ML20205D157
| ML20205D157 | |
| Person / Time | |
|---|---|
| Site: | 07003008 |
| Issue date: | 09/13/1985 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Davies J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 2NRC-5-130, NUDOCS 8509230447 | |
| Download: ML20205D157 (34) | |
Text
_ _ _ -
'Af 41 ) 78 5 41 Nuclear Constrtiction Division Robinson Plaza. Building 2. Suite 210 Tecon Pittsburgh. PA 15205 September 13, 1985 United States Nuclear Regulatory Conunission Washington, DC 20555 gO'3O ATTENTION:
Mr. John G. Davies, Director Nuclear Material Safety and Safeguards
SUBJECT:
Beaver Valley Power Station - Unit No. 2 Docket No. 50-412 Application for Special Nuclear Material License
REFERENCES:
1.
2NRC-4-162, dated September 28, 1985 2.
Letter from N. Ketzlack to E. J. Woolever dated February 8, 1985 Gentlemen:
Please find enclosed one (1) original plus six (6) copies of Revision 1 to the Special Nuclear Material (SNM) License application submi t t ed fo r review on September 28, 1984 (Reference 1).
This revision incorporates the additional information requested in Reference 2 and Kishore Kodali's comments provided in a meeting on June 13, 1985.
This revised application is filed pursuant to 10CFR70 for authoriza-tion to rece ive, possess, store, ins pect package fo r t rans po rt, and ins tall unrad iated nucle ar fuel assemblies, flux mapping unveable incore detectors,
neutron detector systems, primary source rods, and various detectors and monitors and their calibration and check sources fo r Beave r Valley Power Station Unit 2.
Communications pursuant to this license application should be sent to:
Mr. J. J. Carey Vice President DUQUESNE LIGHT COMPANY Robinson Plaza II Suite 210 PA Route 60 Pittsburgh, PA 15205 If you have any additional que s t ions, please contact Mr. Thomas Zogimann at (412) 787-5141 DUQUESNE LIGHT COMPANY SUByRIBEDANDfyWORNTOBEFORKMETHIS
// DAY OF y/pga//nf, 1985.
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Mr. B. K. Singh, Project Manager (w/a)
Mr. G. Walton, NRC Resident Inspector (w/a)
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United Ststec Nuclear Rsgulatory Comnission Mr. John G. Davies, Director Application for Special Nuclear Material License Page 2 COMMONWEALTH OF PENNSYLVANIA )
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1.0 GENERAL INFORMATION 1.1 Reactor and Fuel 1.1.1 General The Beaver Valley Power Station (BV?S) is located in Ship-pingport Borough, Beaver County, Pennsylvania.
The site is approximately 25 miles northwest of Pittsburgh, Pennsyl-vania. A detailed description of the geographic location is provided in the BVPS-2 Final Safety Analysis Report (FSAR)
Section 2.1.1 (attached), Docket No. 50-412.
The BVPS-2 construction permit number is CPPR-105.
1.1.2 Fuel Assenblies The 157 nuclear fuel assenblies consist of slightly enriched 1
uranium dioxide pellets encased in Zircaloy-4 rods.
The average outside diameter of the slightly enriched uranium dioxide pellets is 0.3225 inches.
The fuel rod pitch in a fuel assembly is 0.496 inches. The zircaloy fuel rods have a nominal outside diameter and wall thickness of 0.374 inches and 0.0225 inches, respectively. Each assembly contains 264 fuel rods, 24 Zircaloy-4 control rod guide thimbles, and 1 Zircaloy-4 instrumentation thimble arranged in a 17 x 17 matrix (see FSAR Figure 4.2-1 [ attached]).
The 17 x 17 1
matrix is maintained by 8 inconel grid assemblies located along tha length of the fuel assembly.
The assembly top and bottom nozzles are constructed of stainless steel.
The assembly is approximately 160 inches in length with a nominal active fuel length of 144 inches.
Each assembly is approximately 8.4 inches square.
1.1.3 Assembly Enrichment and Weights The initial core contains nominal assembly enrichments of 2.10 weight percent (w/o), 2.60 w/o, and 3.10 w/o U-235 (1 0.05 w/o).
The total uranium weight per assenbly is nominally 460 Kg of which less than 15 Kg is U-235.
Fuel pellet design density is 95 percent of theoretical.
Tne 1
total assembly design weight, including structural compo-nents, is 665 Kg. The fuel assenblies contain no U-233, Pu, depleted uranium, or thorium.
The maximum fuel assenbly enrichment to be stored under the BVPS-2 Special Nuclear Material License is 3.10 w/o U-235 (10.05 w/o).
The highest anticipated enrichment assumed for nuclear criticality safety analyses is 3.6 w/o U-235.
1.1.4 Total Fuel Assemblies and Uranium The total number of fuel assemblies in the initial core is 157; no spare fuel assemblies were purchased.
The total,
1 1
weights of U-235 and uranium are approximately 1,875 Kg and 72,220 Kg, respectively.
1.2 Storage Conditions 1.2.1 Fuel Storage Area Fuel storage and handling operations will be perfonned in the fuel building.
One hundred and fif ty-seven (157) f uel assenblies will be stored in the spent fuel racks or up to seventy (70) fuel assenblies may be stored in the new fuel racks with the renaining a3semblies stored in the spent fuel racks (1,088 available spent fuel rack storage locations).
Detailed elevation and plan views of the Fuel Building are shown in FSAR Figures 9.1-1 and 9.1-2 (attached).
Temporary storage of new fuel assenblies in their shipo4g containers may be necessary for short periods of time during new fuel receipt. The shipping containers for fuel movement will be those covered by certificate of compliance number 5450.
If such storage is required, the new fuel will be stored in a horizontal position in closed shipping con-tainers stacked no more than two high and in groups of not more than 60.
1 The new fuel storage racks are located in a different area of the Fuel Building and are separated from the new fuel receipt and inspection area by an eighteen (18) inch thick reinforced concrete wall.
