ML18018B619
| ML18018B619 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 04/12/1984 |
| From: | Utley E CAROLINA POWER & LIGHT CO. |
| To: | Jennifer Davis NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NLS-84-075, NLS-84-75, NUDOCS 8404180189 | |
| Download: ML18018B619 (72) | |
Text
REGULATORY ORMATION DISTRIBUTION SYS>>1 (RIDS)
ACCESSION NBR: 8404180189-. DOO ~ DATE: 84/04/12 NOTARIZED: NO DOCKET-'
FACIL".50 BYNAME 400 Shear on .Har ris Nucl ear power Pl antg Uni t. 1> Car ol ina<<05000400 AUTH AUTHOR 'AFFILIATION UTLEYgE,E, Carolina"f'ower<<,L Light Co ~
"RECIP,'NAME: RECIPIENT AFFILIATION Material DAVIS p J,G, Of f i ce', of Nuc'l ear 'Safety 8 'Saf eguardsiDir ector ~
SUBJECT:
Application for"'SNM license',to permit receiptipossessioni
>storage 8 .pr eparation for 'SNM tr anspor t. Fuel <<shipment
>>scheduled <<to begin in Jan 1985 ~ .
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TITLE: Licensing 'Submittali >>PSAR/FSAR Amdts. 8 Related 'Cohr espondence'OTES!
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Carolina Power 8 Light Company SERIAL: NLS-84-075 P. O. Box 1551 ~ Raleigh, N. C. 27602 APR 12 1984 E. E. UTLEY Executtve Vlcc President Power Supply and Engineering tt Constructton Mr. John G. Davis, Director Office of Nuclear Material Safety and Safeguards U. S. Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NO+ 1 DOCKET NO~ 50-400 SPECIAL NUCLEAR MATERIALS LICENSE APPLICATION
Dear Mr. Davis:
Carolina Power & Light Company hereby submits an application for a Special Nuclear Material License. Eight (8) copies of the application, as required by Regulatory Guide 3.15, are enclosed for your review and approval.
The license herein applied for will permit the receipt, possession, storage, and preparation for transport of Special Nuclear Material required for the operation of the Shearon Harris Nuclear Power Plant, Unit 1, which will be licensed pursuant to the requirements of 10 CPR Part 50. As specified in Regulatory Guide 3.15, a fee is not required for this application since a construction permit has been issued for the SHNPP.
Section 1.3 of this application contains privileged commercial information relating to the physical protection of the plant and Special Nuclear Material which is being withheld from public disclosure under the provisions of 10 CFR t't2.790 and 9.12. Therefore, Section 1.3 is being submitted under a separate letter along with the required affidavit.
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Mr. John C. Davis Fuel shipment to the SHNPP is presently scheduled to begin in January 1985.
Therefore, this license will be required by December 31, l984. Your prompt consideration of this Special Nuclear Material License application will be appreciated.
Yours very truly, E. E. Utley GAS/lcv (8629NLU)
Enclosures CC ~ Mr. G. 0. Bright (ASLB) Mr. J. P. O'Reilly (NRC-RlI)
Mr. B. C. Buckley (NRC) Mr. R. G. Page (NRC)
- Dr. J. H. Carpenter (ASLB) Mr. Travis Payne (KUDZU)
Mr. Wells Eddleman Mr. Daniel F. Read (CHANGE/ELP)
Mr. J. C. Kelly (ASLB) Mr. John D. Runkle Dr. Phyllis Lotchin Dr. Richard D. Wilson Mr. G. F. Maxwell (NRC-SHNPP) Wake County Public Library Chapel Hill Public Library
t p, 5 3
SHEARON HARRIS NUCLEAR POWER PLANT APPLICATION FOR SPECIAL NUCLEAR MATERIAL LICENSE Carolina Power & Light Company (CP&L) and the North Carolina Eastern Municipal Power Agency (NCEMPA) (hereafter called "Applicants" ) pursuant to 10 CFR 70, hereby apply for a license to permit the receipt, possession and storage of Special Nuclear Material (SNM) consisting of unirradiated nuclear fuel assemblies, incore fission chambers, and sealed sources for irradiation surveillance capsules as described herein for the Shearon Harris Nuclear Power Plant (SHNPP), Unit 1. The ter m of the license requested is for the period beginning January 1, 1985, until such time as it may be superseded by the permanent operating license.
The applicants are co-owners of the SHNPP. The NCEMPA owns 16.17$ of the facility and CP&L owns the remaining 83.83$ of the facility. CP&L has exclusive responsibility for the design and construction of the facility.
CP&L is an investor -owned utility engaged in the generation, distribution, and sale of electricity with its principal office located in Raleigh, North Carolina. The applicants are not owned, controlled or dominated by an alien, foreign corporation or foreign government.
CP&L is a public service corporation formed under the laws of North Carolina in l926. The names and addresses of CP&L's principal officers, all of whom are citizens of the United States, are as follows:
NAME POSITION Sherwood H. Smith, Jr. Chairman/President and Chief Executive Officer E. E. Utley Executive Vice President Edward G. Lily, Jr. Executive Vice President and Chief Financial Officer William E. Graham, Jr. Executive Vice President Wilson W. Morgan Senior Vice President M. A. McDuffie Senior Vice President James M. Davis, Jr. Senior Vice President Lynn W. Eury Senior Vice President James M. Davis, Jr . Senior Vice President Russell H. Lee Senior Vice President Charles D. Barham, Jr. Senior Vice President and General Counsel Paul S. Bradshaw Vice President and Controller (8540NLU/lcv)
J. L. Lancaster, Jr. Seer etary and Manager L. Thompson Quarles Treasurer The address of the foregoing principal officers of CP&L is:
Post Office Box 1551 Raleigh, North Carolina 27602 The NCEMPA is a public body corporate and politic and an instrumentality of the State of North Carolina created pursuant to the Joint Municipal Electric Power and Energy Act, Chapter 159B of the General Statutes, as amended, of North Carolina. The names of NCEMPA's Board of Commissioners, all of whom are..
citizens of the United States, are, as follows:
The Honorable Frederick E. Turnage, Chairman City of Rocky Mount Mr. Peter G. Vandenberg, Vice Chairman City of Laurinburg Mr. Charles O'. Horne, Jr. Secretary-Treasurer City of Greenville Mr. Ralph W. Shaw, General Manager Mr. Ronald Wicker Mr. Jordan C. Horne Town of Apex Town of Ayden Mr. Steven S. Weatherman Mr. Steven L. Harrell Town of Belhaven Town of Benson Mr. Charles R. Stewart Mr. Willis Pr ivott Town of Clayton Town of Edenton Mr. Joseph B. Anderson Mr. Connie Price City of Elizabeth City Town of Fremont Mr. J. A. Wooten, Jr. Mr. W. P. Riley, Jr.
Town of Farmville Town of Hamilton Mr. E. A. Warren Mr . R. G. Anthony City of Greenville Town of Hobgood Mr. Jesse Harris Hon. Simon C. Sitterson, Jr.
Town of Hertford Town of Kinston Mr. Gene C. Hill Mr. Peter Vandenberg Town of Hookerton City of Laurinburg Mr. Edward B. Walters Mr. Harry L. Ivey Town of LaGrange City of Lumberton (8540NLU/lcv)
Hs. Lois Brown Wheless Mr. C. Vance Greeson Town of Louisburg City of Pikeville Mr. Boyd C. Hyers Hr. Ralph S. Mobley City of New Bern Town of Robersonville Mr. John McNeill Mr. Joe Edwards, Jr.
Town of Red Springs Town of Selma Hr. Frederick E. Turnage Mr. Hugh C. Talton Ci.ty of Rocky Mount Town of Smithfield Mr. N. 0. McDowell, Jr. Mr. J. Ray King Town of Selma Town of Tarboro Mr. W. Robert Thorsen Hr. Abbot t N. Sawyer City of Southport Town of Washington Hr. Rodney V. Byard Hr. T. Bruce Boyette Town of Wake Forest City of Wilson The office address of the NCEMPA is:
Post Office Box 95162 3117 Poplarwood Court Raleigh, North Carolina 27625 Communication pursuant to this application should be sent to:
E. E. Utley Executive Vice President Carolina Power & Light Company 411 Fayetteville Street Box 1551 Raleigh, North Carolina 27602 1.0 GENERAL INFORMATION Reactor and Fuel The SHNPP is located in the southwest corner of Wake County and the southeast corner of Chatham County, NC. The site is approximately 16 miles southwest of Raleigh and 15 miles northeast of Sanford, NC.
CP&L has the overall responsibility to ensure that the plant is designed, constructed, and operated without undue risk to the health and safety of the public. Ebasco Services, Incorporated is the architect/engineer responsible for the design, engineering, and equipment and material procurement for SHNPP. This includes all plant structures, systems, and components except for those provided by Westinghouse Electric Corporation, the Nuclear Steam Supply System (NSSS) Supplier. Daniel (8540NLU/lcv)
Construction Company, Inc., as the constructor, performed the major part of the plant construction. Selected portions of the work, however, were performed by other contractors under direct supervision of CP&L-The docket number for the SHNPP, Unit 1 is 50-400. The construction permit number is CPPR-158.
The nuclear fuel assemblies consist of a 17x17 array of 264 fuel rods, 24 control rod guide tubes, and 1 instrument tube. The Zircaloy-4 tubes are supported laterally in a 0.496-inch rod pitch by 8 inconel spacer grids, and supported on the top and bottom by stainless steel end fittings to form a rod bundle 159.77 inches long, with an 8.426-inch envelope. Each fuel rod contains 144 inches of slightly enriched uranium dioxide in the form of 0.3225-inch diameter pellets. The cladding has a 0.374-inch outer diameter and is 0.0225 inches thick.
The initial SHNPP core will contain three fuel regions, with assembly enrichments of 2.1, 2.6, and 3.1 weight percent U-235. The weight of a new fuel assembly, including structural material, is 1,467 pounds, with a maximum U-235 weight of 31.5 pounds. Fresh fuel contains no U-233, plutonium, thorium, or depleted uranium.
This application is for a full core of 157 assemblies containing 159,705 pounds of uranium; 4,157 pounds of whi,ch is U-235.
1.2 Stora e Conditions General Arrangement Drawings CAR-2165 G-022 through G-026 (Attachment 1)* show the Fuel Handling Building (FHB) of the SHNPP.
The SHNPP Fuel Storage Facilities are contained within the plant FHB. The FHB houses the fuel storage facilities consisting of two new fuel pools, two spent fuel pools, two 100 percent cooling systems, and cleanup equipment to remove the particulate and dissolved fission and corrosion products resulting from the spent fuel.
The new fuel pools are designed for the storage of both new and spent PNR fuel. Consequently, they are designed for both wet and dry storage. The maximum storage capacity of the two new fuel pools is 1160 PWR fuel assemblies, each pool having a capacity of more than 3 cores. The fuel is stored in a combination of 6 x 10 and 7 x 10 PNR rack modules, which are designed for underwater removal and installation. The fuel racks consist of individual vertical cells fastened together through top and bottom supporting grid structures to form integral modules. A neutron absorbing material is encapsulated into the stainless steel walls of each storage cell. These free-standing, self-supporting modules have a center-to-center spacing of 10.5 inches between cells, which is sufficient when combined with the neutron absorber material to maintain a subcritical array even in the event the pools are flooded with unborated water. The fuel rack dimensions are as follows:
- Full size drawings only sent with copy to Hr. R. G. Page (NRC)
(8540NLU/lcv)
C-C Spacing 10.500 in.
Cell I.D. 8.750 in.
Poison Cavity 0.090 in.
Poison Width 7.500 in.
Cell Gap (Nominal) 1.330 in.
Poison Thickness 0.075 in.
Wall Thickness 0.075 in.
Wrapper Thickness 0.035 in.
Poison(gms. B-10 sq.cm) 0.020 The fuel pools are interconnected by means of a transfer canal which runs the length of the FHB. The pools are normally isolated by means of removable gates. The pools are concrete structures with stainless steel liners for compatibility with the pool water. Provisions are made to limit and detect leakage of the new fuel pools. This is further discussed in FSAR Section 9.1.3.
The FHB is designed in accordance with Regulatory Guide 1.13, Revision 1 and provides protection to the fuel racks and other pieces of equipment against natural phenomena. Further information on this subject can be found in FSAR Sections 3.3, 3.4, and 3.5.
The design and safety evaluation of the fuel racks is in accordance with the NRC position paper, "Review and Acceptance of Spent Fuel Storage and Handling Applications." The SHNPP will use spent fuel racks for the storage of new fuel. The racks, being ANS Safety Class 3 and Seismic Category 1 structures, are designed to withstand normal and postulated dead loads, live loads, loads due to thermal effects, loads caused by the operating basis earthquakes and safe shutdown earthquakes in accordance with Regulatory Guide 1.29, and stress allowables defined by the ASME Code, Section III. The racks can withstand an uplift force equal to the maximum uplift capability of the spent fuel bridge crane. The design of the racks is such that with fuel of the highest anticipated enrichment (3.9 w/o U-235 Westinghouse 17x17 optimized fuel) and the pool flooded with unborated water, K ff is < 0.95. If the unborated water is replaced by other effective moderators such as foam or water mist, Keff is < 0.95.
Consideration is given to the inherent neutron absorbing effect of the materials of construction. Fuel handling accidents will not alter the rack geometry to the extent that the critically acceptance criteria is violated. Further information on the critically safety analysis is located in FSAR Section 4.3.2.6.
All materials used in construction are compatible with the storage pool environment, and all surfaces that come in contact with the fuel assemblies are made of annealed austenitic stainless steel.
The maximum storage capacity of the two spent fuel pools is 3024 PWR assemblies. The total storage capacity of both the new and spent fuel pools is 4184 PWR assemblies. Fuel is stored in a combination of 6x8, 6x10, and 7x7 PWR rack modules designed for underwater removal and (8540NLU/1cv)
installation on existing pool floor embedments should rack rearrangements be desired. Rearrangement of the racks would have no effect on maximum stored fuel critically.
