ML19259B278
| ML19259B278 | |
| Person / Time | |
|---|---|
| Issue date: | 01/12/1979 |
| From: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Jacobs S STONE & WEBSTER, INC. |
| References | |
| REF-PROJ-M-1 NUDOCS 7901240049 | |
| Download: ML19259B278 (22) | |
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UNITED STATES j
1 NUCLEAR REGULATORY COMMISSION
,1 WASHINGTON, D. C. 20555
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JAN 121979 Project M-1 Mr. S. B. Jacobs Chief Licensing Engineer Stone and Webster Engineering Corporation P. O. Box 325 Boston, Massachusetts 02107
Dear Mr. Jacobs:
Our letter of July 12, 1978 provided you with the staff's evaluation of the Stone and Webster Engineering Corporation Topical Report SWECO-7601, " Interim Spent Fuel Storage Facility." On the basis of this evaluation, we concluded that the conceptual design described in the topical report, in conjunction with the additional information to be provided by a utility applicant, should allow the staff to conduct a safety review of a specific application in a timely manner.
The purpose of this letter is to identify, on the basis of the staff's evaluation, the subsectiors of SWECO-7601, as modified through Revision 4 (December 8,1978), that may be referenced in future applications as design requirements for construction of independent spent fuel storage installations. We find that the design requirements specified in the following subsections are acceptable for referencing in applications for the proposed construction of such installations:
Subsection Ti tle 3.2 Classification of Structures, Systems and Components 3.3 Wind and Tornado Loadings 3.4.2 Water Level (Flood) Design - Analysis Procedures 3.5 Missile Protection - Barrier Design Pro-cedures 7 9 012 4 0 0 YT
Mr. S. B. Jacobs JAtt 2 & i373 3.6 Seismic Design 3.7 Design of Category I Structures 8.1 Radiation Exposure 8.2 Radiation Sources 8.3 Radiation Protection Design Features 8.4 Dose Assessment 13.1 Quality Assurance During Design and Construction Enclosed for your information are copies of the safety evaluation report (SER) inputs for the subsections identified above which discuss the bases for the staff's conclusions.
As we noted in our letter of July 12, 1978, the reliance on parent plant facilities and the need for further information to be provided by a future applicant raises a number of interface items with respect to the design described in SWEC0-7601. The interface information supplied in Table 1.5-1 of the report and Appendix A of the enclosed SER sections adequately identifies the items which need to be addressed in an application. To assist the staff in a safety review of a site-specific installaticn, we request that an application that references the identified sections of SWECO-7601 include Table 1.5-1, modified to cross-index the location in the application of the required interface information.
You may wish to consider issuing and submitting a revised version of SY:CO-7601 to include this letter and enclosure accepting the specified suesections of the report.
It is the staff's intent that the design requirements specified in these subsections need not be reevaluated when referenced in a site-specific application for construction of an
i Mr. S. B. Jacobs,_,
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independent spent fuel storage installation. We will inform you, to tha best of our ability, if it appears that our present conclusions concerning SWECO-7601 are invalidated by subsequent changes in Nuclear Regulatory Commission (NRC) criteria or regulations in order to enable you to revise the report and resubmit it if you so desire.
Nevertheless, you are advised that the responsibility for recognizing the impacts of any such published changes in NRC criteria and regulations on your design rests with you. We will be pleased to meet with you or your staff at any time to discuss such matters or questions you may have.
Sincerely,
/
Leland C. Rouse, Acting C ief Fuel Reprocessing and Recycle Branch Division of Fuel Cycle and Material Safety
Enclosure:
As stated
L APPEN0!X A INTERFACE INFORMATION Section of SWECO-7E,01 Applicant Responsibilities Interface Requirements Chapter 1 - Introduction & General Description 1.1 Introduction supply introduction to facility.
1.2 General facility Description Supply facility description.
1.3 Identification of Agents and Contractors Supply description of agents and contractors.
1.4 Naterial Incorporated by Reference Supply material incorporated by reference.
Chapter 2 - Site Characteristics 2.1 Geography and Demography Describe site population and population distribu-Facility must be located within tion, and use of adjacent lands and waters.
existing site exclusion boundary.
2.2 Heasby Industrial Transportation, and Evaluate effects of such facilities.
Hilitary facilities
- 2. 3 Heteorology Provide site-met data established for the adjacent Show conformance with envelope of parent facility including data collection program, site characteristics as described site specific short term and Icng ters diffusion in section 2.3.