In addition, the new fuel receipt and inspection operating area is. located sixteen (16) feet below the new fuel rack area (Fuel Building elevation 735'-6" and 752'-3", respectively).
1.2.2 Fuel Storage Area Activities Only those activities which involve new fuel receipt, fuel inspection, and fuel handling and storage are normally con-ducted in or adjacent to the fuel handling and storage areas.
No construction activities or test activities which could possibly result in damage to the fuel will be allowed in the fuel storage or adjacent areas during fuel receipt, I
handling, inspection, or storage.
All construction and test activities in or adjacent to the fuel handling and storage areas which could effect fuel receipt will be conpleted prior to the scheduled fuel receipt date.
1 1.2.3 Fuel Handling Building Equipment and Systens The Fuel Handling Building structures, components, equip-ment, systens, and the design criteria used to ensure their 2
structural integrity are described in FSAR Sections 9.1 and 3.8.4 (attached).
Prior to receipt of new fuel, required fuel handling equip-ment and storage facilities will be inspected and tested to ensure safe operation during new fuel handling activities.
1.2.4 Fire Alarm and Control Systems As presented in Fire Protection Evaluation Report (FPER)
Figure Al-2 and FPER Table 1 (attached), the Fuel Handling i
Building combustible loading classification is low based on combustible loading calculations.
The barriers separating the fire areas are constructed of concrete block or poured, reinforced concrete, or both, with approved fire doors, fire L
danpers, and penetrations of an equivalent rating.
Fire protection will be provided by hose stations and portable extinguishers. A detailed discussion of the fire protection progran is presented in FSAR Section 9.5.l.2 (attached).
BTP CHEB 9.5-1 requires - that the fire protection progran l
(plans, personnel and equipment) for buildings storing new j
reactor fuel and for adjacent fire zones which could affect the fuel storage zone to be fully operational before fuel is received at the site.
Therefore, the Fire Protection Pro-grans for the Fuel Handling Building will be in effect prior to receipt of fuel on the site as stated in the BVPS-2 Fire Protection Evaluation.Repert Section 1.1.6.
1.3 Physical Protection See tne Beaver Valley Physical Security Plan which was submitted under separate cover (2NRC-3-055) to Mr. Harold R. Denton, Office of Nuclear Reactor Regulation on July 28, 1983.
The Physical Security Plan will be implemented in the Fuel Handling Building by the.date of fuel receipt and will be in effect whenever fresh fuel is stored onsite.
DLC is licensed to operate a nuclear power reactor (BVPS-1) pursuant to 10CFR50 and has a Physical Security Plan (10CFR50.34) in effect at the present time.
The Physical Security Plan will be expanded.to include BV-2 prior to receipt of fuel on site.
1 1.4 Transfer of Special Nuclear Material Should it De necessary to ship Special Nuclear Material prior to Duquesne Light will repackage receiving-the operating license, the material in accordance with 10CFR Part 71 prior to delivery to a carrier for transport.
1.5 Financial Protection and Indemnity In accordance with the Code of Federal Regulations, Chapter 10, Section.140.13, financial protection in the anount of $160,000,000 3
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has been obtained from Nuclear Energy Liability Insurance Associ-ation (NELIA) and Mutual Atomic Energy Liability Underwriters (MAELU).
Copies of the policies are on file with the Nuclear Regulatory Commission.
The broker of record is Conover & Associ-ates, 931 Penn Avenue, Pittsburgh, Pennsylvania 15222.
Proof of Secondary Financial Protection will De submitted to the Nuclear Regulatory Commission at such time as necessary to facilitate the issuance of a Part 50 License.
2.0 HEALTH AND SAFETY 2.1 Radiation Control 2.1.1 Training and Experience The training and experience of the key BVPS Health Physics personnel meet the requirements of ANSI N18.1-1971 and Regulatory Guide 1.8.
Additional information regarding the qualification of BVPS personnel is described in FSAR Table 13.1-1 (attached).
2.1;2 Procedures and Equipment Administrative controls which govern the safe handling and storage of fuel will be the responsibility of the Refueling Supervisor. Tne Refueling Supervisor position meets the requirements of ANSI N18.1-1971 (Attachment G).
Those 1
procedures which control the safe handling of fuel will meet the intent of 10CFR70.58 and be approved by the Onsite Safety Committee (OSC). The function of the OSC is described in Section 13.4.1 of the BVPS-2 FSAR (attached).
The manipulation of the new fuel assemblies will be per-formed by BVPS-2 Operations personnel trained in proper fuel handling techniques and, in addition, will be done in accor-dance with approved written fuel handling procedures con-taining provisions to-ensure that all fuel assemblies are handled correctly.
All training relating to the reciept, inspection, and storage of new fuel will be complete before receipt of fuel.
Radiation and contamination monitoring will be perfonned prior to the initial handling and storage of new fuel.
All new fuel that has not been unloaded or unpacked will De handled as coritaminated material with all appropriate radio-logical controls in effect until contanination checks are performed. New fuel will be checked for radioactive contan-ination by BVPS-2 Radiation Protection personnel as part of the new fuel inspection procedure. Swipes or smears will be taken of. the fuel in order to obtain a representative sanple of the surf ace contanination of the entire' assenbly and will be counted for alpha and beta / gamma activity to determine the anount of contanination present.
If the anount of 4
1
contanination is found to exceed allowable limits (2 0.005 microcuries of renovable contanination), the source of the contamination will be determined and appropriate decontani-nation steps will be initiated as required.
This practice should identify any possible radiation hazards associated with external contanination of new fuel assenblies and allow proper planning for ALARA controls deened necessary by Radiation Protection personnel.
The BVPS-2 Health Physics i
Progran outlined in FSAR Section 12.5 (attached) describes the procedures and equipment involved in radiological controls.
Due to the fact that the fuel will be unirradiated, there will be no significant radiation hazard associated with the low level radioactivity of the fuel itself. The handling and storage of the fuel, as outlined above, will be sufficient to maintain radiation exposures ALARA.