The design of the spent fuel storage racks precludes fuel insertion in other than prescribed locations, thereby, preventing any possibility of accidental critically. A lead-in opening is provided for each storage location, and the storage cells provide full length guidance for the fuel assembly.
FSAR Section 9.1.4 (Attachment 2) gives a description of the SHNPP fuel handling system.
The new fuel inspection stand is a platform at elevation 286'hat is permanently fixed to the wall above the new fuel container set down area. After the fuel assemblies are removed from the shipping containers, they will pass by the stand for inspection prior to being loaded into the fuel racks.
Due to the design of the FHB and the New Fuel Storage Pool, the term "adjacent areas" as used in this submittal is meant to include the area within the FHB at the 286'levation. This elevation is separated from the balance of the plant and the environment by structural concrete walls. Access to this elevation is limited to hatches and stairwells.
Prior to receipt and storage of new fuel, all major construction and testing of the necessary portions of the FHB will be complete. The FHB is considered complete for receipt of the first core when the new fuel system (New Fuel Handling Tool, Rod Cluster Control Assembly Handling Tool, and Unit 1 New Fuel Elevator), the FHB Auxiliary Crane, the Spent Fuel Pool Bridge Crane, and the Unit 1 New Fuel Pool Cooling and Cleanup System are tested, in addition to the installation of all fuel storage racks in the storage location for the first core of new fuel (see Attachment 3).
The capacity of the spent fuel pool bridge crane is two tons; the hoist will be interlocked at 2250 pounds to preclude excessive loads from being carried by this crane over new fuel.
The capacity of the auxiliary crane is 12 tons. This crane will be used to install the fuel racks in the new fuel storage area, after which time, the auxiliary crane travel will be restricted administratively to north of the storage location for first core of new fuel, to which this li,cense applies, while handling excessive loads (refer to Attachment 3) or a safe load path prepared in accordance with NUREG-0612 will be used.
The activities that will be conducted in the areas adjacent to the new fuel storage area can be grouped into one of four categories as described below:
- a. ~Activit PrevenCive and corrective maintenance on equipment/systems in the FHB.
(8540NLU/lcv)
Potential Effects on Storage Minimal Work will be performed by or under the supervision of CP&L Maintenance personnel. Any work within the security control area established to protect the fuel will require maintenance personnel to be authorized for entry to the area.
- b. '~Activit Construction completion activities to resolve discrepancy items on equipment/systems turned over to the Start-up Test Group or to complete minor work on equipment/systems in preparation for turnover.
Potential Effects on Storage Minimal The security control area has been established and protective covering will be provided to ensure that any work in adjacent areas will not damage the fuel. Additionally, work to resolve discrepancy items on equipment/systems turned over to the Start-up Group will be controlled by the Start-up Manual Construction Work Request/Authorization Form which requires the Start-up Engineer's approval prior to starting work.
Any work within the security control area will require the construction personnel to be authorized for entry to the area.
- c. Activit Start-up testing on equipment/systems that could inc u e checkout,'lushing, hydrostatic testing, and preoperational testing.
Potential Effects on Storage Minimal Testing will be performed by or under the supervision of assigned system Start-up Engineers who will ensure that the testing activities have no impact on the stored fuel. Testing support for the Start-up Engineer will be provided by maintenance technicians, maintenance mechanics, operators, chemistry technicians, or construction personnel as appropriate for the testing being performed. Any work within the security control area established to protect the fuel will require personnel to be authorized for entry.
The following equipment/systems could be subject to flushing, hydrostatic testing, or preoperational testing after
'receiving/storing new fuel:
Security Systems Service Air Instrument Air Demineralized Water Fuel Pool Cooling Fuel Pool Cleanup Fire Protection Refueling System ds. ~Activit Routine housekeeping.
(8540NLU/lcv)
Potential Effects on Storage Minimal Housekeeping activities will be planned and supervised to ensure that the activities have no detrimental effect on the stored fuel. The fuel will be covered to provide protection from dust, etc. Chemicals used in the vicinity of the new fuel will be reviewed for compatibility with fuel materials prior to use. The Manager, Environmental and Radiation Control, will develop and approve this list of accepted chemicals.
- e. ~Activist New fuel receipt and inspection.
Potential Effects on New Fuel None Receipt and inspection of new fuel will be conducted in accordance with approved SHNPP procedures and under the direction of a shift foreman or other qualified individual.. Refer to Section 2.2.
The FHB is separated from other structures by three-hour fire barriers. Operations in this building are not related to safe shutdown of the reactors. Safety related equipment is, however, present in this building. Combustible loading is minimal. Hand portable extinguishers and standpipe hose stations are installed throughout the building. Location of the hose stations in the new fuel storage area and wetting by fire protection water is considered acceptable, because the new fuel storage racks are designed to retain the subcritically of the storage array even when flooded with unborated water. Due to the absence of combustible materials and the large room volume provided by a 50 foot ceiling height, automatic fire detectors are not provided in general areas. Manual fire alarm stations are located throughout the FHB near hose stations with local alarm and annunciation in the control room. Protection for the spent fuel pool area is provided by portable extinguishers and hose stations. The manual hose stations and portable extinguishers located in the vicinity of the new fuel storage and new fuel inspection areas will be installed prior to receipt of new fuel.
Additional information on fire protection is located in Section 9.5.1 of the FSAR.
A description of the controls for preventing unauthorized access to areas where SNM is stored is given in Section 1.3.
Ph sical Protection This section contains privileged commercial information relating to the physical protection of the plant and SNM which is being withheld from public disclosure under the provisions of 10 CFR 2e790 and 9.12. As such, this section is being transmitted via a separate letter along with the required affidavit.
Transfer of S ecial Nuclear Material Packaging and shipment of new fuel to the SHNPP site will be the responsibility of the fuel fabricator, Westinghouse Electric Corporation. Shipping container handling, removal of fuel from containers, and fuel inspection for shipping damage will be performed (8540NLU/icv)
according to specifications supplied by Westinghouse. Following receipt inspection, the new fuel will be placed into the new fuel racks for storage.
SNM records are maintained showing receipt, inventory, disposition, and transfer of all SNM in CP&L's possession. The DOE and NRC are notified via DOE/NRC Form 741 as to the amount of material shipped, date of shipment, and shipping and receiving parties. This DOE/NRC Form 741 is issued the day of shipment by the responsible shipper.
The SHNPP Onsite Operations Section/Technical Support Unit/Reactor Engineer will be responsible for maintaining records for the new fuel received at the SHNPP.
1.5 Financial Protection and Indemnit The applicants will apply for nuclear energy liability insurance with the American Nuclear Insurers in the Amount of $ 1,000,000 to satisfy the financial protection requirements of 10 CFR 140.13. The effective date of insurance coverage will be from the time new fuel is received at the SHNPP until the first fuel assembly is loaded into the reactor. Proof of such financial protection will be furnished prior to issuance of a Special Nuclear Material License pursuant to this applicati.on.
2.0 HEALTH AND SAFETY 2.1 Radiation Control Minimum qualifications for those positions responsible for radiation safety require experience in applied radiation protection at a nuclear facility dealing with radiation protection problems and programs similar to those at nuclear power plants. These positions also require a bachelor's degree or equivalent in a science or engineering subject including some formal training in radiation protection.
Experience in applied radiation protection shall be in compliance with the requirements of Regulatory Guide 1.8 (Revision l), "Personnel Selection and Training" as stated in the FSAR Section 1.8.
At SHNPP, the key radiation safety personnel are the Manager Environmental and Radiation Control and the Radiation Control Supervisor. Their responsibilities are as follows:
a~ The Manager Environmental and Radiation Control is responsible to plant management for the following:
Planning the overall activities and tasks of the Radiation Control Subunit in cooperation with other unit managers in order to develop an integrated plant operations program.
Developing and approving Radiation Control procedures and instructions to ensure safe and reliable operation of equipment in an ALARA manner.
(8540NLU/lcv)
Establishing and regularly reviewing the required Radiation Control records. Supervising the preparation of Radiation Control reports to ensure they are complete, accurate, and submitted in a timely manner.
Assuring that programs, procedures and policies of other uni,ts have been reviewed by the Radiation Control Subunit to ensure ALARA philosophies have been incorporated.
- b. The Radiation Control Supervisor is responsible to the Hanager-Environmental and Radiation Control for the following:
Ensuring that plant activities are conducted in a manner to protect the plant, employees, visitors, general public, and the surrounding community consistent with the SHNPP Radiation Control programs.
Ensuring that Radiation Control programs and related procedures meet SHNPP needs and regulatory requirements.
Providing adequate documentation related to individual radiation exposures and audits.
Ensuring that an adequate supply of materials are provided as required for radiation protection activities.
Ensuring adequate procedural guidance is available to facilitate sound Radiation Control practices.
The person responsible for radiation safety at the SHNPP is the Hanager Environmental and Radiation Control. The training and experience of the manager are shown in the attached resume (Attachment 4).
Each unirradiated fuel assembly will be tested for contamination using a standard smear technique and counted for contamination with an alpha scintillation survey meter, a low-background alpha-beta gas flow proportional counter, or a beta-gamma counter-sealer system. If a fuel assembly reveals the presence of smearable alpha contamination in excess of 20 dpm/100 cm that assembly will be isolated and an inspection will be conducted to determine the cause of the contamination.
A gamma survey using a beta-gamma survey instrument will be made on each new fuel shipping container upon receipt. Additionally, each container will be tested for smearable beta-gamma contamination. If any container reveals abnormally high radiation levels or smearable contamination in excess of 1000 dpm/100 cm , the container will be isolated and an investigation initiated.
Calibration of most ranges of the portable gamma and beta-gamma detection instruments will be performed using a shielded calibrator. Each instrument will be calibrated on a six-month basis. The sources used for calibration will be traceable to the National Bureau of Standards or (8540NLU/icv)
other standards laboratory. Additional alpha, beta, and gamma sources will be used as necessary to calibrate or check the lower ranges of the various instruments. A calibration or check source will be used to verify instrument operability. This verification will be performed as a minimum on a daily basis prior to use.
Programs will be developed to meet applicable sections of 10 CFR 20.
These programs such as Radiation Control and Protection Program, ALARA Program, and Respiratory Protection Program, will establish the SHNPP radiation protection philosophy and policy. Detailed procedures implementing the overall programs will be developed and used as day-to-day guidance in personnel radiation protection and radioactive material security and accountability. Equipment to be used in conjunction with programs and procedures relative to storage of unirradiated fuel are listed below:
Devices such as thermal luminescent dosimeters (TLD) and self-reading pocket dosimeters (SPRD) will be used to monitor personnel. SPRD's will be read at SHNPP. TLD's will be read at the Harris Energy and Environmental Center (HE&EC) using an automated TLD system and Records Management system.
Low background gas flow proportional counters used for gross alpha and gross beta measurements.
A beta-gamma counter-sealer used for gross beta-gamma measurements.
GM beta-gamma survey meters (most sensitive range 0-.2 mR/hr.,
maximum range 0-2 R/hr., with internal probe) used for detection of radioactive contamination on surfaces and for low-level exposure rate measurements.
Ionization chamber beta-gamma survey meters 0-5 rem/hr.
(0-5 mrem/hr. most sensitive range) used to cover the general range of dose rate measurements necessary for radiation protection evaluations.
Remote monitoring (telescoping probe) GM tube-beta, gamma survey meters, 0-1,000 R/hr., 0-2 mR/hr., most sensitive range used for exposure rate measurements.
Alpha scintillation survey meters, 0-500K cpm used for measurement of alpha surface contamination.
Small amounts of radioactive waste, that may be generated from tests used to determine the presence of radioactive contamination will be collected, properly packaged, and temporarily stored in a controlled area in the Radiation Control facility located in the Waste Processing Building until it is shipped to a licensed radioactive waste disposal facility for burial. When construction has been completed, radio-active waste generated from these tests will be stored in the designed radioactive waste storage area of the Waste Processirg Building until it is shipped to a licensed radioactive waste disposal facility.
(8540NLU/1cv)
2.2 Nuclear Critically Safety The movement and inspection of new fuel assemblies will be a team effort. The team leader will be responsible for critically safety and direction of new fuel handling and will meet one of the following minimum qualifications:
a) Previously held an NRC SRO license at another facility and has had direct fueling handling experience.
b) Individual(s) has been in responsible position in charge of new fuel receipt and handling at another facility.
The individual responsible for critically safety will be trained on SHNPP fuel handling procedures.
The team will also include an individual qualified as a radiation control technician. This individual will be responsible for:
a) Survey of the transport vehicle prior to removal of new fuel shipping containers and periodic surveys prior to and during the removal, inspection and movement of the new fuel.
b) Establishing a radiation control area surrounding the fuel inspection area.
c) Dosimetry for the work party doing inspection and fuel movement to the storage area.
d) Operation of a low volume airborne radiation monitor during load inspection and movement.
e) Smears of new fuel assemblies to determine the presence and quantity of tramp uranium.
The team will also include additional personnel as required to complete fuel inspection and operate cranes and other equipment.
A general outline of the tasks to be completed during the inspection and movement of the first core of new fuel assemblies is provided below:
a) Radiation survey of new fuel shipping containers prior to movement.
b) Lifting the new fuel shipping containers from the carrier by t'e auxiliary crane and placement of the containers in the new fuel inspection area.
c) Unbol'ting, removal, and inspection of the new fuel container lids.
d) Upending the fuel assemblies.
(854ONLU/acv)
e) Inspecting assemblies.
f) After inspection, transporting acceptable new fuel assemblies to the new fuel storage pool.
g) Unacceptable assemblies may be temporarily stored dry in the new fuel pool or returned to the shipping containers and stored in the new fuel inspection area.
New fuel=assemblies will be removed from the shipping containers one at a time and placed into the new fuel storage racks. The new fuel racks consist of individual vertical cells fastened together through top and bottom supporting grid structures to form integral free-standing modules with a center-to-center spacing of 10.5 inches. This arrangement prevents the insertion of assemblies into areas not designated to hold fuel. A neutron-absorbing material is encapsulated into the stainless steel walls of each stored cell. The fuel rack structure is illustrated in Attachment 5.
