle Safety Branch O fb o T ',,'r A

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. jck' . UNITED STATES :

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AMEN 0 MENT TO INDEMNITY AGREEMENT N0. B-102-AMENDMENT NO. 4 l

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l LEffectiven., . , Indemnity; agreement- No. .B-102,1between .

Commonwealth Edison Company and the > Nuclear Regulatory Commission; dated 1

• October 8,.'1985, as amended,-is .he'reby further amended as follows: .l Item'3 of-the' Attachment to the. indemnity; agreement is deleted in its entirety and the following. substituted:therefore:-

L Item 3 - License number or numbers ,

1 SNM-1938 (From12:01~a.m., October.8,1985,to ,

12 midnight, October 16, 1986,. ~ '!

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,NPF-59. (From 12:01'a.m., October' 17, 1986, to 12 midnight, May 20, 1987,. .

inclusive) l NPF-70 (From 12:01-a.m., May 21, 1987, to 12 midnight, July 1, 1987,

' inclusive).

'NPF-72 (From-12:01'a.m.. July 2,1987)

SNM-1945 (From 12:01-a.m. g 37 g )

FOR THE'U.S. NUCLEAR REGULATORY COMMISSION f:

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.Jfsse L.. Fu7ttles, Chief Policy Development and Technical Support Branch Program Management, Policy Development and Analysis Staff 0ffice of Nuclear Reactor'. Regulation

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' AMENDMENT TO INDEMNITY' AGREEMENT NO.'B'-102 AMENDMENT NO. 4 RJ 7 1987 ..

,,: Indemnity agreement No. B-102,;between.

? Effective Commonwealth Edison Company and the Nuclear Regulatory Commission, dated.

October 8,1985, as: amended,s is hereby. further amended as ,follows:

Item 3lof the.' Attachment to the' indemnity agreement is. deleted in its entirety and the following substituted therefore:

Item'3 - Licenseinumber or numbers SNM-1938 ~ (From 12:01 a.m. ,~ 0ctober 8,1985, to~

12 midnight, October 16, 1986; inclusive)

NPF-59 (From 12:01 a.m.,' October.' 17', 1986, to 1 12 midnight, May 20, 1987,  !

-inclusive)

NPF (From 12:01 a.m., May 21,.1987, to

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l 12 midnight, July 1, 1987,.  !

inclusive)~

NPF-72 (From12:01'a.m., July 2,1987)

SNM-1945 (From 12:01 a.m. M 2 7 M )

FOR THE U.S. NUCLEAR. REGULATORY COMMISSION:

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Jsse L. FuncTies, Chief folicy Development and Technical fj Support Branch Program Mai agement, Policy Development and Analysis Staff Office of Nuclear Reactor Regulation 1

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commonwealth taison co.

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