2.1.3 Detection Calibration and Testing Testing of the detectors used to measure radioactive contan-ination on new fuel assemblies will consist of daily checks as required on background radiation, detector ef ficiency, the updating of a daily trend plot of detector ef ficiency, and the updating of a daily trend plot of detector perfor-mance.
In addition, the instrumentation will be calibrated as a minimum on a quarterly basis using appropriate calibra-tion sources.
2.2 Nuclear Criticality Safety After receiving the shipping containers at the plant site, only one metal shipping container with fuel assemblies will be opened at any one ti:ne.
Each fuel assenbly will be renoved froin its shipping container and inspected at a new fuel inspection station.
If no defects are found, the fuel assenbly will be moved to the fuel storage racks, f
- 2. 2.1 New Fuel Racks Tne new fuel storage racks consist of a stainless steel support structure into which seventy (70) stainless steel fuel guide assemblies are bolted in 14 parallel rows of five fuel guide assenblies each. FSAR Figure 9.1-3 shows the new fuel storage racks.
The center-to-center spacing between fuel assemblies is 2111/16 inches.
The guide assenbly wall thickness is 1/8 1 1/64 inches (see FSAR Figure 9.1-3, detail "B").
1 The new fuel storage racks are designed to include storage for one-third core plus 17 spare assenblies.
The design of the fuel storage rack assembly is such that it is impossible 5
to insert the new fuel assenblies in other than prescribed locations.
1 The plastic covering around each assembly will be opened at tne bottom to allow water drainage should flooding and then drainage of the fuel storage area occur.
These racks are designed to prevent accidental criticality even if unborated water is present.
The design of the new fuel storage racks is such that the effective multiplication factor (keff) does not exceed 0.98 with fuel of the highest anticipated enrich-ment in place, assuming optiman moderation.
The new fuel storage racks are designed to withstand nonnal 4
operating loads as well as Safe Shutdown Earthquake (SSE) and Operating Basis Earthquake (0BE) seisinic loads meeting i
the design criteria of ASME B&PV Code Section III, Appendix XVII.
The fuel storage racks are also designed to meet the seismic Category I guidance of NRC Regulatory Guide 1.29, Revision 2, February 1976.
Damage to the fuel assemblies and the new fuel racks by excessive uplif t forces fron the new fuel handling hoist are prevented by operating proce-7 dures and by a load cell attached to the crane.
2.2.2 Spent Fuel Racks The spent fuel racks consist of two parts, a subbase bean system and 17 individual rack assenblies which provide fuel storage on 10-7/1611/8 inch center-to-center spacing.
The inside dimension of the fuel storage cell is 8-15/16 inches. The systen of interconnected base beams is provided 1
to bridge the space between embedment pads so load transfer from the racks to the floor occurs only through the sabed-ment pads. Each rack assembly, consisting of an 8 x 8 array of storage cells, is bolted to the base beams. Because the entire complenent of base beans and 8 x 8 assemblies form a single structural unit, relative sliding between assemblies is eliminated.
The storage locations on the edges of adja-cent 8 x 8 arrays are spaced at 11-5/811/4 inch centers.
The storage racks are positioned such that adequate clear-ances are provided between the racks and pool walls to avoid impacting during seismic events. The manufacturing tolerance is controlled such that the total tolerance per rack assem-bly is 1 1/8 inch.
Figure i shows a rack assenbly.
j
. Sufficient space does not exist between storage cells or adjacent to storage cells to allow a fuel assembly to be placed closer than the minimum center-to-center spacing.
Where it might be possi'>le to place a fuel assembly between the fuel pool wall and the fuel racks, fuel guards are utilized to ensure that a twelve (12) inch boundary is maintained.
l The racks are designed with an internal neutron absorber to allow a high density storage design.
The absorber is j
6 l
l
?
provided as boron carbide in a non-metallic binder (BORA-FLEX). The Boraflex binder is Dow-Corning 170, the standard binder used in the manufacturer of Boraflex.
Each storage 1
location consists of a square stainless steel tube with 0.090 1 0.005 inch wall thickness.
As shown on Figure 1, each storage location has sheets of Boraflex attached to all sides.
The Boraflex is provided as an 0.06810.010 inch x 7.510.0625 ing x 145 incn sheet with an areal density 0.017 grans B per square centimeter.
The Boraflex of sheets are located vertically in such a manner that the horizontal centerline is at the mid-plane of the active fuel length of a stored fuel assenbly.
The " canning plates" provide the support to hold the Boraflex poison sheets in place on the outside of tne fuel cell walls.
Since the retainers and poison are in contact with each other, a gap dimension is not provided.
1 In the production of Boraflex, isotropic analysis is used to ensure that the required concentration of boron is provided at 95/95 confidence level.
The weight-percent of B10 !
present in each lot of boron carbide powder is directly determined by isotopic analysis.
This value is used by the Boraflex manufacturer in calculating a minimum required specific gravity for each Boraflex production batch to ensure that the required areal 810 content is met at the minimum material thickness.
If the actual thickness is above the minimum thickness on which the calculations are based, the actual B10 content will exceed the minimum required. Representative values for the Boraflex are:
Minimum Weight-percent of B C: 49%
4 MaximumWeightgrcentofDow-Corning 170: 517.
1.77 grans/cm} grams /cm2 conter.t: 0.0l Minimum Areal B 1
Boraflex density:
In the assembly of the spent fuel racks, in addition to the vendor quality assurance
- progran, independent quali ty inspections were utilized to ensure that the specified amount of Boraflex was installed in each assembly.
During the f abrication of the storage cell subassembly, top and bottom Boraflex locating stops are welded in position on the exterior of each side of each storage cell.
The Boraflex material is then laid between the stops and the Boraflex cover is welded on, securing the Boraflex in place.
All operations are listed on NES manufacturing travelers, which are signed off at each step.
As a final check, the canning plates (0.0293 1 0.005 inch) contain a 1/2 inch di aneter hole which allows the verification of the presence of Boraflex material af ter the canning plate has been welded in place.
1 The spent fuel racks meet the requirenents of SRP 3.8.4 Appendix D and are designed to withstand a maximum uplif t 7
1
load of 4000 pounds.