Information, rainfall, high air pollution potential, and relative humidity.
2.4 Hydraulic Engineering Provide Hydrologic Engineering data established for Show conformance with Hydrologic parent f acility including flooding; PMF; da:s envelope of site characteristics failure; Probable maximum surge and seiche flooding; described in section 2.4.
Proable maximura Tsumaat, ice flooding; low water; envitonmental acceptance of effluents; groundwater; lech Specs and' Emergency Operating requirements.
Provide site drainage system. Discuss flood protec-tion and the use of cams or dikes if the design flood is higt.er than yard grade.
A-1
t I
d Section of SWECO-7601 Applicant Responsibilities Interface Requirements 2.5 Geology, Seismology, & Geotechnical Supply site geology and seismology data including Show conformance with Geologic Engineering discussion of Vibratory Ground motion, surface envelope as described in faulting, stability of subsurface materials, and section 2.5.
slope stability.
Chapter 3 ' Design of Structures, Components, Equipment, and Systems 3.1 Conformance with Applicable Criteria Identify and evaluate codes and standards to determine applicability.
Design wind velocity for seismic 3.3 Wind and Tornado Loadings Cat. I structures 120 mph.
3.4 flood Protection Supply above grade flood protection if applicable.
3.5 Hissile Protection Present turbine missile strike probabilities and probability of penetration.
Justify ductility ratio >10 for concrete barriers if required.
3.6 5elsmic Design Supply additional analyses for any deviations from presented design.
Supply subgrade characteristics and verify design aJequacy.
Justify higher seismic dasping values if necessary.
Provide seismic analysis of dams if needed.
Supply details of seismic qualification program.
Supply values of Poisson's ratio for site materials.
A-2
I Section of SWTCO-7601 Applicant Responsibilities Interface Requirements 3.7 Design of Category I structures Verify seismic design for soil-structure interaction.
Supply description of any site related variations for materials, quality control, and special construction techniques.
Chapter 5 - Electric Power 5.1.3 Design Criteria Identify extent of compliance to National
. Electrical Codes depending on f acility locatitu and component classification.
5.2 Offsite Power Describe utility grid.
See Table 1.5-2.
Describe of fsite power system.
5.4 Communication Supply description of 15FSF to parent facility Commercial telephone and page/
and 15FSF to offsite communications.
party public address and evacua-tion system to parent facility.
Chapter 6 - Auxiliary Systems 6.1.1 Spent Fuel Storage Select spent fuel storage rack design.
NSS vendor's spent fuel storage requirerents.
Provide criticality analysis, demonstration of Hixed oxide fuel shall not be structural integrity, administrative control allowed without reanalyses.
to prevent inadvertent criticality, and identi-fication of personnel qualification program.
6.1.2 Spent Fuel Pool Cooling and Supply description of Seismic Category I source See Table 1.5-2.
Purification of fuel pool makeup.
Supply Fuel Pool Vacuum System.
O A-3
Section of SVECO-7601 Applicant Responsibilities Interface Requirements Select type of purification filter, cartridge Flitration requirements are or etched disc.
stated in section 6.I.2.
6.1.3 Fuel Handling System Select the fuel handling bridge.
Meet the ISFSF arrangement envelope and seismic require-ments of section 3.8.2.
6.2.1 Facility Cooling Water Elect installation of dry internals cask See Table 1.5-2.
cooldown system if needed.
Determine procedure for handling defective fuel.
6.2.2 Demineralized Water Supply description of demineralized water makeup See Table 1.5-2.
including water treatment equipment. Determine makeup rate required and demineralizer water chemistry.
6.2.3 Potable and Sanitary Water Supply description of system.
6.5.1 Water Fine Protection Supply description of water source, liaison with See lable 1.5-2.
local fire force, personnel training and qualification.
Chapter 7 Radioactive Waste Management 7.2 Liquid Waste Hanagement Supplydescriptionofpipingconnectingfacility See Table 1.5-2.
to parent facility and parent facility s liquid waste capabilities.
Supply description of facility effluent handling Effluent activity and volume are and dilution factor.
dependent on available dilution factor and effluent handling.
Provide dose estimates.
Effluent W.e is 20 gpm.