Fuel assemblies may be removed from storage for purposes of inspection, and for initial core loading. All fuel handling activities will follow approved fuel handling procedures. If fuel assemblies are to be removed from storage for inspection, only one assembly will be allowed out of the storage racks at one time. The handling of more than one fuel assembly will not be permitted at any time.
Critically of fuel assemblies in the fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between assemblies and inserting neutron-absorber material (Boraflex) between assemblies. The following assumptions were made in performing the critically analysis:
a) The fuel contains the maximum design enrichment (3.9 w/o U-235) without any control rods or any noncontained burnable poison.
b) Since the design of the new fuel pools permits flooding, the assumed moderator is pure water at the temperature within the design limits of the pool which yields the largest reactivity.
Since the presence of poison plates removes the conditions necessary for optimum moderation, a conservative value of 1.0 gm/cm is used for the density of water. No dissolved boron is included in the water.
c) The array is infinite in both lateral and axial dimensions.
d) Uncertainties and biases due to mechanical tolerances during construction are treated by using the most conservative values.
e) Credit is taken for the'eutron absorption in full-length structural materials (Type 304 stainless steel) and in solid materials (Boraflex) added specifically for neutron absorption. A minimum poison loading is assumed in the poison plates and B4C particle self-shielding is included as the bias in the reactivity calculation.
(8540NLU/Icv)
The calculation method and cross-section values are verified by comparison with critical data for assemblies similar to those for which the racks are designed. This benchmarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low moderator densities.
The design method which ensures the critically safety of fuel assemblies in the fuel storage rack uses tge AMPX system of codes(
for cross-section generation and KENO IV ~ for reactivity determination.
The 218 energy group cross-section library ) that is the common starting point for all cross-sections used for the benchmarks and ghq storage rack is generated from ENDF/B-IV data. The NITS, program<
includes, in this library,. the self-shielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatjal weighting of cross-sections is performed by the XSDRNPM program<2) which is a one-dimensional SN transport theory code. These multi-group cross-section sets are then used as input to KENO IV( ) which is a three-dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 27 critical experiments have been analyzed using the above method to demonstrate its applicability to critically analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (boral, steel water) that simulate LWR fuel shipping and storage conditions(4~ ) to dry, harder spec)rum uranium metal cylinder arrays with various interspersed materials< ) (plexiglass, steel and air) that demonstrates the wide range of applicability of the method.
The average K of the benchmarks is 0.9998 ff associated which demonstrates that there is no blas with the method. The standard deviation of the Keff values is 0 0057 b, k. The 95/95 one sided tolerance limit factor for 27 values is 2.26. Thus, there is a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.013 A k.
The total uncertainty to be added to a critically calculation is:
e (KS) th d s 0.013 as discussed above, and (KS) < 1 is the statisticaT uncertainty associated with the particular FEHO calculation being used.
The most important effect on reactivity of the mechanical tolerances is the possible reduction in the water gap between the poison plates. The worst combination of mechanical tolerances are those that result in the maximum reduction in the water gap. For a single can it is found that reactivity does not increase significantly because the increase in reactivity due to the water gap reduction on one side of (8540NLU/lcv)
the can is offset by the decrease in reactivity due to the increased water gap on the opposite side of this can. The analysis, for the effect of mechanical tolerances, however, assumed a worst case of a rack composed of an array of groups of four cans with the minimum water gap between the four cans. The reactivity increase of this configuration is included as a bias in calculating the Keff rack. It is included as a bias term since cans can be weHed to a common grid during manufacturing which is the likely cause of the water gap reduction.
The final result of the uncertainty analysis is that the critically design criteria are met when the calculated effective multiplication factor, plus the total uncertainty (TU) and any biases, is less than 0.95.
These methods conform with ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurizer Water Reactor Plants," Section 5.7, Fuel Handling Systems; ANSI N210-1976, "Design Objectives for LWR Spent Fuel Storage Faci.lities at Nuclear Power Stations," Section 5.1.12; ANSI N16.9-1975, "Validation of Calculational Hethods for Nuclear Critically Safety," NRC Standard Review Plan, Section 9.1.2, "Spent Fuel Storage;"
and the NRC guidance, "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications."
Some mechanical tolerances are not included in the analysis because worst case assumptions are used in the nominal case analysis. An example of this is eccentric assembly position. Calculations were performed which show that the most reactive condition is the assembly centered in the can which is assumed in the nominal case. Another example is the reduced width of the poison plates. No bias is included here since the nominal KENO case models the reduced width explicitly.
For normal operation and using the method described above, the Keff for the rack is determined in the following manner:
eff Knominal mech method part nominal method where Knomina 1 nomina l case KENO Kef f B
mechh
= K ff bias to account for the fact that mechanical eff toferances can result in water gaps between poison plates less than nominal.
B method h d
= Nethod bias determined from benchmark critical comparisons.
B part = Bias to account for poison particle self-shielding.
KS nominal =
inal 95/95 uncertainty in the method bias.
Substituting calculated values, the result is Keff ~ 0.9242.
(8540NLU/lcv)
Since Keff I
is less than 0.95 including uncertainties at a 95/95 probability/confidence level, the acceptance criteria for critically is met.
The neutron absorber material (Boraflex) is a flexible material tested to withstand radiation exposure and hot borated water temperatures up to 240'F. Boraflex has a silicon rubber base, which serves as a binder for up to 49 w/o B C. Specifications call for 0.02 grams of B-10 per square centimeter of oraflex. It does outgas helium and must be vented to prevent rack swelling.
New fuel will be removed from rack locations from time to time as required for relocation or inspection. All fuel moves will be performed according to fuel vendor specifications to assure mechanical integrity of the fuel. In order to maintain a sub-critical configuration during all fuel-handling conditions, only one assembly will be removed from storage at any one time.
The applicants request exemption from the requirements of 10 CFR 70.24 as provided in Subsection 70.24(d). As described in this Section, the fuel assemblies will be stored in critically safe storage racks. In addition, other admi.nistrative procedures as discussed herein preclude the achievement of conditions which would cause critically.
REFERENCES
- 1. W. E. Ford,,III, et al, "A 218-Group Neutron Cross-Section Library in the AHPX Haster Interface Format for Critically Safety Studies,"ORNL/CSD/TH-4(July 1976).
2, N. H. Greene, et al, "AHPX: A Modular Code System for Generating Coupled Hultigroup Neutron-Gamma Libraries from ENDF/B," ORNL/TH-3706 (March 1976)
- 3. L. H. Petrie and N. F. Cross, "KENO IV An Improved Honte Carlo Criticality Program," ORNL-4938 (November 1975).
- 4. S. R. Bierman, et al, "Critical Separation Between Subcritical Clusters of 2.35 wt % U Enriched U02 Rods in Water with Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2438 (October 1977).
- 5. S. R. 11irman et al,, Critic"al Separation Between Suberftleal Clusters of 4.29 wt % U Rods in Water with Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2615 (March 1978).
- 6. J. T. Thomas, "Critical Three-Dimensional Arrays of U (93.2) Metal Cylinders,"
Nuclear Science and Engineering, Volume 52, pages 350-359 (1973).
(8540NLU/lcv)
2.3 Accident Analysis The design basis fuel handling accident is the postulated drop of an irradiated fuel assembly resulting in the rupture of the cladding of all the fuel rods in the assembly. This postulated accident is discussed in more detail in FSAR Section 15.7.4 (Attachment 6). The possibility of a fuel handling accident is remote because of the many interlocks, administrative controls and physical limitations imposed on the fuel handling operations.
Even if a fuel handling accident did occur, the consequences would be relatively insignificant, as compared to those given in the FSAR for spent fuel, since all the new fuel will be unirradiated.
discussed above, fuel storage facilities are designed to be criticality 4
As safe even for postulated optimum moderator scenarios. In addition, handling of fuel assemblies prior to placement of an assembly in the fuel storage racks will be conducted such that there is no potential for accident criticality during these activities.
Administrative controls, physical limitations on the fuel handling operations and experience and training of personnel performing fuel movement assure that damage to the new fuel assemblies will not occur. In the unlikely event that a fuel assembly is dropped, the health physics technician monitoring fuel handling activities will be responsible for performing activities to assess the presence and extent of the spread of radioactive materials from the dropped fuel assembly. The requisite equipment for this will be maintained in the FHB during fuel inspection and fuel movement to the storage area.
3.0 OTHER MATERIALS RE UIRING NRC LICENSE Other SNM for which a license is requested consists of U-235 and U-238 in the following forms and quantities.
a) Incore Fission Chambers Form Amount*
U-234 U-234, 3/4 of 1%
- .0041 gm of'ranium enriched to greater than 90% in U-235 per fission chamber.
The five fission chambers are manufactured by Westinghouse, Model WL-23435, for core power distribution mapping.
b) Sealed sources for irradiation surveillance capsules:
Form Amount Np-237 16-18 mg. (Np02 19-21 gm)
U-238 12 mg. (U308 14.25 gm)
-1 7- (8540NLU/1cv)
Sealed sources manufactured by Westinghouse/Amersham (Np-Amersham, U-Westinghouse). Sealed source enclosed in a stainless steel capsule,
.250" 0.D. x .375" long, sealed in a steel block, and enclosed in a stainless steel capsule. There are in each surveillance capsule 1 sealed source of Np-237 and 1 sealed source of U-238 (total of 6 capsules). There are approximately 24 x 10 microcuries of U-238 for 6 capsules and 72.6 microcuries of Np-237 for 6 capsules.
The -incore instrumentation fission chambers will be used in .the SHNPP incore instrumentation system to provide information on core neutron flux distribution. Prior to installation, the fission chambers will be stored in locked cabinets under the jurisdiction of the Radiation Control group. The radiation protection provisions for these sources will be as described in Section 2 of this application.
(8540NLU/1 cv)
SHNPP FSAR h
tachment (2) 9+1 ~ 4 FUEL HANDLING SYSTEM There is one Fuel Handling Building which serves the two Units. Contained io-inside this building are two new fuel pools and two spent fuel pools with a f transfer canal system that permits transfer of fuel between pools. The Fuel Handling System (FHS) is designed in conformance with Regulatory Guide 1.13 as detailed, in Section 1.8.
The Fuel Handling System will provide the following services on SHNPP:
. a) provides the means for safely moving the fuel as necessary to accomplish, receipt and storage of new and spent fuel, 'refueling,, receiving shipments of offsite spent fuel, and shipment of spent fuel to- offsite locations.
b) provides the means for safely preparing the plant facilities for fuel movement, such as placement of fuel transfer canal gates in appropriate positions, dismantling and replacing reactor vessel components to allow for refueling and placement of portable barriers for safe fuel cask handling.
c) provides the means for safely transferring spent fuel among all fuel pools.
E d) provides shielding for protection of personnel from excessive radiation exposure during refueling, inspection, and fuel storage.
e) provides that either:
- 1) a load drop resulting from a single electrical or lifting cable failure is precluded, or;
- 2) the consequences of a load drop can be accommodated without affecting the ability to bring the plant to a safe shutdown condition or to control the release of significant amounts of radioactive material.
h f) is designed such that maximum 'design load on the wire rope hoisting cables shall not exceed 1/5 ultimate strength of the cables.
g) provides appropriate containment isolation boundaries for containment penetrations.
h) is designed such that lifting devices have appropriate interlocks and stopping capability.
i) is designed such that fuel lifting and handling equipment and structures will not fail in such a manner as to damage Seismic Category I equipment or structures in the event. of an SSE.
Structures, systems, and components designed as Seismic Category I are shown in Table 3.2.1-1 ~ Structures, systems, and components which could damage
- 9. 1;4-1 Amendment No. 10
~
SHNPP FSAR ~
safety-related equipment upon failure are designed to withstand an SSE event without causing such damage. Components that are designed for Safety Class 1, 2, or 3 are shown in Table 3 '.1-1.
9 ' ~ 4.1 ~ 1 Fuel Transfer Decay Heat
'l The refueling water provides a reliable and adequate cooling medium 'for spent fuel transfer. The'ew and spent fuel pools are connected to the pool cooling and clean-up systems, which are discussed in detail in Section 9.1.3 9 ' '.1 ~ 2 Fuel Transfer Radiation Shielding I
Adequate shielding from radiation is provided during reactor refueling by transferring and storing .spent fuel underwater and maintaining a safe shielding depth of water above the 'fuel assemblies during refu'cling. TMs permits visual control of the operation at all times while maintaining acceptable radiation'levels for periodic occupancy of the area by operating personnel.
9.i+4 ' S stem Descri tion 9 1 4 ~ 2o1 System The Fuel Handling System consists of the equipment and associated structures used to handle fuel from the time of receipt until it leaves the plant and the handling equipment used to prepare the reactor to discharge and receive fuel.
The equipment consists of.:
a) Containment building circular bridge crane.
b) Manipulator crane.
'I c) Spent 'fuel bridge crane.
Spent fuel cask handling crane.
e) Auxiliary crane.
f) Fuel handling tools -and fixtures g) Fuel transfer system.
h) . Fuel racks.
i) New fuel elevator.
The following areas are associated with the fuel handling equipment:
a) Refueling cavity.
b) Spent fuel pools and fuel transfer canal syst'm.
- 9. 1. 4-2
SHNPP FSAR c) New fuel storage areas.
d) Spent fuel cask loading pool and decontamination area.
Refer to Figures 1.2.2-55 through 1.2.2-59 for general arrangements of the Fuel Handling Building.
For each Unit, the associated fuel handling structures may be divided into two areas:
a) , The refueling cavity which is flooded only during Unit shutdown for l
refuel pg.
b): The. new and spent fuel- pools and fuel transfer canal system.