All materials that cone in contact with the fuel assenblies are made of corrosion resistant stainless steel.
The spent fuel storage racks are designed to withstand shipping, handling, nonnal operating loads (dead loads of fuel assemblies), as well as SSE and OBE loads. These racks meet ASME B&PV Code,Section III, Appen-dix XVII requirenents.
The racks also meet the seismic category I guidance of Regulatory Guide 1.29, Revision 2, February 1976.
The spent fuel storage racks have been designed to withstand the impact of a dropped fuel assembly from the maximum height of 23.5 inches.
The analysis results show that the fuel cell deforms in compression and shortens in lengtn.
The accident would not result in an unsafe geonetric spacing of fuel assenblies.
Cranes carry-ing loads heavier than 3,000 pounds will be prevented by interlocks fran traveling over the new and spent fuel storage areas when fuel is stored in the new and spent fuel storage racks.
The fuel storage cells, 8 x 8 rack array, and cell weldment I
drawings are provided to assist in the independent criti-cality analysis. (SWEC File Nos. 2003.520-040-012, 2003.520-040-026, and 2003.520-040-016, respectively.)
1 2.2.3 Criticality Analysis Nuclear criticality safety evaluations for fuel stored in the new fuel racks and the spent fuel racks were perfonned assuming fuel of the highest anticipated enrichinent in place and optimum moderation conditions.
Criticality analyses were performed assuming moderator densities ranging from
.0001 gm/cm3 to 1.0 gm/cm3 The results of the criticality analyses indicate that keff does not exceed-.98 under optimum moderation conditions.
The design basis for the spent fuel storage criticality analysis is that there is a 95% confidence level that the keff of the fuel storage array will be less than 0.95 per ANSI Standard N18.2-1973.
The results of the analysis for an infinite array of 17 x 17 assemblies enri'hed to 3.6 w/o U-235 denonstrate that for a 14.0 inch cente
'-center rack spacing there is at least a 95% probability,
- % confi-dence level that keff will not exceed 0.95.
In the analysis for both the new fuel storage racks and the spent fuel storage racks, the fuel assemblies are assumed to be in their most reactive condition, nanely fresh or unde-pleted and with no centrol rods or' removable neutron absor-bers present. Assemblies cannot be closer together than the design separation provided by the storage racks.
The mechanical integrity of the fuel assembly is assumed.
Detailed explanations of the criticality safety studies and 8
t
9 their results are presented in FSAR Sections 9.1 and 4.3.2.6 (attached).
2.2.4 Fuel Manipulation New fuel elements will tje renoved fron their usual storage locations fron time to time for such activities as fuel assenbly relocation in storage and fuel inspection.
The manipulation of the new fuel assenblies will be perfonned by DLC operations personnel trained in proper fuel handling techniques and, in addition, will use fuel handling proce-dures which cor.tain provisions to ensure that fuel assem-blies are handled correctly.
Equipnent and structures used for fuel handling activities are designed to provide for safe operation as described in FSAR Section 9.1.4.
In order to prevent accidental nuclear criticality, only one new fuel assembly will be allowed to be renoved frc;a a ship-ping container or an approved storage location at any one time.
Further discussion of the criticality of fuel assen-blies is found in FSAR Section 4.3.2.6.
Because of the fuel storage facilities design and admini strative controls limiting the maximum number of fuel assen-blies allowed out of the storage locations, the possibility of accidental criticality during receipt, ins pect ion, and other handling activities is eliminated.
Therefore, an exemption in whole fran the requirenents of 10CFR70.24 is requested as provided by 10CFR70.24(d).
The minimum qualifications for the key positions having l
nuclear criticality safety and fuel handling responsibili-ties are as follows:
A.
Mechanical Maintenance Department Mechanic First Class, Mechanic Second Class, Crane Operator, and Forenan are trained in accordance with the Maintenance Department training / testing progran in their respective positions.
All personnel have read and understand the Corrective Maintenance Procedure prior to receiving or handling new fuel assemblies.
Additional training requirenents can be seen for the Crane Operator in Attachment A.
Maintenance forenan qualifications are a high school diplona or equivalent and four years experience in the discipline to be occupied.
B.
Operations Quality Control Department Tne 0QC inspector is trained in accordance with the 0QC training progran on receipt and handling of new fuel and the position meets the requirenents of ANSI N18.1-1971.
See Attachment B for specific qualifications.
1 9
h W
1 1
C.
Nuclear Material Control The nuclear material control personnel qualifications are as follows:
The Plant Manager has been trained to be a Licensed Senior Reactor Operator and meets the requirenents of ANSI N18.1-1971.
The Director of Administrative Services is respon-sible for reading, understanding and following the Nuclear Material Control Manual. The qualifications are five years' experience in power generating stations and an educational equivalent to a gradua-tion fran an accredited college with a degree in Business Administration perferred.
The Director, Budget and Fuel Contract, is qualified in accordance with ANSI N18.1-1971.
Tne Nuclear Shift Supervisor is a Licensed Senior Reactor Operator and has met the appropriate quali-fications, passed the NRC examination and meets the requirements of ANSI N18.1-1971.
D.
Refueling Supervisor 1.
B.S. in Engineering or equivalent.
2.
Has a technical understanding of plant systens and operations.
3.
Has read and understands the Corrective Maintenance Procedure, the Fuel Accountability Procedure for receipt and handling of new fuel assenblies, and the refueling Technical Specifications.
4.
Meets the requirements of ANSI N18.1-1971.
E.
Westinghouse Service Representative (NFD)
Westinghouse personnel qualifications are based on experience gained in the field at operating PWR nuclear plants and at the Service Center Facility.
(See Attach-ment E for specific qualification.)
F.
Radiation Control The Radtech will be trained in accordance with the Radcon Manual.
(See Attachment D.)
The BVPS training program is based on government regula-tions, ANSI Standards (N18.1-1971), and other training 1
10
~
J
.. ~ _
g documents considered by Duquesne Light Company (DLC) as desirable for maintaining a knowledge and well-trained i
organization.
The program _ conforms to the intent of Regulatory Guide 1.8, Revision 1-R, Septenber 1975, Personnel Selection and Training.
i' The responsibilities for the key personnel responsible for nuclear criticality safety and fuel handling are as follows:
A.
The Mechanical Maintenance Department is responsible for the handling and unpackaging of the fuel assenblies and their shipping containers.
i B.
The Operations Quality Control Department is responsible for the inspection and acceptance of every fuel assembly
~
received onsite.
C.
The Director of Administrative Services or his designee i.
shall be responsible for the reciept and retention of all Fuel Receiving Inspection Reports.
(See Attachment C.)
4 i
D.
The Refueling Supervisor insures that the fuel is moved in accordance with applicable procedures.
E.
A Westinghouse Service Representative (NFD) will be made l
available onsite to witness the receiving and inspection of the fuel assemblies and provide technical support if necessary.
F.
The Radiation Control Group will make a radiological survey of the material and prescribe handling restric--
tions, if any, in accordance with Radcon procedures.
Specific actions are identified in Attachment F.
G.
The Radiological Control Manager is in charge of radia-l tion control during fuel movement.
This position meets j
the requirements.of RG 1.8, Rev. 1-R, September 1985.
1
-2.3 ' Accident Analysis l
Interlocks and administrative controls, prevent the Fuel Building r
handling equipment capable of carrying loads heavier than a fuel assembly from traveling over the fuel-storage area. The spent fuel storage racks are designed.to maintain a safe geometric spacing of fuel! assemblies despite the impact of a fuel assembly dropped from
- the maximum lift height of the spent fuel handling hoist.
~ Because the license will involve only the handling and storage of 4
unirradiated reactor fuel and the concamitant absence of any fis-sion products in the fuel' handling areas, there would be no signif-icant ' safety hazard as a result of a fuel handling accident. Should i
a fuel-handling accident result in the release of any of the J
11
1 uranium contents of the new fuel, Health Physics personnel would implement the appropriate radiological protection measures.
3.0 OTHER MATERIALS REQUIRING NRC LICENSE 3.1 Neutron Detectors The name, anount, and specifications of other speci al nuclear materials which OLC proposes to store are as follows:
U235/ Detector Total Total Name of Device Model Number Quantity Net. Weight Activity U235 (gram / unit)
(uCi)
(gram)
Flux Mapping Moveable In-core Detectors (WL-23680) 10 0.0041 0.09 0.041 R.G. 1.97 Neutron Detec-tor Systen (Unknown) 2 42 0*
84
- Deactivated U235 The incore detectors will be stored initially in the BVPS-1 New Fuel Storage Facility prior to installation at BVPS-2.
Security 1
and accountability practices consistent with NRC requirenents have been established and are maintained in accordance with procedures specifically developed for this purpose.
The General Manager, Nuclear Service Unit meets the requirentns of ANSI N18.1-1971 and has the overall responsibility for implenentation of the security and accountability practices and radiological control procedures.
Security and Radiological control procedures have been established to prevent access to the detectors by un authorized personnel.
Additional ly, personnel radiation exposures will De controlled by issuance of Radiological Work Permits when access to or work with the detectors is required.
1 The incore detectors will ultimately be used as part of the Incore Instrumentation Systen. The Incore Instrumentation provides infor-mation on the neutron flux distribution and fuel assembly outlet tenperatures at selected core locations. From this infonnation the core power distribution can be detennined.
The moveable flux detectors can traverse the entire length of selected fuel assem-blies, thus providing an extrenely accurate, three-dimensional map of the neutron flux distribution.
The Neutron Detector Systen is a two channel systen intended to monitor neutron flux according to Category 1 guidance of R.G.1.97, Revision 2, and satisfies 10CFR50 Appendix R guidance.
The systen is designed to operate over a range of 10-8 to 100 percent power and renain operable for a least one month following a design basis accident. Each channel includes a fission chamber neutron detector, 12
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preamplifier, processing and di splay un i t, and interconnecting cables and connectors.
3.2 Primary Source Rods The following infonnation is for the two (2) primary source rods used to provide a minimum base neutron count rate of two counts per second at the nearest source range detector for at least six months after the initial start-up of the first reactor cycle.
Following incore service, the primary source assemblies are stored in the spent fuel storage area or other appropriate facilities.
Upon receipt, the primary source rods will be stored at the BVPS-1 New Fuel Storage Facility.
Security and accountability practices consistent with NRC requirenents have been established and are maintained in accordance with procedures specifically developed for this purpose.
The General Manager, Nuclear Service Unit has the overall responsibility for implementation of the security and accountability practices and radiological control proced ures.
Security and radiological control procedures have been established to prevent access to the primary source rods by unauthorized personnel.
Addi tionally, personnel radiation exposures will be controlled by issuance of Radiological Work Pennits when access to or work with the source rods is required.
1
- Element and mass number per source Atom %
Californium - 254
<0.025 253 0.500 252 82.500 251 2.000 250 11.000 249 4.000
- Chemical and physical fonn of source material Chemical: Paladium-Californium 0xide (Pd-Cf 0 )
23 Physical:
Cennet wire encapsulated within stainless steel tubing
- Name of manufacturer and model nunber of sealed source Manufacturer:
Monsanto Research Corporation Model Nuaber:
A2765-AA00
- Maximan number of millicuries in sealed source Approximately 100 uCi's
- Maximun activity per source which will De possessed at any one time-3.1 x 108 neutrons per second (based on a maximurn of 113 ug of Cf-252) l 13 i
- Container in which sealed source is to be stored and model number The container is a shipping cask approximately 60" x 40" x 195" which weighs approximately 9 tons.
The model number is Model 2501, USA Dot SP5916 Type B or Model 2511 Type A.
3.3 The following information is for the various detector calibration and check sources (sealed) and monitors used in the Digital Radia-tion Monitoring Systen. Upon receipt, this material will be stored in Storeroon No. 22 (non-exempt sources) and Warehouse "D" (exempt I
sources) in accordance with ANSI N45.2.2.