9.1.4.2.2 Components 9.1.4 2.2.1 Reactor Vessel Head Lifting Device reactor vessel head lifting device consists of a welded and bolted The structural steel frame which enables the crane operator to lift the head and store it during refueling operations This device is part of the Integrated Reactor Vessel Head (IRVH) ~
9.1.4.2.2.2 Reactor Internals Lifting Device" J
The reactor internals lifting device -is a structural frame mechanism which provides the means of gripping the upper and lower internal packages to transmit the lifting load to the cran'e (refer to Figure 9.1.4-1)'y the use of auxiliary brackets, the assembly is guided onto the internal packages.
Attachment is accomplished by manually connecting the assembly to the
"'internals'i'th handling tools operated from the internals lifting rig platform. The upper internals are stored in the flooded refueling cavity during refueling. Although their removal is not required for refueling, the lower internals may be stored in the flooded refueling cavity when required 9.1.4".2.2.3 , Manipulator. Crane The manipulator crane is a rectilinear bridge and trolley crane with a vertical mast extending down into the refueling water (refer to Figure 9.1 ~ 4-2) The bridge spans the refueling cavity and runs on rails set into the floor along the edge of the refueling cavity. The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core. A long tube with a pneumatic gripper on 'the end is lowered out of the mast to grip the fuel assembly. The gripper tube is long enough so the upper end is still contained in the mast when the gripper end contacts the fuel. A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube, The fuel is transported while inside the mast tube to its new posi tione Controls for the manipulator crane are mounted on a console on the trolley.
The bridge is positioned on a coordinate system laid out on one rail. The
- 9. l. 4-3
SHNPP FSAR electrical readout system on the console indicates the position of the bridge.
The trolley is positioned with the aid of a scale on the bridge structure.
The scale is read directly by the operator at the console. The drives for the bridge,'rolley, and winch are variable speed and include a separate inching control on the winch.
The manipulator crane will not collapse nor become disengaged as a consequence of an SSE.
9.1.4.2,2.4 Spent Fuel Bridge Crane The spent fuel bridge crane, as shown on Figure 9 '.4-3, is a wheel~ounted walkway spanning the width of the Puel Handling Building, which carries an electric monorail hoist on an overhead structure. The monorail hoist has access to all spent and new fuel pools, as..mell as interconnecting transfer canals. The fuel assemblies are moved within the fuel pools by means of a long-handled tool and short-handled tool (refer to Figures 9.1 ~ 4-4 and 9.1.4-13, respectively) suspended from the hoist. The hoist travel and tool length are interlocked to limit the maximum lift of a fuel assembly to a safe will'ot shielding depth. The spent fuel bridge crane drop its load nor leave the rails as a consequence of an SSE. The capacity of the hoist is two tons;-
the approximate weight of the handling tool and a fuel assembly is 2,000 lbs The capacity of the spent fuel bridge crane precludes excessive loads from being carried over spent fuel storage area.
ii 'I The spent fuel bridge crane hoist which has a design capacity of 2 tons is equipped with a load monitor preset to prevent hoist operation at a load of 250 lbs above the weight of a fuel assembly and handling tools Changing of the set- point to lift heavier loads will be under administrative control. i, 9.1 ~ 4 '.2.5 ', Fuel Transfer System
~ The Fuel Transfer System includes an underwater conveyor car running on tracks extending from the refueling cavity through the transfer tube and into the fuel transfer canal and an upending frame at each end of the transfer tube (refer to Figure 9.1.4-5). To remove a fuel assembly from the reactor, the upending frame in the refueling cavity receives a fuel assembly in, the vertical position from the manipulator crane. The fuel assembly is then lowered to a horizontal position for passage through the transfer tube and raised to a vertical position by the upending frame in the fuel transfer canal. The hoist on the spent fuel bridge then takes the fuel assembly to a position in the spent fuel racks via the fuel transfer canals.
To seal the reactor Containment during Unit operation, a .blind flange is bolted on the end of the transfer tube in the refueling cavity inside Containment, and 'a manually operated valve is locked closed 'in the fuel transfer canal in the Fuel Handling Building (Section 6 '.4) ~ The transfer tube and the blind flange 'are designed to Seismic Category I requirements.
9.1.4.2.2.6 Rod Cluster Control Assembly (RCCA) Changing Pixture
, An RCCA changing fixture is mounted on the refueling cavity wall for transferring RCCA's from fuel assemblies removed from control positions and inserting RCCA's into the fuel assemblies to be placed in the control
SHNPP FSAR positions (Figure 9.1.4-6). The fixture consists of two main components: a guide tube mounted to the wall for containing and guiding the RCCA and a wheel-mounted carriage for holding the fuel assemblies and positioning fuel assemblies under the guide tube. The guide tube contains a pneumatic gripper on a winch which grips the RCCA and lifts it out of the fuel assembly. *By repositioning the carriage, another fuel assembly is brought under the guide tube; and the gripper lowers and releases the RCCA. The manipulator crane loads and removes the fuel assemblies into and out of the carriage.
9.1.4.2.2.7 Spent Fuel Cask Handling Crane and Auxiliary Crane The spent fuel cask handling crane (150-ton) transfers the spent fuel cask between the railroad car and the spent fuel cask loading pool, (refer to Figures 1.2.2-55 through 1.2.2-59). Design of the Fuel Handling Building and the spent fuel cask handling crane prevents the possibility of the cask passing, over or falling into either the spent fuel or the new fuel pools.
Permanent mechanical stops, which will withstand the impact of the crane at maximum operating speed, are provided to limit the crane movement so that travel of the center of the main hook is limited to 12 in. south of the centerline of the cask loading pool. Additionally, only the micro drives will be functional in the last 5 ft. of .crane travel in the southerly direction as controlled by limit switch. In the unlikely event that the crane comes in contact with the mechanical stops while at maximum operating speed, the maximum swing of the bottom of the cask from its normal position in the vertical plane will be 14.5 inches in the southerly direction. When the cask reaches this deflected position, if it is still entirely over the cask pool.
Therefore, dropped while in this extended position, the cask will not come
. in contact with spent fuel in the fuel pools.
The spent fuel cask handling crane is equipped with stops which limit main hook vertical travel to ensure the shipping cask could never fall more than 30 feet through air to any load-bearing surface and will not be raised more than 12 inches above the operating floor.
Two independent systems are provided to prevent the centerline of the main hook of the spent fuel cask crane from coming within 10 ft. 6 in. of the north or west edges of the new fuel pool. A removable barrier is provided with its west face in line with the east edge of the cask unloading pool on top of the dividing wall between that 'pool and the cask head storage area.
The auxiliary crane will',be used for the installation and re-removal of this barrier.
Figures 9.1 4-7 through 9.1.4-12 show the envelope of travel of the main hook of the spent fuel cask handling crane as controlled by design and administrative control of the crane, and within the main hook envelope, the area to which cask travel will be restricted by administrative control.
There is no safety-related equipment within the possible area of main hook (and, therefore, fuel cask) travel', either on the operating floor level or on floors beneath. Nevertheless, the floors of the cask decontamination area, cask head storage area, cask loading pool, railroad car unloading bay, and the 9.1.4-5 Amendment No. 11
SHNPP FSAR operating deck south of column line 73K within the administratively controlled cask travel envelope will withstand a postulated drop of fuel cask from the maximum height. to which it can be raised.
The auxiliary crane (design capacity of. 12 tons) operates on the same runway as the spent fuel cask handling crane as shown in Figures 1.2.2-55 through 1.2.2-59. Two independent systems are provided to prevent the two cranes from coming in contact with each other. Design of each system provides that the auxiliary crane can operate in the common operating area only when the cask crane is in its parking position which is at a safe distance away from the end of travel of the auxiliary crane. While the cask crane is operating, the auxiliary crane is limited to operate at a safe distance away from the common area.
P A redundant supporting system is provided on the auxiliary crane in regard to hook, reeving, and braking mechanisms. Provisions are made to manually move the crane to a laydown area for emergency manual lowering of the load. A detailed description of the auxiliary crane is given in Table 9.1.4-1.
Both cranes are capable of retaining the maximum load during an SSE although the crane may not be operable after the seismic event. The bridge and trolley are provided with means for preventing them from leaving their runways with or without hook load during operation or under any postulated seismic event 5 There is no other lifting device that can carry excessive loads over the fuel storage areas.
9.1.4.2.2.8 Containment Circular Bridge Crane The overhead crane in the Containment (250 ton/50 ton) used for reactor servicing operations is of the polar configuration, and is seated on a girder bracketed off the containment wall. The crane is capable of retaining a 175-ton lifted load (weight of integrated reactor vessel head with lifting rig, which is the heaviest component to be lifted during refueling operations) during an OBE or SSE, although the crane may not be operable after the seismic event. The bridge and trolley are prevented from leaving their runways with or without the 1?5-ton lifted load during operation or under any seismic event.
centerline of the'rane is offset from the reator vessel centerline to
'he assure the, alignment of lifting devices'with all possible loads and to provide clearance for containment spray header piping risers which run vertically along the containment liner.
9.1.4.2.3 Fuel Handling Description (New and Spent)
New fuel assemblies received for initial refueling are removed one at a time from the shipping container and moved to the new fuel assembly inspection After inspection, the acceptable new fuel assemblies are stored in the racks in the new or spent fuel pools.
Should the need exist in the future, spent fuel from other nuclear plants in the CP&L system would be brought to the SHNPP site in an approve approved sshippin ipp ng cask. Th e cask would be placed in the flooded shipping cask pool. The spent 9.1.4-6 Amendment No. 5
SHNPP FSA3 fuel would then be removed from the cask and transported to the storage racks.
This procedure would be carried out with the spent fuel assemblies totally submerged.
The fuel handling equipment handles the spent fuel assemblies underwater from the time they leave the reactor vessel until they are placed in a container
.for shipment from the site. Underwater transfer of spent fuel assemblies provides an effective, economic, and transparent radiation shield, as well as a reliable cooling medium for removal of decay heat.
The associated fuel handling structures are generally divided into two areas:
the refueling cavity which is flooded only during a plant shutdown for refueling, and the fuel pools and fuel transfer canals, which are kept full of water. The refueling cavity and the Fuel Handling Building are connected by -a fuel transfer tube which is, fitted with a blind flange on the Containment end ani: a gate valve on the Fuel Handling-:.Building end. ;The blind flange is in place except during refueling to ensure containment integrity. Fuel is carried through the tube on an underwater transfer car.
\
Fuel is moved between the reactor vessel and the refueling cavity by the manipulator crane. A rod cluster control changing fixture is located in the refueling cavity for" transferring control elements from one fuel assembly to another. The fuel transfer system is used to move fuel assemblies between the Containment Building and the Fuel Handling Building. After a fuel assembly is placed in the fuel upender, the lifting arm pivots the fuel assembly to the horizontal position for passage through the fuel transfer tube. After the transfer car transports the fuel assembly through the transfer tube, the lifting arm at the end of the tube. pivots the assembly to a vertical position so that the assembly can be lifted out of the fuel upender.
In the Fuel Handling Building, fuel assemblies are moved about by the spent fuel bridge crane. When lifting fuel assemblies, the hoist uses a long-handled tool to ensure that sufficient radiation shielding is maintained.
Initially, a short tool is used to handle new fuel assemblies, (see Figure 9.1.4-13) but the bridge crane is used to lower the assembly to a depth at which the fuel handling machine, using the long-handled tool, can place the new fuel assemblies into or out of the fuel upender.
Decay heat; generated by the spent fuel assemblies in the fuel pools, is removed by the Spent Fuel Pool Cooling and Cleanup System, which is described in Section 9.1.3.
After a sufficient decay period, the spent fuel assemblies are removed from the fuel racks and loaded into the spent fuel shipping cask for removal from the site.
9.1.4.2.4 New Fuel Receiving and Inspecting Procedure
'I a) New fuel arrives at the north end of the Fuel Handling Building on truck or, rail car.
b) The airtight door is opened to admit the carrier and closed behind it.
c) The equipment hatch cover is removed.
9.1.4-7 Amendment No. 5
SHNPP 'FSAR d) The new fuel containers are lifted from the carrier by the auxiliary crane and placed in the new fuel inspection area. The equipment hatch cover is replaced.
e) The new fuel container lids are unbolted and removed by the auxiliary crane. and stored.
f) The fuel assemblies are prepared for upending. I g) The fuel assemblies are raised to vertical and lifted with the spent fuel bridge crane to the operating level and inspected.
h) After inspection, acceptable new fuel assemblies are transported to the-appropriate new fuel storage pool or to the new fuel elevator and stored wet in the spent fuel pool.
i) Unacceptable new fuel may be temporarily stored dry in the new fuel pool or returned to the shipping containers and stored in the new fuel inspection area.
j) The new fuel au~iliary crane and container lids are returned to the containers by the bolted.
k)'he equipment hatch cover is dt ~
removed.
- 1) The empty new fuel containers are lifted by the auxiliary crane and loaded back on the carrier. The equipment hatch cover is replaced.
V m) . The airtight door is opened and the carrier, loaded with empty new fuel containers, leaves the building and the airtight door is closed.
Offsite Spent Fuel Receiving Procedure a) Offsite spent fuel arrives at the north end of the Fuel Handling Building in approved shipping containers by truck or railcar.
b) The airtight door is opened to admit the vehicle and closed behind it.
c) The equipment hatch cover is removed.
d) The cask is prepared for lifting.
e) The cask is lifted by spent fuel cask handling crane and transported to the decontamination or work area. The equipment hatch cover is replaced.
f) Cask is, prepared for pool entry.
g) Cask loading pool is flooded and the gate removed.
h) Removable barrier is put in place by the FHB auxiliary crane.
i) Cask is transported and lowered into the cask loading pool.
j) Cask head is removed and stored in an appropriate location.
9.1.4-8 hmo~8mont Nn .
0 SHNPP FSAR k) Fuel is removed from the cask using the appropriate long-handled spent fuel tool (PWR or BWR).
- 1) The spent fuel is transported, using the spent fuel bridge crane, to its pre-assigned storage location.
m) After the cask is emptied the head is returned to the cask and replaced.
n) Cask is lifted by the spent fuel cask handling crane and placed in the decontamination area.
o) Removable barrier is placed in storage.
p) Cask is prepared for shipment and decontaminated to acceptable levels.