Security and account-1 ability practices consistent with NRC requirements have been establi shed and are maintained in accordance with procedures specifically developed for this purpose.
The General Manager, Nuclear Services Unit has the overall responsibility for implemen-tation of the security and accountability practices and radiologi-cal control procedures.
Radiological control procedures have been established to prevent access to the check sources by unauthorized personnel.
Additionally, only qualified radiological control personnel are authorized to utilize the check sources in a manner consistant with established practices to protect personnel fran radiation exposure.
This material is being supplied by GA Techno-1 logies, Inc., and is shipped in DOT-approved containers.
- Various area and process monitor detector calibration and check sources that are individually in exempt quantity (each less than 10 microcurie Cs 137):
One (1) Wide Range Gas Monitor (Monitor Type 5001 WRGM Detection) 1 with two (2) 100 microcurie Cs 137 check sources For calibrating the area monitor detectors:
Two (2) each 10 millicurie Cs 137 sources Two (2) each 100 millicurie Cs 137 sources Four (4) Liquid Monitors (Monitor Type 2301) each with 100 microcurie Cs 137 sources 1
For calibrating the process monitor detectors, the calibration kits will contain:
Two (2) each 50 microcurie Ba 133 sources Two (2) each 100 microcurie Cs 137 sources Six (6) adjacent-to-line monitors:
- Five (5) check sources in exempt quantity
- 0ne (1) check source containing 100 microcurie Cs 137 1
One (1) Steam Line Monitor with check source (< 100uCi Cs 137) 3.4 Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma enitting material or 5 microcuries of alpha enitting material shall be free of 2. 0.005 microcuries of renovable contanination.
This limit, based on 14
10CFR70.39(c) limits for plutonium, will ensure that leakage fran byproduct, source, and special nuclear material sources will not exceed allowable intake values.
The detectors, source rods, and calibration and check sources will be retained in their shipping crates (the crates will be resealed if required to be opened.for receipt inspection).
Each of the detectors, source rods, and calibration and check sources shall be surveyed for contanination and/or tested for leakage upon receipt and prior to use or transfer to another user unless they have been
' leak-tested within six months prior to the date of use or transfer.
Records of results of the contanination surveys will be maintained
- for inspection.
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Attachment A, DUQUESNE LIGHT COMPANY Figure 0.7 Beaver Valley Power Station Training Manual LESSON PLAN Crane Operator Training.
8 Course Course Ecurs D. Brown
/- 2 6~- 81 1
Instru tor Date
/gAI LP-MM-043 (Ralph No. 0512)
"- Approved By:
Lesson Plan No. (Sequentially From 1)
References To Be Quoted:
ANSI B30.2.0 -1976 NUREG -612 DLCo Accident Prevention Manual
.27 CTR 1910 (OSHA) l 1
Items Issued:
(At.ach copy of all passouts, quizzes, etc.)
Crane Operator' Handout, Crane Operator Quiz In. trod.u.ctio.n.:..
1.
Purpose:
To instruct potential crane operator in general crane safety, rigging, standard hand signals and crane operation.
2.
Motivation:- (Discuss how you plan to nocivate students)
Provide enough instruction for potential crane operators to perfor:n safely at.d proficiently 'when operating a crane.
3.
General outline: (T.ist detailed outline Section I)
See attached sheet.
4 General Student Goals:
('ist detailed student objectives Secticn !!)
Crane coeratse cendue:. handling a load. orceer r12:inz, inseettien of j
- snes and rigging ecuirnant. s:2ndard hand signals.
I ISSd. 3
e Attaghment A - Page Two General Outline - Crane Ooerator Training I.
Classroom A'.
Introduction 3.
Phase I, General Crane Safety 1.
Videotape, Phase I 2.
Lecture, Crane Safety a.
Crane control 1)
Cab-controlled crane 2)
Pendant controlled crane b.
Crane inspections - Figure 2 of handout c.
Appemi12 I and II review d.
Safety 3.
Slides and description 4.
Students read handout - Parts I and II C.
Phase II, Rigging 1.
Videotape, Phase II 2.
Lecture s.
Weight of Lift b.
Rigging equipment inspection c.
Safe working loads d.
Rigging techniqu'es e.
Safety 3.
Students read handout - Part III D.
Phase III, Standard Hand Signals 1.
Videotape, Phase III 2.
Lacture a.
!=por: ant of Signals r
~v r-,a, nn-----n,-
---,-,----w-wv+
Attachment'A - Page Three Canarril Outlinn - Crana Oesrator Traininz b.
Review of signals Re-emphasize comEunication between signal: nan & crane operator c.
~
3.
Students read handout - Parts IV, V and VI 4.
Finish Videotape - Short Segment 5.
Review 6.
Quiz II.
Practical A.
Operate cranes 3.
Complace Check-off (Attached)
I l
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Attachment B
'CCC 10.2 Fev. 4 Page 10.3 2e fc11owing lesson plan shall be ctmupleted prier to perfcz=ing inspection of new fael shipants for eacn refaeling.
It applies only to 00C persennel performing the asceipt rrw.icn function and is not intended to nullify the perfczmance of this r M ure should any part boccme unavailable.
Personnel Prerecuisites and Limitations Car ified Invel I IrMor in asceiving Inspection functions to i g lement r d ures and record data.
Car.ified Invel II Inspecter in Deceiving Inspection fanctions to evaluate acceptability of irw.icn results and to report results.
1 i
Recuirac Readi.xt
' 1.
OQC Procedure 10.2, current revision, " Fuel Assembly and Shipping Container Receipt Tr % =.
2.
CMP 1-75-286, " Site Receipt and Handidng of New Assemblies Fev.
and 2. sir Shipping Containers".
4 3.
Mditional reading when ewelata and as assigned by the l
G mgineer.
4 i
Reviews i
1.
View video tape of past Beceipt Inspection of new fael shipnents 2.
Mditional tral. ting, such as that supplied by W, *.en apprg:riate 4
and as assigned by the QC Engineer.
1 i
ATTACHMENT C NUCLEAR MATERIAL CONTROL ORGANIZATION (Excerpt from the Beaver Valley Nuclear Material Control Manual)
A.
The Plant Manager of Beaver Valley Power Station, who reports to the General Manager, Nuclear Operations, has overall responsibility for the custody and physical control of nuclear materials at the station and for the implementation and enforcenent of the nuclear material control and accounting system.
B.
The Director, Administrative Service, who reports to the Manager, Administrative Services, is responsible for managing the nuclear material control and accounting system.
C.
The clerical force, who reports to the Director, Administrative Services, maintains the nuclear material control and ' accounting records, prepares reports and participates in the control of nuclear material as directed by the Director, Administrative Services.
D.
The management of nuclear material control and accounting systen and the management of the operating organization reports independently to the Plant Manager. This provides a separation of functions so that the activities of one individual or organization serve as controls over and checks of the activities of other individuals or organizational units.
E.
The responsibilities of the nuclear material control organizations members are assigned by position title.
In absence of such individ-uals, the assunption of their responsibilities by their designated alternates, as described in Site / Station Administrative Procedures, is implied and will not be specifically identified hereinafter.
F.
The Director, Budget and Fuel Contracts, who reports to the General Manager, Nuclear Services, cooperates with the Plant Manager in the responsibility of nuclear fuel management. He has overall responsi-bility for the procurenent, utilization, final disposition of nuclear fuel, and offsite fuel accountability. He negotiates all contracts with suppliers and reprocessors-for the purchase, conversion, enrich-ment, and fabrication of nuclear fuel. He arranges for all aspects of transportation of new and spent fuel and the final disposition or reclaimed nuclear fuel. He directs these activities through the Nuclear Fuel Management Coordinator.
G.
The Nuclear Shift Supervisor must be aware of all fuel moves.
H.
The Safety and Licensing Engineer, who reports to the Nuclear Fuel Management Coordinator, provides the total element and isotopic conpo-sition of nuclear material in each irradiated nuclear fuel assenbly.
This information is derived fran incore measurenents perfonned through the supervisor of Testing and Plant Performance, who reports to the Plant Manager.
}
B.V.P.S. - R.C.M.
Attachment D CEAPTER 1 - STANDARDS AND REQUIRDENTS B.
RADIOLOGICAL CONTROL RESPONSIBILITIES AND TRAINING 1.
Essponsibilities The BVPS Administrative Procedures delineates the specific responsibilities of the Company and personnel assigned to BVPS.
2.
Training a.
Training and retraining is provided for the BVPS Radiation Technicians (RT's) to ensure their proficiency is maintained.
The training provides a format for review of the current status of procedures, regulations and requirements. The training requirements are set forth in the BVPS Training Manual. Training is conducted in accordance with the BVPS Training Manual.
l b.
Radiation Workar Training Radiologiesi training is provided to each individual assigned.
to BVPS who is designated as a Radiation Worker. This training is to fulfill and enhance the worker's knowledge and attitude toward radiological safety. Each Radiation Workar is trained to the level indicated in BVPS Training Manual.
l This training is commensurate with the duties and responsibilities of the Radiatien Workar.
c.
Training of Others i
Each person who is assigned to BVPS, is given BVPS Power Station indoctrination training prior to assuming duties at BVPS as described in the BVPS Training Manual. This indoctrination includes familiarization with the Emergency Preparedness Plan. It includes familiarization with the location of the Controlled Area, Radiation Areas, and the Emergency Preparedness Plan Assembly Areas. Training is also provided in the proper use of personnel dosimetry.
Personnel whose duties do not necessitate their entering a controlled area are made aware of the reasons for kaaping out of controlled areas.
Prior to being issued dosimetry equipment, the female Radiation Workar is given specific instructions about prenatal exposure risks to the developing embryo and fetus. This instruction shall include both orally and in writing the applicable infor=ation in NRC Regulatory Guide 8.13.
1 i
_...eu._...__..___,.._._..~.,...__.__.,,.._.___._..______
Attachment D - Page Two
)
BVPS-RCM I.C.
CHAPTER 1 - STANDARDS AND REQUIREMENTS C.
COMPLIANCE WITH FEDERAL AND STATE REGULATIONS 1.
Periodic Radiolonical Audits Audits of the Radeon Program and radiological posture of BVPS are conducted periodically.
2.
Revtew and Update of Radeon Progr'an The BVPS-RCM is developed basically in accordance with 10 CFR 20, the BVPS Technical Specifications, and other applicable regulations and 4
guidas. It may be revised where necessary and whenever a change or clarification of any applicable regulation occurs. The BVPS Administrative Procedures set forth the requirements for such revisions.
Supplementary information is contained in BVPS-RCM Appendix 1.
3.
As Law as taasonably Achievable Guidelinas The Company endeavors to maintain reviews, plans, procedures and practices consistant with the intent of Regulatory Guide 8.8 and 8.10 to maintain all doses as low as reasonably achievable (ALAEA).
All Radiation Workers I
at BVPS are instructed of their responsibilities in this regard and are directed to follow practices which =4a4=4*e their exposure and the exposure of the general public to radiation and radioactive material.
NOTE: This provision indicates an objective and is not -intended as a literal commitment to each specific item in Regulatory Guide 8.8 and 8.10.
t l
l ISICE 3
Attachment E To: Jenkins. J (48:WES3064)
From: McKenzie. B.D (WST5166) Posted: Mon 1-Apr-85 12:45 EST Sys 49.(1
Subject:
FSS Qualification 4
Qualification For A FSS Engineer (Westinghouse Service Representative) 1.
Indoctrinated and trained in FSS activities 2.
Passes physical exam including vision exam 3.
Reviews and becomes f amiliar with applicable product Assurance Procedures and Specifications 4.
Completes Training program which includes
- Tour of manufacturing facilities
- Familarization with inspection requirements
- Supervision by qualified FSS engineer during fuel receipt, core loading and unloading 5.
Completes REM program certification t
i l
B.V.P. S. - R.C.M.