't')
The equipment hatch cover is removed.
r) Cask is lifted from the decontamination area and returned to the truck or railcar for removal and the equipment hatch cover is replaced.
s) The airtight door is opened and the vehicle, loaded with the empty cask, leaves the building and the airtight door is closed.
9.1-4.2.6 Spent Fuel Shipping Procedure a) A truck or railcar arrives at the north end of the Fuel Handling Building carrying an empty approved spent fuel shipping container.
b) The airtight door is opened to admit the vehicle and closed behind it.
c) The equipment hatch cover is removed.
d) The cask is prepared for lifting.
e) The cask is lifted by spent fuel cask handling crane and transported to the decontamination or work area. The equipment hatch cover is replaced.
f) Cask is prepared for pool entry.
g) Cask loading pool is flooded and the gate removed.
h) Removable carrier is put in place by the FHB auxiliary crane.
i) Cask is transported and lowered into the cask loading pool.
g ) Cask head is removed and stored in an appropriate location.
k) Spent fuel is loaded into the cask using .the appropriate long-handled spent fuel tool (PWR or BWR) and the spent fuel bridge crane.
After cask is loaded, the head is returned to the cask and replaced.
m) Cask is lifted by the spent fuel cask handling crane and placed in the decontamination area.
9.1.4-9 Amendment No. 5
SHNPP. FSAR ~;
Removable barrier is placed in storage.
o) Cask is prepared for shipment and decontaminated.
p) The equipment hatch cover is removed.
q) Cask is lifted from the decontamination area and returned to the truck or railcar fox removal; and the equipment hatch cover is replaced.
r) The a'irtight door is opened and the vehicle, loaded with the full cask, leaves the building and the airtight door is rlosed.
9.1.4.2.7 Refueling Procedure I
'4 I
- 9. 1 4:2. 7 1 Preparation a) The reactor is shut down and cooled to ambient conditions with a final K
ff ( 0 95 and'if the levels are sufficiently low, the
')
A radiation survey is made containment fs entered.
c) , The reactor vessel coolant level is lowered slightly below the reactor vessel flange.
d) IRVH cables are disconnected and removed to storage.
t tJ, e) Reactor vessel head insulation.and instrument leads are xemoved.
f) . The reactor vessel head nuts are loosened with the hydraulic tensioner.
g) . The reactor vessel hea'd studs and nuts are removed to storage.
h) The refueling cavity drain valves are closed, and the fuel transfer tube blind flange is removed.
i) Checkout of the fuel transfer device and manipulator crane is started.
installed in three vessel flange holes, I')
Guide studs are and the remainder of the holes are plugged.
k) The reactor vessel cavity seal ring is installed.
l) Final preparation of underwater lights and tools is made. Checkout of fuel transfer system is completed. Manipulator crane is parked.
/
m) The reactor vessel head is unseated and raised one foot.
n) The refueling cavity is filled to the level of the vessel flange with water from the refueling water storage tank.
o) The reactor vessel integrated head is lifted slowly while the water is pumped into the refueling cavity by the residual heat removal pumps from the refueling water storage tank through the reactor vessel. The water level and
)9. 1. 4-10 hmoeAmont Mn
SHNPP FSAR vessel head are raised, and the water level is maintained just below the head until the water level is several feet above the vessel flange.')
The refueling cavity isw~
filled.
q) The reactor vessel integrated head is taken to the storage pedestal.
r)'he control rod drive shafts are unlatched from the spider.
s) The reactor vessel internals lifting rig is lowered into position and latched to'. the upper internals package.
I t). The reactor vessel upper internals package and drive shafts are lifted'-
out of the vessel and, placed in,. the underwater storage rack.
u) Checkout of manipulator crane is complete (Core Index).
v) The core is now ready for refueling.
9.1.4.2.7.2 Refueling Sequence The refueling sequence is now started with the manipulator crane. As determined in the Nuclear Design Report which is prepared before each refueling, spent fuel assemblies are removed from the 'core; some partially spent fuel assemblies have their positions changed; and new assemblies are added to the core..
For fuel assemblies containing rod'cluster control assemblies (RCCA), the
.re'fueling "sequence is modified as required. If a transfer of the RCCA between fuel assemblies is necessary, the assemblies are taken to the RCCA changing fixture for the exchange. Such an exchange may be required whenever a spent fuel assembly containing an RCCA is removed from the core and whenever a fuel assembly is placed in or taken out of a control position during refueling rearrangement.
- 9. 1. 4. 2. 7. 3 Reactor Reassembly a) The:- fuel transfer conveyor car is parked, and the refueling cavity is isolated, from the fuel transfer canal by closing the manual gate valve on the FHB side.
b). The manipulator crane is parked.
c) The reactor vessel upper internals package is placed in the vessel.
Th'e reactor vessel internals lifting 'rig is unlatched and removed to storage.
d) The full-length control rod drive shafts are relatched to the RCCA spiders.
e) The old seal rings are removed from the reactor vessel head, the grooves cleaned, and new rings installed.
f) The reactor vessel head is picked up and positioned over the reactor vehsel.
9.1.4-.11 Amendment No. 5
SHNPP FSAR g) The reactor vessel head and water level in the refueling cavity are lowered simultaneously.
h) When the cavity water level is just above the flange, the reactor vessel head is lowered to about one foot above the vessel flange and the refueling cavity is drained and the flange surface is cleaned.
i) The reactor vessel head is seated.
j) The guide studs are removed to their storage rack. The stud hole plugs are removed.
k) The head studs are placed and retorqued.
- 1) The refueling cav'ity drain ho3,es are'pened, and the flange for the fuel transfer tube is replaced.
m) . 'lectrical leads are reconnected to the IRVH.
n) Vessel head insulation and instrumentation leads are replaced.
o) The cavity seal ring is removed.
ghq p) A hydrostatic test is performed on the reactor vessel.
q) Control rod drives are checked.
r) Pre-start-up tests are performed.
9.1.4.2.8 , Codes and Standards
,r,
'a) Cranes Crane Manufacturers Association of America (CMAA)
Specification No. 70 and/or AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings.
b) Structures r
ASME Code, Section III, Appendix XVII.
c) Electrical Applicable standards and requirements of the National Electric Code, NFPA 70, and NEMA standards MAI and ICS for design installation and manufacturing.
d) Materials Main load-bearing materials to conform to the specifications of the ASTM, ASME, or AISC Standards.
e) i Safety OSHA Standards, load-testing requirements.
20CFR1910 and 10CFR1926, including ii ANSI N18.2 iii Regulatory Guide 1.29 and GDC 61 and 62.
iv ANSI B30.2, "Safety Standards for Overhead and Gantry Cranes."
Fuel Transfer Tube: ASME Section III, Code Class 2.
Amendment No. 5
SHNPP FSAR
- 9. l. 4 3 Saf et Evaluation The extent of compliance of the fuel handling system with Regulatory Guide 1.13 is discussed in Section 1.8.
9.1.4 3 1 . Fuel Handling Equipment.
Electrical interlocks and limit switches on the bridge and trolley drives of the manipulator crane protect the equipment. In an emergency, the bridge, trolley, and winch, can be operated manually using a hand-wheel on the motor shaft+
The manipulator crane design includes the following provisions to ensure safe handling of fuel assemblies:
Safety Interlocks Operations which could endanger the operator or damage the fuel are prohibited by mechanical or fail-safe electrical interlocks or by redundant electrical interlocks. All other interlocks are intended to provide equipment protection and may be implemented either mechanically or by electrical interlock, not necessarily fail-safe FaQ.-safe electrical design of a control system interlock may be applied according to, the following rules')
Fail-safe operation of. an electrically operated brake is such that the brake engages on loss of power.
b) Pail-safe operation of an electrically operated clutch is such that the clutch disengages on loss of power c) Pail-safe operation of a relay is such that the de-energized state of the relay inhibits unsafe operation..
d) Fail-safe operation of a switch, termination, or wire is such that breakage or high resistance of the circuit inhibits unsafe operation. The dominant failure mode of the mechanical operation of a cam-operated limit switch is sticking of the plunger in its depressed position. Therefore, use of the plunger-extended position (on the lower part of the operating cam) to energize a relay is consistent with fall-safe operation.
e) Pail-safe operation of an electrical comparator or impedance bridge is not defined.
Those parts of a control system interlock required to be fail-safe which are not or cannot be operated in a fail-safe mode as defined in these rules, may be supplemented by a redundant component or components Co provide the requisite protection i')
When the gripper is engaged, the machine shall .not traverse unless the guide tube is in its full up position.
9.1.4-13
SHNPP FSAR b) When the gripper is disengaged, the machine shall not traverse unless the gripper is withdrawn into the mast.
c) VerticaL motion of the guide tube shall be permitted only in a controlled area over the reactor (avoiding the vessel guide studs), fuel transfer system,, or rod cluster control changing .fixture.
d) Traverse of the trolley and bridge shall be limited to the areas of item c and a clear path connecting those areas e) A key-operated interlock bypass switch shall be provided to defeat interlocks a through d to allow operation of an inspection camera on the gripper. 1
"~
f) The gripper shall be monitored'y limit switches to confirm operation to the fully engaged ov fully disengaged position. An audible and a visual alarm shaLL be actuated if both engaged and disengaged switches are actuated at the same time or if neither is actuated. A time delay may be used to allow for recycle time of normal operation.
g) The loaded fuel gripper shall not reLease unless it is in its down position in the core,, or in the fuel transfer system or rod cluster control changing fixture, and the weight of the fuel is off the mast.
r h) Raising of the guide tube shaLl not be permitted if the gripper is disengaged and 'the load monitor indicates Chat fuel assembly .
it still is attached to the 2p i) exceeds Raising of the guide tube shall not be permitted if the hoist loading Che allowable limit set in accordance with the Westinghouse Specification F-5 "Instructions, Precautions, and Limitations for Handling New and Partially Spent Fuel Assemblies "
))
in the Lowering of the guide tube hoist.
shall not be permitted if slack cable exists k) 'he 'guide tube shaLL be prevented from rising to a height where there is less than the'safe shielding depth of water over the fuel assemblies
- 1) The guide tube shall travel only at-a controlled speed of about 2 fpm when: l) the,bottom of the fuel begins to enter the core, and 2) the gripper approaches .the top of the core. In addition, )ust above those points, the guide tube shall automatically stop lowering, and shall require acknowledgement from the operator before proceeding.
m) 'The fuel transfer system upender shall be prevented from moving unless the engaged gripper is in the full up position or the disengaged gripper is withdrawn into the mast, or unless the manipulator crane is out of the fuel transfer zone. An interlock shall be provided from the refueling machine to
'the fuel transfer system to accomplish this.
Suitable restraints are provided between the bridge and trolley structures and their respective rails Co prevent derailing due to the SSE. The manipulator
- 9. l. 4-14
SHNPP FSAR I
crane prevents disengagement of a fuel assembly from the gripper during an SSE The following safety features are provided for in the fuel transfer system:
a) Transfer car permissive switch - The transfer car controls are. located in the Fuel Handling Building; and conditions in the Containment are, therefore, not visible to the operator The transfer car permissive switch allows a second operator in the Containment to exercise some control over car movement if conditions visible to him warrant such control.
C
)
Transfer car operation is possible only when both lifting arms are in pe down-position as indicated by the limit switches. The permissive switch is a backup for the transfer car lifting arm interlock. ~suming.the fuel is in the upright position in the Containment and the lifting arm 'ontainer interlock circuit fails in the permissive condition, the operator in the Fuel Handling Building still cannot operate the car because of the permissive switch interlock. The interlock, therefore, can withstand a single failure.
b) Lifting.arm (transfer car position) Two redundant interlocks allow lifting arm operation only when the transfer car is at the respective end of its travel and therefore can withstand a single failure.
Of the two redundant interlocks which allow lifting arm operation only when the transfer car is at the end of its travel, one interlock is a position limit switch in the control circuit.. The backup interlock is a mechanical latch device on the lifting arm that is opened by the,car moving into position c) '" 'ransfer car (valve open) - An interlock on the transfer tube valve-permits transfer car operation only when the transfer tube valve position switch indicates the valve is fully open.
d) Transfer car (lifting arm) The transfer car lifting arm is primarily designed to protect the equipment from overload and possible damage if a~
attempt is made to move the car when the fuel upender is in the vertical position. This interlock is redundant,and can withstand a single failure.
- The basic interlock is a position limit switch in the contxol Circuit~ The backup interlock is a mechanical latch device that is opened by the weight of the fuel upender when in the horizontal position.
e) Lifting arm (refueling machine) - The refueling canal lifting arm is interlocked with the manipulator crane. Whenever the transfer car is located in the refueling cavity, the lifting arm cannot be operated unless the mast is in the fully retracted position or the manipulator crane is over the core or the gripper is released and inside the core.
f) Lifting arm (fuel handling machine) - The lifting arm is interlocked with the spent fuel bridge crane. The lifting arm cannot be operated unless the spent fuel bridge crane is not over the lifting arm area.
9 ~ 1 ~ 4-15
SHNPP FSAR
- 9. 1.4.3, 2 Overhead Cranes Overhead cranes used in refueling and fuel handling operations include the 250/50-ton circular bridge crane, the 150-ton spent fuel cask handling crane, and the 12-ton (design Load) auxiliary crane. These cranes are classified as non-nuclear safety (NNS) since they neither, provide nor support any safety system function.,
a) , Circular Bridge Crane The crane is used for removal of the reactor vessel integrated head and the upper internals package during the refueling shutdown. This crane is provided with seismic restraints to prevent derailment in the event of an SSE or OBE.
~t b) Spent Fuel Cask Handling Crane, This 150-ton crane is provided for handling the spent fuel shipping cask.