Attachment F CHAPTER 3 RADCON PROCEDURE 1.8
' )
RECEIPT OF FUEL 1.0 Purpose 1.1 This RP describes the Radeon practice to be followed at BVPS upon receipt of reactor fuel.
2.0 Prerequisites, Precautions and Limitations 2.1 Initiate a Radiological Work Pennit for the fuel receipt.
2.2 Place a continuous air monitor (AMS-3 or equivalent) in operation in the Fuel Handling Building at elevation 735' near the fuel inspection rack.
2.3 Rope off and post unloading, storage and other work areas, as required, for access control.
2.4 Utilize RCM Fonn 1.6, Fuel Receiving Record, to docunent container and fuel suney data.
2.5 When perfonning survey on vehicles, shipping containers and fuel, notify Radcon supervision of any contamination levels that exceed BVPS clean limits.
Survey for beta-gamma and al pha contamination.
3.0 Procedure 3.1 Shiament Arrival 3.1.1 Upon notification of the shipnent arrival, perfonn a survey of the shipment yehic1e in accordance with RP 3.9, Monitoring Vehicies. Utilize RCM Fonn 3.4.
3.2 Unloading Containers From Vehicles
- 3. 2.1 Perfonn a radiation and contamination survey on the external surfaces of each shipping container. Docunent the suney data on RCM Fonn 1.6.
3.3 Rel ease Of Vehicles 3. 3.1 Suney enpty shipment vehicles in accordance with RP 3.9, prior to i
releasing it fran the site.
3.3.2 If the outgoing vehicles are to transport anpty containers, a survey of the anpty containers is required.
3. 3. 2.1 Ensure the enpty shipping containers are 1abeled " EMPTY."
3.3.2.2 Renove or cover any unnecessary vehicle radioactive markings, e.g.,
" Radioactive" placards.
-l -
ISSUE 2 l
C.M.
. Att'achment F - Page Two RP 1.8 - RECEIPT OF FUEL 3.3.2.3 Consult the vehicle driver to determine what docunentation is required, "s
e.g., vehicle / container survey data, bill of lading, etc.
3.4 Opening Shipping Containers Carrying New Fuel Provide Radcon konitoring during the opening of each shipping container 3.4.1 and transfer of each fuel assably.
3.4.2 Perfonn a contanination survey on the interic: surfaces of each shipping container. Document the survey data on RCM Fonn 1.6.
3.4.3 Label enpty shipping containers with " EMPTY" label s.
3.5 Monitoring New Fuel
- 3. 5.1 Perfonn a radiation and contamination survey on each fuel assenbly upon renoval of the covering. Docunent the survey data on RCM Fonn 1.6.
NOTE: This survey is to be perfonned before an assenbly is handled by other personnel.
3.5.2 As appropriate, affix radiolegical labels / stickers on the fuel assenbly covers. Do not affix labels or stickers on an uncovered assenbly.
4.0 Reconds and Fonns 4.1 RCM Fonn 1.6, Fuel Receiving Record RCM Fona 3.4, Vehicle Radiological Status Survey Map 4.2 4.3 Other survey maps as required.
. ISSUE 2 i
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FIGURE 3.1.8.1 ISSUE 2
.- 3_
RE11SI0fi 1.
L
ATTACHMENT c POSITION:
SUPERVISOR, REFUELING POSITION RESPONSIBILITIES 1.
- Plans, coordinates, and directs all station refueling activities.
Coordinates with the refueling shift supervisor (SRO Licensed), the vendor refueling group, and refueling shift coordinators prior to and during fue1 movements.
2.
Directs refueling preparations and identifies potential problems that impact refueling.
Resolves problems through technical
- skills, knowledge, and the requisitioning of vendor technical services.
Works with vendor equipment designers and technicians to provide solutions and repair to the equipment.
3.
During refueling outages dictates the sequence of containment polar crane operations, on a
day-to-day
- basis, to assure-the refueling operations are carried out in a timely fashion.
4.
Identifies and arranges for all the necessary refueling manpower.
Coordinates activities with the Maintenance Supervisor for maintenance support.
Contacts vendors to support refueling and for necessa ry repairs and modifications for associated systems, such as the fuel transfer system and fuel pool.
5.
Coordinates refueling efforts with the supervisors of Radiological Operations, Maintenance, Operations,
- Training, Licensing and compliance, Ouality Assurance, Security, Chemistry, and Material Control.
6.
Provides technical input to Muclear Engineering and Construction Unit and the various other engineering organiza tions in the
- repair, development, and modification of the refueling equipment.
7.
Ensures that all procedures required to support refueling operations are written, approved, and used as described.
8.
Requisitions and expedites materials and services needed to support the refueling objectives.
9.
Coordinates the receipt and storage o' new fuel.
10.
Advises canagerent of the stTtus of refueling activities and identifies those activities wit 1 potential i pact to the refueling outage completion.
- ~ _ - - -
i 11.
Responsible for preparing the operations and maintenance
- budgets, monitors expenditures, and provides a
reconciliation of deviations.
In addition, provides information for the preparation of the capital improvement budget.
12.
Assures that personnel within his cognizance are qualified to perform their duties and fully ' understand
~
their responsibilities.
13.
Directs the preparation of post refueling outage critiques.
14.
Administers the Operations qualifty Assurance Program as it applies to refueling.
15.
Administers the industrial safety and housekeeping programs as applicable.
16.
Institutes and maintains a
program for employee performance evaluation and development.
17.
Performs other duties as assigned.
POSITION QUALIFICATIONS:
1.
Bachelor Degree in Engineering or equivalent.
2.
Four (4) or more years experience in nuclear power plant technical maintenance or operations work.
3.
Demonstrated ability to solve problems of a technical na ture.
4.
Ability to plan and direct the work of others.
5.
Demonstrated ability to prepare procedures and reports of a technical nature.
6.
Experience in managing or supervising rajor nuclear outage activities.
7.
Demonstrated ability to plan and coordinate numerous activities to meet an overall objective.
l
.