Crane design and building arrangement preclude travel of this crane hook over the fuel pools. This crane will maintain its structural integrity and hold its load under the dynamic loading conditions of the SSE as described in Section F 1~4 2i2i.7 ' A postulated drop of the fuel cask will. not cause damage to spent fuel and safety-related equipment c) 'uxiliary Crane The auxiliary crane is used for handling of the removable barrier, pool gates, fuel racks and other miscellaneous items weighing less than 10 tons
~-
The crane is designed to maintain'ts'tructural integrity and hold its load-under- the dynamic loading conditions of the SSE Load drop is precluded due to its redundant supporting system as described in Section 9,1.4.2 2.7 and Table 9.1.4-1 ~
9 1~4 4 Ens ection and Testin Re uirements As part of normal plant operations, the fuel-handling equipment is inspected prior to the refueling operations During the operational testing, procedures are followed to affirm the correct performance of the fuel handling system interlocks The test and inspection requirement fo'r the equipment in the fuel handling system are:
a) Manipulator crane, spent fuel bridge crane, rod cluster control changing fixture, and new fuel elevator.
The minimum acceptable test shall include the following:
P
- 1) Hoists and cable shall be load tested at 125 percent of the rated load.
- 2) The equipment shall be assembled and checked for proper functional and running operation.
9.1 '-16
SHNPP FSAR The following maintenance and checkout tests are recommended to be performed prior to using the equipment: .
- 1) Visually inspect for loose or foreign parts. Keep free of dirt and grease.
'2) Lubricate exposed gears with proper lubricant.
- 3) Inspect hoist cables for worn or broken strands.
- 4) Visually inspect all limit switches and limit switch actuators for any sign of damaged or broken parts
- 5) Check the equipment for proper functional and running ",.
aperation.
b) Reactor vessel head lifting device and reactor internals lifting devicei The minimum acceptable test at the shop site shall include the following:
') The 1) load.
The devices The devices following maintenance prior to using the tools:
. 1) =
shall shall
~
be load tested to 125 percent of the rated be assembled and checkout to ensure proper component tests are recommended Visually inspect, for loose or foreign parts or damaged surfaces.
to be performed fit up ,
- 2) Visually inspect all engagement surfaces and lubricate with proper lubricant.
- 3) On the reactor internals lifting device, check for the proper functioning of the engagement and protective rig operators.
c) New fuel assembly handling tool 'and spent fuel assembly handling tool The minimum acceptable test at the shop site shall include the following:')
The tools shall be load tested to 125 percent of the rated load.
- 2) The tools shall be assembled and checked for proper functio'nal operation.
The following maintenance and checkout tests are recommended to be performed prior to using the tools.
- 1) Visually inspect the tools for dirt, loose hardware, and for any signs of damage such as nicks and burns.
F 1.4-17
SHNPP FSAR
- 2) Check the tools for proper functional operation.
d) Fuel transfer system The minimum acceptable test at the shop site shall. include the following:
- 1) The system shall be assembled and checked for propex functional and running operation.
The following maintenance and checkout tests are recommended to be perfoxmed prior to using the tools.
- 1) Visually inspect for loose c, foreign parts. Keep free of dirt and grease.
M
- 2) Lubricate exposed gears with proper lubricant.
- 3) Visually inspect all limit switches and limit switch actuators for any signs of damaged or broken parts.
- 4) Check the system for proper functional and running operation.
e) Reactor vessel stud tensioner The minimum acceptable test at the shop site shall include the following:
I) The tensioner shall be as'sembled and checked for proper functional and running operation The following maintenance and checkout tests are recommended to be performed prior to using the equipment Visually inspect for loose or foreign parts.
2)
Inspect hydraulic lines for w~e r or. damage.
- 3) Check the hydraulic unit for proper pressurization and if any leaks occur at operating .pressure.
9 ' ~ 4.5 Instrumentation Re uirements Instrumentation requirements of equipment, including interlocks, are discussed in Sections 9.1.4.2 and 9.1 4.3
- 9. 1~ 4-18
SHNPP FSAR TABLE 9. 1.4-1 FHB AUXILIARY CRANE a) The crane is, designed, fabricated, installed; inspected, tested, and operated,,in accordance with the requirements of ANSI B30.2, "Safety Standards for Overhead and'antry Cranes",'nd CMAA Specification No. 70,,
"Specifications for Electric Overhead Traveling Cranes", as applicable.
b) Design, fabrication, workmanship, materials, and construction of the crane structure is in accordance with the AISC, "Specification for the Design, Fabrication and-Erection of StructuraL Steel for Buildings" except that permissible unit stresses in welds are as contained in Table 9.3-1 and Paragraph 9.3 of AWS Dl.l~ "Structural Welding Code", or ~i Specification No. 70, "Specifications fo'r Electric Overhead Traveling Cranes,' whichever is more restrictive.
- 1) Allowable Stresses and Safety Factors The following allowable stresses and factors of safety are used unless greater strength requirements are specified in CMAA Specification No 70 for Class A Cranes., Factors of safety are.
based on the loading combinations and'he ultimate strength of the materials
-~
1 (a) Hoist rope.has 'a factor'of safety of six based, on the lead line stress computed from the rated load, modified by a factor
'epresenting 'the efficiency of lifting tackle; but excluding impact.'load.-
(b) All parts subject to dynamic strains such as gears, shafts, drums, blocks, and other integral parts have a factor of safety of fivei (c) Bending, stress combined with torsional stress in pins and axles does not exceed 20 percent of the yield stress of material used.
(d) Normal torsional deflection in bridge drive shaft is limited to 0.08 degrees per foot at two thirds of rated motor torque or a total movement at wheel circumference of 1/2 in., whichever is more restrictive.
(e) Hook stresses do not exceed 1/5 of .the ultimate strength or 1/3 of the yield strength of the material used, whichever is smaller.
- 2) Design and Loading Conditions.
All structural and mechanical parts of the crane are designed to resist dead and live loads, seismic loads and the forces produced by impact and thrust. Fatigue analysis is considered for load 9~ 1. 4-19
SHNPP FSAR.,
bearing components for a usage factor of 20,000 to 100,000 full load cycles.
(a) The crane structures, components and subsystems essential to retaining and holding the load in a stable or. immobile safe position, and means provided for safely moving the crane manually with:.load and emergency lowering of the load are designed'o sustain an SSE event.
Structures, components and subsystems other than those covered above are designed such that they will not fall off from the cx'ane during a seismic event and the failure of which will not damage the seismically qualified items covered above'b)
Bridge-.trucks and trolley are provided with restraints to prevent them fxom leaving their runways with or without the design load during normal operation or under any seismic excursions (c) The portions of the vertical hoisting system components, which include the head block, rope reeving system, load block and dual load-attaching, device are each designed to support a static load .of 200 percent of the- design rated load, using allowable stresses and safety factors specified above.
(d) All parts of the cxane are designed to resist any of the following conditions of loading, using allowable stresses and safety factors specified above and those specified herein:
(1) Dead load. plus live 'load plus impact (2) .Dead load plus Uve load, plus the lateral or longitudinal thrust (3) Dead load plus live load plus SSE plus pendulum and
., swinging load effects, (a lifted Load of 10-ton is considered)
(4) Rated breakdown torque of motors (5) Collision with bumper'stops with no load For conditions (1), (2), and (5), members are designed in accordance. with basic allowable stresses of CMAA Specification No 70, AXSC Code, or AWS Dl.l, whichever governs.
For condition (3), members are designed for 1 ~ 6 times the allowable stresses. Local overstressing is permitted for condition (3) provided it can be demonstrated that the crane retains the design load and'does not fall from the runway.
Ability of the crane for emergency manual. transferring and .
lowering of the load is maintained after the seismic event.
- 9. 1. 4-20
~, ~
SHNPP FSAR
'3 For condition (4) stresses elastic limit of the material.
do not exceed 90 percent of the c) The crane is not used during the plant construction phase.
,d) e) 'll Minimum operating temperature of the crane is 50 F.
ferritic material which is used in load bearing structural members are impact-tested to determine fracture toughness of the material. Load bearing structural members are defined as structural members stressed in the process of transferring hook loads (vertical or hori.zontal) through the crane to the main runway. ASTM A-514 material is not used in any load bearing structural members; other low alloy steel may be used with CP&L's (or it' agent's) written approval.
Either drop weight'test per ASTM E-208 or Charpy tests per ASTM A-370 may be used for impact testing. The minimum operating temperature, as obtained by following procedures in Subarticle HC-2300 or ND-2300 of ASME B&PV code,Section III, Div. 1, based on the drop weight test or the Charpy V-notch impact test respectively, are not higher than 50 F.
f) Welding is performed by using welding procedures, welders, welding operators, and tackers qualified in accordance with AWS D.l.l ~
O') Postweld heat treatment of welded assemblies is performed, if necessary, when an assembly is under restraint during welding, when machining is to be. performed, or for welded steel greater than l-l/2 in. in thickness at the welded joint. Welds pn all load bearing structural members are Postweld head-treated in accordance with Subarticle NF-4620 of ASME B&PV Code,Section III, Div. 1, or other requirements as approved by CP&L (or its agent).
& h) Where practical, weld joint designs susceptible to laminar tearing are not used. Weld joints susceptible to laminar tearing are ultrasonically tested for soundness of base metal and weld metal ot the completed weld joint.
i) Full penetration butt welds on all load-bearing structural members are 100 percent radiographed for soundness of weld metal and base metal where accessible. Full penetration tee welds on all load-bearing structural members and full penetration butt-welds on all load-bearing structural members which cannot be radiographed are tested as follows:
- 1) Magnetic particle or liquid penetrant test of root pass and final weld layer.
All fillet welds and partial-penetration welds are visually inspected in accordance with and to the acceptance criteria of AWS Dl.l Paragraph 9.25.
Fillet welds and partial-penetration welds joining load-bearing structural G
- 9. 1. 4-21
SHNPP FSAR members are inspected by liquid penetrant or magnetic particle methods after the final weld layer is applied.
jthat
) The automatic and manual controls for all motions are designed such.
a malfunction in the control system will not prevent the load from being maintained at a safe, holding position.
k) The haisting system is designed to provide two completely independent load paths such that the failure of any single component in either load path system will result in the other assuming the full load and retaining safe, stable position.
it in a The following basic load path components are provided with redundant counterparts or otherwise protected against subsystem or component failure:
1),Hook
- 2) Load block
- 3) Reeving
- 4) Head block
- 5) Drum
- 6) - Braking system
~ ~
\
- 7) Gear train Hoist cable redundancy is achieved by the provision of two balanced reeving systems consisting of two separate load-sharing wire ropes reeved in such a manner that the breakage of one rope will result in its shaxe of the load being immediately transferred, without development of slack, to the other.
The centers of lift of the two independent cable systems are coincident so as to minimize the swinging or twisting motion imparted to the load block when load tx'ansfer occurs following a cable break.
Each load path is designed to resist the full 'load as specified in Paragraph b) 2).
An equalizer system of the beam type is provided whose main functions are as follows:
- 1) Continually adjust the hook load during normal hoisting operations so that the load will be shax'ed equally by all parts of the reeving system.
- 2) Transfer the shock of a cable break in a safe dynamic fashion to the remaining cable by means of a shock absorbing arrangement permitting load transfer from one side of the equalizer system to the other without the imparting of unacceptably large dynamic impact to the cable or crane.
9.1.4-22 Amendment No. 1
SHNPP FSAR The equalizer system is provided with a set of proximity limit switches actuated by an exaggerated displacement of the equalizer assembly such as would be experienced in the event of a cable break. Limit switches so actuated will set the holding brakes to stop and retain the load in a safe, stable position.
The equalizer"assembly is attached to the'rolley frame by means of a redundant system of supports.
Following the= failure of a component or subsystem, means- are provided to safely.- move and lower the, load to a laydown area to allow the failed component(s) or subsystem(s) to be repaired, ad]usted or replaced as retained }
to retu~ the crane to service. s, e An electronic load indicating device is provided 'to monitor both. load paths and will set the holding brakes in the event, of a cable break or rope load unbalance.
- 1) Dual load attaching points of redundant. design are provided as part of the load block assembly for the support of the redundant subcomponents of the hook. Each hook subcomponent and attaching point is capable of supporting three. times the rated load, statically applied,'ithout permanent deformation of any part of the hook and load block assembly other" than that due to localized stress concentrations in areas where.-additional. material has been provided for weap.
l A 200 percent static load test is performed on each redundant sub-component of the hook. Measurements of the geometric configurations are made before and after the load test.
Nondestructive examination is performed before and after the load test,
'onsisting of'agnetic particle and ultrasonic testing in accordance with the: following:-
t (a) The hook forging is ultrasonically examined in accordance with ASTM A-388. The results of 'the ultrasonic examination are analyzed and documented.
(b) Hook shank and .load areas are magnetic-particle examined by the longitudinal method'in- accordance with. ASTM E-109, appendix:.
Al, Paragraph'1.2.1.3. Other magnetic particle examination methods may be used provided care is taken to prevent. local overheating', burning or arcing of the surface to be tested. All cracks are unacceptable and all linear, indications or aligned porosity exceeding 1/4 in.-in length are unacceptable. All repairs require approval of CPGL (or its agent).
The load block is nondestructively examined by surface and volumetric techniques m) Crane motions have the following maximum speeds with full rated load:
- 1) Hoist: 5 fpm 9.1 4-23
SHNPP FSAR
- 2) Bridge Travel: 100 fpm
- 3) Trolley Travel: 50 fpm Slow speeds for precise handling and setting are provided by inching drives at five percent, of the full rated speed, for bridge and trolley travel and between 6 to 12 in. per minute for the hoist.;,
n) The maximum fleet angle does not exceed 3-1 /2 degrees at any point during hoisting, except t!hat for the last three ft. of maximum lift elevation, the fleet angle may increase slightly.
Use of reverse bends for running wire ropes are limited so that disproportionate reduction in Pire rope. fatigue life would not be expected.
diameter is at least 24 times the diameter of rope for the drums and "V'itch sheaves, and at least 12 times the diameter of rope for equalizer sheaves (Ropes are stainless cable AS'-492, type 304 with independent wire rope core having six strands of 37 wires per strand.)
o) A limit switch activated by the hook block and a gear type limit switch are provided to prevent the hoisting system from "two blocking" ~
p) An overload protection device of redundant design is provided for the hoisting system which will be actuated by,a-load in excess of 110 percent of the hook rating, thereupon opening the main hoist circuit and setting the holding brakes.
q) Brakes are. as follows:
1):Bridge Travel: Electrically released, spring-set; friction-shoe type with capacity at least equal to full operating torque of the bridge drive. Brake will operate when motor controller is in OFF
'position, when main power supply switch is in OFF position, or in the event of power failure, or an overspeed or overload condition.
- 2) Trolley Travel: Electrically released, spring-set, friction-shoe type brake with. capacity at least equal to full operating torque of the trolley drive. Brake will 'operate when 'motor controller is in OFF position, when main power supply switch is in OFF position, or in the event of a power failure, or an overspeed or overload condition.
- 3) Bridge and trolley braking systems are designed to be Single FaiLure Proof and capable. of manual operation for emergency service.
- 4) Hoist: Two electric stopping and holding brakes, and one electrical hoist-control device as further specified in sub-paragraphs (a) and (b) following:
(a) Electric stopping and holding brake for the hoist operates automatically and is of the electrically released, spring-set, friction-shoe type capable of stopping and holding 1-.1/2 times the full rated load when the power is off. Brakes are mounted on 9.1.4-24
0 SHNPP FSAR the motor shafts extending from opposite sides of the motor.
When the power is off, or tripped by overspeed or overload devices, or activated by one of the limit switches described elsewhere, the brake is capable of stopping and holding the load.
(b) Electrical hoist control devices are of the eddy current type. They are capable of controlling the lowering speed under all conditions with up to 1-1 /2 times the rated load on hooki No lowering of load occurs unless power is applied to hoist motor in a lowering direction.
- 5) Two hoist holding brakes are capable of operation for emergency lowering after a single failure. Provisions are made for manual operation of'he holding brakes in this event by means of alternate lowering and holding to provide time for adequate heat dissipation. Design for manual brake operation during emergency lowering includes features to indicate and control the lowering speed.
r) The drum is provided with structural and mechanical devices to prevent it from dropping, rotating or dispngaging from its holding brake system should failure occur in the drum shaft, bearing, nr bearing support,
'he hoist drum is provided with an overspeed switch which will cut power to the hoist and set the holding brakes should the drum attain 40 percent overspeed.
s) 'esign of the crane limits the. torque during fogging and plugging to acceptable values.
t) Drift point in the electrical power system for bridge or trolley movement is provided only for the lowest operating speeds.
u) The crane is pendant controlled from the operating floor. The pushbutton pendant control is suspended from the bridge and supported from a motorized messenger track so that the pendant can be placed at any position along the length of the bridge. The pushbuttons are spring loaded to ensure automatic return to "OFF" position when buttons are released. The pendant control cable is mounted to an electric hoist system to enable the operator to vary the height of the control.
An additional control station, mounted on the building wall near the storage position of the crane, is provided to control the movements of the crane bridge and the pendant station in an area within 50 ft, from the storage position of the crane. Once the pendant station takes contro1 of the crane, the wall station will be inoperative until the main switch on the pendant is switched to the OFF position.
Purpose of the wall mounted station is to bring the crane in and out of the storage position only. Handling of loads is not controlled by the wall-mounted station.
G
- 9. 1~ 4-25
SHNPP FSAR v) The bridge and trolley are provided with accessible enclosed limit switches, which, when the bridge or trolley has traveled to within nine in. -3 of its end stop, will interrupt the current to the drive motor. Reversing the motion of the bridge or trolley will reset the switch. A bypass control is provided to allow the. bridge or trolley to approach end stops.
Also provided are two independent methods to prevent Auxiliary Crane from coming in contact with the Cask Crane as described in Section 9.1.4.2 '.7.
w) A capacity plate showing both design rated load capacity .(12 ton) and maximum working load capacity (10-ton) of the hoist is placed on each crane girder and is easily legible from the operating deck.
x) Instruction manuals are provided by the manufacturer for each component covering installation, operation and maintenance instructions in accordance with;Ebasco Instruction;.,Hanual Guidelines, General Instructions (Form 567-A),
Hechanical Equipment Including Ha)or Heating, Ventilating and Air Con'ditioning Equipment (Form 567-C) and Electrical Equipment (Form 567-E).
Instruction manual also includes the following:
- 1) Qualification requirements for crane operators
- 2) Procedure for emergency manual operations for moving the crane and lowering the load,. and locations of manual controls
- 3) Field test procedures Haintenance instructions, are based on maintaining the crane at design rated load capacity.
y) . The manufacturer is required to provide a competent, experienced representative, on completion of crane installation, to check and certify that the crane has been properly erected; instruct CP6L in the crane operation, lubrication, and periodic maintenance adjustments; and also direct CPSL in the initial crane operation to demonstrate it's satisfactory performance for acceptance by CPSL.
The crane system is static load tested at 125 percent of the design rated load. The test includes all positions generating maximum strain in the bridge and trolley structures and other positions as recommended by the manufacturer.
After satisfactory completion of the 125 percent static test and adjustments required as a result of the test, the crane is full performance tested with 100 percent of the design rated load for all speeds and motions for which the system is designed. This will include verifying all limiting and safety control devices.
The features provided for manual lowering of the .load and manual movement of the bridge and trolley during an emergency are tested with the maximum working load attached to demonstrate their ability to function as intended.
Q
- 9. l. 4-.26 Amendment No. l
SHNPP FSAR The protective overload devices are tested to ensure proper functioning of the devices by a test procedure recommended by the manufacturer.
Above tests are run within temperature range as specified for the plant operation phase.
s) The crane manufacturer has an accepted quality assurance program consistent with the pertinent provisions of Appendix B to 10CFR Part 50.
- 9. 1. 4-27
UNIT I CONTAINMENT Plant North STORAOE LOCATIOII FOR IVI CORE NEW EVEL VMLOAOVIO SAV sEACOIo AREA (InSPeCtiOn)
Attachment 3 GENERAL ARRANGEMENT OF FUEL HANDLING BUILDING Elevation 286'dministrative limit of travel for auxiliary crane with excessive loads following receipt of new fuel.
ATTACHMENT 4 RESUME OF JOSEPH R. SIPP EXPERIENCE:
November 1983 to present Carolina Power & Light Shearon Harris Nuclear Power Plant P. 0. Box 165 New Hi.ll, North Carolina 27562 Position Held: Mana er Environmental and Radiation Control May 1981 to November 1983 General Public Utilities Company 100 Interpace Parkway Parsippany, New'Jersey 07054 Position Held: Mana er of Chemical En ineerin Res onsibilities:
Review of the chemical processes in the facility to insure their optimization. This involves the review of plant chemistry, and operational data and comparing it to the system chemistry specifications and other design criteria.
Preparation of Chemistry Reports summarizing systems chemical history.
Chemical Engineering coordinates the activities between outside organizations and the plant where appropriate.
Conducting of investigations and making engineering recommendations and providing implementation with management approval.
Review of chemical processes against state-of-the art and making recommendations for their upgrade.
Provide all the Chemical Engineering for the water treatment systems.
Write all Chemistry Specifications for the plants.
Upgrade Specifications based on state-of-the art data.
Implement the necessary programs to insure the power'lant can operate at the required specifications.
(9550FXT/cfr)
Provide all the Chemical Engineering for the water treatment systems.
Provide all the Chemical Engineering for the waste water treatment systems.
Provide input to all plant procedures affected by Chemistry/Chemical Engineering.
Manage a staff of eight professionals, and a secretary.
Act as the Project Manager on unique 0 and M items at the plant.
Prepare and provi.de the technical portion of licensing responses to regulatory body findings.
Provide safety review of important-to-safety systems-April 1979 to May 1981 Public Service Company of New Hampshire Seabrook Station P. 0. Box 300 Seabrook, New Hampshire 03874 Position Held: Chemistr De artment Su ervisor Res onsibilities:
Directed a Chemistry staff of twenty-six people (had responsibility for setting up staff and preparing department budget).
Development, implementation and maintenance of the station's analytical and radiochemical monitoring program to insure compliance with state, federal, and vendor requirements for start up and operation of two 1200 MWE Nuclear Units.
Review of all chemical systems for system optimization and recommended the necessary system changes.
Preparation of all required reports, evaluation of personnel, and preparation of the department's budget.
Member of emergency team.
March 1973 to April 1979 Vermont Yankee Nuclear Power Corporation P. 0. Box 157 Vernon, Vermont 05354 (9550FXT/cfr)
Position Held: Plant Chemist Res onsibilities:
The position included the responsibility for serving on an alternate basis as onshift health physics supervisor during refueling and major maintenance outages and during normal operations'uring the subject time interval, the plant underwent an average of one refueling and two maintenance outages per year. The onshift health physics supervisor's responsibilities include the preparation and signoff for radiation work permits; and directing technicians in the implementation of work permits, conduct of radiation surveys and implementation of health physics procedures.
Directed a staff of four people as a Plant Chemist and directed a staff of thirteen people when acting as Chemistry and Health Physics Supervisor.
Assisted the Chemistry and Health Physics Supervisor in the development, implementation and maintenance of the department programs.
Participated in the review of department personnel, procedures, and generated necessary reports to satisfy regulatory requirements.
Administratively controlled all plant chemical processes.
Directed chemical decontamination of plant components.
Acted for the Chemistry and Health Physics Supervisor in his absence.
Provided expertise in the selection and evaluation of reactor water cleanup decontamination.
Directed the chemical operation in the plant to optimize their performance (Radwaste, condensate filter, demineralizers, charcoal filters, . . . etc.).
Reviewed all potential release points and ensured they were properly monitored.
Participated in the whole body counting program and respiratory protection program.
Maintained the Chemical and Radio chemical laboratories, Chemical and Radio. chemical on line monitoring equipment (calibration and data reduction).
Qualified to perform filter testing for HEPA and charcoal filters.
Member of emergency team and fire brigade.
(9550FXT/cfr)
Responsible for calibration of process radiation monitoring calibrations'ebruary 1970 to March 1973 Vermont Yankee Nuclear Power Corporation P. 0. Box 157 Vernon, Vermont 05354 Position Held: Chemistr and Health Ph sics Assistant Chemistr Res onsibilities:
Assisted and directed the Chemistry and Health Physics Technicians in routine analysis and surveillance in the chemistry laboratory.
Provided chemistry coverage in the preoperational and operational startup of the plant ~
Performed the preoperational testing of radwaste and set up the water treatment plant ~ Performed the original set up and calibration of the process monitoring system.
Health Ph sics Res onsibilities:
Directed the Health Physics Technicians in operational radiological activities during maintenance and refueling outage'his position included the responsibility for acting as the onshift health physics supervisor during major maintenance outages. These outages included several shutdowns in the startup program to correct problems and to perform reconstitution of burnt fuel assemblies'his function, carried out on a twelve hours onshift twelve hours offshift basis included direction of all onshift radiation protection activities'he initial criticality on the Vermont Yankee plant was March 24, 1972. The onshift health physics supervisor's responsibilities included the preparation and signoff for radiation work permits; and directing technicians in the implementation of work permits, conduct of radiation surveys and implementation of health physics procedures'ssisted in the continuous monitoring of all radiological activities within the plant and its systems, including the surrounding environment, to ensure maximum safety at all times'rote Radiation Work permits, performed radiation surveys and performed calibration of Health Physics survey instruments'sed the following sources for calibration of survey instruments on on-line process instruments:
(9550FXT/cfr)
137 50 Ci Cs calibrator 24 uCi Co 60 Ci Ir 239 Pu sources.
Also used liquid sources for calibration of Chemistry Instrumentation.
June 1969 to February 1970 Mine Safety Application Research Corporation Evans City, Pennsylvania 16033 Position Held: Research Chemistr Technician Res onsibilities:
Developed new charcoals.
Performed testing of charcoal to evaluate its ability to filter small quantities of organic contaminants from air.
(9550FXT/cfr)
EDUCATION:
B. S. Chemistry, 1969 Geneva College Beaver Falls, Pennsylvania M.B.A., 1978 Western New England College Springfield, Massachusetts Training at Millstone I BWR 3/70 5/70 Millstone, Connecticut Radionuclide Analysis by Gamma 6/70, 1 week Spectroscopy at Northeastern Radiological Health Laboratory In-Place Filter Testing 6/23/75 6/27/75 Department of Environmental Health Sciences Harvard School of Public Health State Fire Service Training Program 6/70 (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />) of Vermont Applied Metallurgy 1/77 5/77 Taylor and Fem Company Hartford, Connecticut Westinghouse PWR Chemistry Course 9/4/79 11/23/79 Pittsburgh, Pennsylvania Seabrook Station, 1979 (60 hours)
Balance of Plant Training Westinghouse PWR Information Course 1979 (60 hours)
Management and Supervision Courses 1979 (145 hours)
Radiochemistry Spring Semester 1980 at Northeastern Fluid Power Controls Fall Semester 1980 at University of Lowell Kepner-Tregoe Decision Analysis 2/25/80 2/29/80 Multimedia First Aid, CPR 1980 The Management and Disposal 10/29/80 10/30/80 of Hazardous and Chemical Waste Leadership Effectiveness Training 1982 (24 hours)
(9550FXT/cfr)
Hanagement Development Program 12/31/83 Phase I, II, III 4.8 CEU's GPUN Hanagement Training 3/9/83 (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />) by Sterling Institute (9550FXT/cfr)
PAPERS WRITTEN:
Presented at Edison Electric Institute Chemistry Meetings.
- l. A New Precoat Material for the Power Industry 1976.
- 2. Product Evaluation of Graver Ecodex and Ecocate at Vermont Yankee 1977.
Presented at the Cooling Tower Institute Annual Meeting.
- 1. Improve Condenser Cleanliness by Using Dispersant to Supplement Chlorination at a Nuclear Power Plant 1978.
Presented at the Western Engineering Conference in the Fall of 1980.
- 1. Prepared discussion of Three Years Operating with Precoat Materials Containing Fibre and Powdered Ion Exchange Resins.
(9550FXT/cfr)
Attachment (
I O.Q 8.Y5c~
SoRA,P LE,X I
sf I
F i pure 1 I'-ac4 hssc~nb I y b~ ta i Is (pgg)
SHNPP FSAR ttachment (6)
- 15. 7~4 DESIGN BASIS FUEL HANDLING ACCIDENTS 15.7 ~ 4 1 Identification of Causes and Accident Descri tion The accident is defined as dropping'f. a spent fuel assembly resulting in the rupture. of the cladding of"all the fuel rods in the assembly. The possibility of a fuel handling accident's remote because of the many interlocks, administrative controls, and physical limitations imposed on the fuel handling operations -All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a senior reactor operator (SRO) ~
Design of the fuel storage racks and handling facilities ig both the Containment and fuel storage area is such that duel will always be in a subcritical geometrical array. The design assumes zero boron concentration in the fuel pool water. The spent fuel pool and reactor cavity water contains boron at the refueling boron concentration. Natural convection of the surrounding water provides adequate cooling of the fuel during handling and storagei Cooling of the water is provided by the Spent Fuel Pool Cooling and Cleanup System At, no time during the transfer from the reactor core to the spent fuel storage rack is a fuel assembly removed from the water. Fuel failux'e during refueling, as a result of inadvertent criticality or overheating, is not possiblei For this evaluation, dropping of a fuel assembly is assumed to occur, breaching the cladding"and releasing the volatile fission products in the gas gap of the fue1.".'ins. In addition to. the area. radiation monitor located in the Fuel Handling Building, portable radiation monitors which emit audible alarms are=located in this area during fuel handling operations 'oors in the Fuel Handling Building are closed to maintain controlled leakage characteristics in the fuel pool stoxage region during refueling operations that involve irradiated fuel. Should a fuel assembly be dropped in the fuel transfer canal, in the new fuel pool, or in the spent fuel pool and release radioactivity above a pxescribed level, the airborne radiation monitors will sound an alarm,. alerting personnel to the problem Airborne radiation monitors in the exhaust ducts from the Fuel Handling Building isolate the normal Fuel Handling Building Ventilation System and automatically initiate the filtration systems 15~7.4i2 Method of Anal sis The following model is postulated for calculation of the fuel handling accident:
a) The accident occurs, at 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following the reactor shutdown; i.e.,
the time at which spent. fuel would be first removed from the reactor and moved into the spent fuel pool.
b) The accident results in breakage of all fuel rods in an assembly.
c) The damaged assembly is assumed to be the one operating at the highest power level in the core region being discharged.
15 7.4-1
SHNPP FSAR d) 'he power in this assembly, and corresp'onding fuel temperatures, establish the total fission product inventory and the fraction of this inventory which is present in the fuel pellet-cladding gap at the time of reactor shutdown.
e) The fuel pellet-cladding gap inventory of fission products is released to the fuel pool water at the time of tQe accident.
f) The fuel pool water retains a.large fraction of the gap activity of.
halogens by virtue of their solubility by hydrolysis." Noble gases are not retained by the water.
- 15. 7.403 Radiolo ical Conse uences 15.7A+3o 1 Postulated Fuel Handling Accident Outside Containment Assumptions and parameters used in evaluating the fuel handling accident outside Containment are consistent with Regulatory Guide 1 25 (3/23/72) recommendations as shown in Table 15,7i4-li The calculational methods and assumptions described in Regulatory Guide I 25 apply since: (a) the values for maximum fuel rod pressurization, (b) peak linear power density for the highest .power assembly discharged, (c) maximum centerline operating fuel temperature for the assembly in item (b) above, and (d) average burnup for the peak assembly in item (b) above are less than the corresponding values in Regulatory Guide 1 25 .
lP The whole body dose'due to immersion and. the thyroid dose due to inhalation have been analyzed. for the 0-2 hour period at the exclusion area boundary and .
for the 0-8 hour period at the LPZ outer boundary. The doses calculated at the exclusion area boundary have been evaluated using the 0-2 hour atmospheric dispersion factor, X/Q, and the corre'dponding doses at the LPZ have been using the 0-8 hour X/Q value For the purpose of these accidents
'alculated the activity has been assumed to'be instantaneously released from the Fuel Handling Buildings The results are listed in: Table 15i7i4-2i The resultant doses are well. within the guidelines of IOCFR100 .
15 7 ' 3 2 Postulated Fuel Handling Accident Inside Containment The possibility of a fuel handling accident inside Containment during refueling is relatively small due to the many physical, administyative, and safety restrictions imposed on refueling operations During fuel handling operations, the Containment is kept in an isolable condition, with all penetrations to the outside atmosphere either closed or capable of being closed. on a containment purge isolation signal (CPIS) initiated by redundant area and airborne radiation monitors. At least one of the two interlock doors on the personnel locks is kept closed. In addition to the area and airborne radiation monitors in the Containment, portable monitors with audible alarms are located in the fuel handling area during refueling.
Should a fuel assembly be dropped and release activity above a prescribed level, the radiation monitors sound an audible alarm, the Containment is isolated and the personnel are evacuated. The containment purge lines are automatically closed upon a CPIS, thus minimizing the. escape of any 15.7 4-2
SHNPP FSAR radioactivity. The consequences of dropping a fuel assembly in the Containment are less severe than the consequences of dropping the assembly in the Fuel Handling Building, since the Containment provides a considerably greater holdup time than the Fuel Handling Building, allowing for radioactive decay of the released fission products.
For analytical purposes, consideration is given to one accident; a drop of a fuel assembly into the refueling cavity by the manipulator crane inside Containment. Assumptions and parameters used in evaluating the fuel handling accident inside Containment are shown in Table 15.7.4-3.
The radiological consequences of a fuel handling accident inside containment were conservatively evaluated by assuming that containment releases for the first 20 second period, it was assumed that further releases were made in a controlled manner through the Reactor Auxiliary Building Filtration System charcoal adsorbers. The assumption'o'f a controlled release was made due to the availabilit'y of these. charcoal adsorbers and based on calculations showing that containment isolation can be achieved prior to radionuclide reaching the first containment isolation valve.
The activity released inside containment as a result of a fuel handling accident will be detected by area radiation monitors. The response time for these monitors is expected to be less than 5 seconds. Following activity detection, the monitors will initiate the closure of the containment isolation valves. The valves will require 15 seconds to close. The containment will be isolated in 20 seconds after the accidental release of radioactivity.
The time required for airborne activity to reach the containment isolation valve is based on 1) travel time from the surface of the reactor pool to the nearest intake header and 2) travel time through the duct. The nearest intake
- ~ header from the pool is located at a.distance of 14.8 feet. The average
'. airflow velocity within a distance of less than 3 feet and the velocity at 3 feet from the intake header was estimated using equations from. Reference
'"'15.7.4-1'. These equations are as follows:
(1) 10X+ A V -1 R 10A av tan .A (2) .
36 10A where, X distance outward along axis, ft. (Note: Equation is accurate only when X is less than 1 1/2 D)
V centerline velocity at distance X from hood, ft/min.
V ~ average velocity within a distance X, ft/min.
Q ~ air flow, cfm A area of hood opening, ft 15.7.4-3 Amendment No. 5
SHNPP FSAR D = diameter of round hoods or side of essentially square hoods.
The air velocity beyond 3 feet, though expected to be smaller, is assumed to remain the same as that at 3 feet from the intake header. The size of the intake header is 24" x 24". The average air velocity up to and including 3 feet was estimated at 179.6 ft/min and the average air velocity beyond 3 feet was estimated to be for a given intake rate of 2500 cfm through the header.
Therefore the activity would take 27.6 seconds to reach the intake header from the surface of the pool. The travel time in the 26 in. x 30 in. duct (106 ft.
in length) would be 3.4 seconds . Therefore the total time required before the.
activity would reach the isolation valve is 31 seconds.
The following conservative assumptions are based on Regulatory Guide 1.25 and inherent plant design parameters used to calculate the activity releases and offsite doses for the postulated fuel handling accidenc inside Containment.
a) The accident is assumed to occur 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following reactor shutdown for refueling.
b) All rods in one fuel assembly are ruptured.
c) The assembly damaged is assumed the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full power operation at the end of core life immediately preceding shutdown. A radial peaking factor of 1.65 is used.,
E) d) , All of the gap activities in the damaged rods is released and consist of the 10 percent of the total 'noble gases other than krypton-85, 30 percent of the krypton-85, and 10 percent of the total radioactive iodine in the rods at the time of the accident.
e) , The iodine gap inventory is composed of inorganic species (99.75 percent) and organic s'pecies (0.25 percent).
f) The refueling cavity water decontamination'actors for the inorganic and organic iodine are 133 and 1 respectively, giving an overall effective decontamination factor of 100.
g) The retention of noble gases in the refueling cavity water is.
negligible.
h) The accident occurs during refueling with the Containment Purge System operating.
i) It was K N conservatively assumed that the radioactivity. released as a result of the fuel handling accident, instantaneously reached the dose point with no credit for containment isolation.
The doses from a fuel handling accident occurring inside Containment have been calculated, assuming no isolation, and have been found to be below the
. guidelines of 10CFR100. The results of this analysis are presented in Table 15.7.4-4.
'SHNPP FSAR TABLE 15 ~ 7. 4-1 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT OUTSIDE CONTAINMENT Design Basis Realistic Parameter Assum tions Assum tions Source Data:
Power level, MPt 00 2775 Radial peaking factor 65 I ~ 55 Burnup 3 full-power 3 fM1-power years years Decay time, hr 48 48 Number of failed assembly I I/17 Fraction of fission product gases contained in the- gap region of the fuel rods, percent Kr 85 30 30 Other Noble Gases 10 10 Iodine 10 10 Activity Release Data:
Fraction of gap activity released to pool ,'00 100 Minimum water depth above damaged rods, fthm 23 23 Pool decontamination factor for noble gases =
Pool decontamination factor, for iodine P Inorganic 133 Organic 1 500 Overall 100 500 Iodine chemical form released to fuel building Inorganic iodine percent 75 75 Organic iodine, percent 25 25 15 7.4-5
SHNPP FSAR TABLE 15.7.4-1 (Cont'd)
Design Basis Realistic Parameter Assum tions Assum tions Filter Efficiency Iodine, inorganic percent 99 99 Iodine, organic percent 99 99 Noble gas percent 0 0 Activity released to atmosphere, (Ci) <<t Isotope I-131 6.4 x 100 I-133 3.4 x 100 Xe-131m 5.2 x 102
'e=133 1 ~ 4 x 105 Xe-135 1 2 x 103 Kr.-85 2 5 x 103 Data:
'ispersion Atmospheric dispersion factors 5 percentile level 50 percentile level X/Qs, (Table 2.3 4-5) X/Q, (Table 2 3.4-5)
Doses Calculation Model Dose Model
.as"discussed in Appendix 15.0A
- 15. 7. 4-6
SHNPP FSAR TABLE 15. 7. 4-2 RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT IN THE FUEL HANDLING BUILDING OUTSIDE CONTAINMENT Design Basis Assumptions Result (one assembly)
Exclusion Area Boundary Dose (0 to 2 hr,)
(rem)'hyroid 2.4 x 100 Whole body 6.4x10 i LPZ Outer Boundary Bose (duration) (rem)
Thyroid 5.5x10 i Whole body 15xlOi J
C iS.7.4-7
SHNPP FS~
TABLE 15.7.4-3 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Design Basis Realistic Parameter Assum tions Assum tions Source Data:
Power level, MVt 2900 2775 Radial peaking factor 1 '5 1o55 Burnup 3 full-power 3 full-power years years Decay time, hr 48 - 48 Number of failed assembly 1 1/17 Fraction of fission product gases contained in the gap region of the fuel rods, percent Kr 85 30 30 Other Noble Gases 10 10 Iodine . 10 10 Activity Release Data Fraction of gap activity released. to pool 100 100 Minimum water- depth above damaged. rods, 23 . 23 ft'ool decontamination factor
, for noble gases Pool decontamination factor for iodine Inorganic 133 Organic 1 500 Overall 100 500 Iodine chemical form released to fuel building Inorganic iodine percent 75 75
,Organic iodine, percent 25 25 Containment Isolation Time Following Accident (Sec) 20 20 Containment Purge (cfm) 37,000 37,000 Containment Volume (cu ft.) 2 '7 x 106 2 '7 x 106 15 7.4-8
SHNPP FSAR '
TABLE 1$ . 7.4-3'Cont'd)
Design Basis Realis tie Parameter Assumptions Assumptions Filter Efficiency Iodine, inorganic percent 90 99 Iodine,'rganic percent 90 99 Noble gas percent 0 0 Activity .released to atmosphere, (Ci)
Before' Isotope Isolation To La 1*
I-131 2.8 5.6 x 10 1 I-133 3.3 x 10 6.6 Xe-131m 2.4 4.6 x 10 2 Xe-133 5.4 x 10 1.0 x 10 XQ-135 1.3 x 10 2.5 x 10 Kr-85 1.4 x 10 2.6 x 103 V
' Dispersion Data P~
4 Atmospheric 5 percentile level 50 percentile level dispersion factors X/Qs, (Table 2.3.4-5) X/Qs, (Table 2.3.4-5)
Dose Calculation Dose model Hodel as discussed in Appendix 15.0A
- Total is the combination of doses before isolation and after through controlled purge.
~ 1S.7.4-9 <<Aa~~W 'lT~
Sf%PP FSAR
.TABLE 15.7.4-4 RADIOLOGICAL CONSEOUENCES OF A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Design Basis Ass'umptions Result Before Isolation Total*
Exclusion Area Boundary Dose, (0 to 2 hr.) (rem)
Thyroid 9.5 x 10 1.9 x.10 Whole body 3.4 x 10 4.7 x 10-1
'?,PZ Outer boundary Dose (duration) (rem)
Thyroid 2.2 x 10 4.3 Whole body 7.6 x 10 1.1 x 10-1
- Total is the combination of doses before isolation and after isolation through controlled purge.
15.7.4-10 Amendment No. 5
~l 1 '4 ~ e