- 7. 3 Solid Waste Management Supply description of resin and miscellaneous See Table 1.5-2.
solid waste onsite transport, handling, and storage.
Provide dose estimates.
A-4 l
e t
Section of SWEC0-7601 Applicant Responsibilities Interface Requirements Chapter 8 - Radiation Protection Supply description of operating procedures, health physics program, and radioactive material safety. Supply portable radiation monitors for use in emergency conditions.
Chapter 9.- Conduct of Operations 9.1 Organizational Structure of Applicant Supply owner's organizational structure.
9.2 Training Supply training program.
9.3 Emergency Planning Supply site related emergency planning.
9.4 Review and Audit Supply review and audit procedures.
9.5 Facility Procedures and Records Supply administrative, operation, and maintenance procedures.
/
9.7 Industrial Security Supply information on parent facility security.
Common site exclusion boundary Describe organization, administration, and with parent facility.
conduct of security program; personnel selec-tion, training, and evaluation procedures:
and types of detection systems, access control Provide a reliable power supply systems, and site related design features.
.for security system.
Chapter 10 Initial Test Program Supply description of test program.
Chapter II Accident Analysis Supply description of site specific accidents.
Chapter 12 License Conditions Provide final specifications and administrative controls, exact number of fuel assemblies the transuranium isotope limits.
Chapter 13 Quality Assurance Supply description of quality assurance program.
Program must be consistent with Topical Report SWSQAP.
4 A-5
i 3.0 DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 3.1 Conformance with NRC General Design Criteria The design of tha ISFSF was reviewed in relation to applicable parts of Appendix A of 10 CFR 50, General Criteria for Nuclear Power Plants.
In addition, the ISFSF design was compared with Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Regulatory Guide 3.24, Guidance On The License Applica-tion, Siting, Design, and Plant Protection For An Independent Spent Fuel Storage Installation, and ANSI Standard N305-1975, Design Objectives For Highly Radioactive Solid Material Handling and Storage Facilities In A Reprocessing Plant.
In this regard, we note several differences between the ISFSF design and the Regulatory Guides.
Regulatory Guide 3.24 states that an " independent spent fuel storage installation" has its own support facilities and operates independently of any other facility.
The ISFSF design relies on a parent facility for liquid and solid waste processing, a source of demineralized water for pool make-up and cask purging and process water for fire protection and cooling tower application.
The staff has evaluated these exceptions and has concluded that they are acceptable.
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Regulatory Guide 3.24 recommends that large spent fuel storage pools should be built as a series of separable modular units or with provisions for isolating sections of the pool when necessary.
The ISFSF fuel pool is designed as a single pool.
The applicant has evaluated the single storage pool design and has concluded that no significant safety or operational consi-derations could be postulated which would preclude using the single storage pool design.
The staff is in agreement with the applicant's conclusion.
Regulatory Guides 3.24 and 3.13 state that a controlled leakage building should enclose the fuel pool. While the ISFSF is not designed for controlled in-leakage, the ventilation system as designed is capable of maintaining exhaust flows along designed pathways even though one of the large access doors is open.
The staff conclusion, based on our evaluation of the system, is that the design is adequate to protect employees as well as the public from any release of radioactivity.
. Regulatory Guide 3.24 recommends that a cask drop analysis be made to determine whether a shock absorber is needed in the bottom of the pool.
The applicants cask drop analysis was made to determine the effect that a dropped cask would have on the cask rather than the pool.
The staff will require any site specific application to include an analysis which shows the effect of a cask drop on the cask pool as well.
(See Appendix A " Interface Information," Section 3.1 "Conformance with Applicable Criteria.")
Regulatory Guide 3.24 states that an automatic interlock with the high radiation level instrumentation should actuate the ventilation confinement system. The ventilation system in the ISFSF relies upon manually switching to actuate the ventilation confinement system.
The staff will require any site specific application to include an analysis to demonstrate that this operational mode is safe and also will require a license condition in any parent application to ensure that the ISFSF is manned at all times.
3.2 Classification of Structures, Systems, and Components Confinement structures, systems and components important to safety are designed to withstand the safe shutdown earthquake (SSE) and are classified as Seismic Class I.
Noosafety related structures, systems and components are classified as nonseismic.
The Seismic Category I structures in the ISFSF are the fuel pool including the cask pool area and the fuel storage racks.
The Topical Report (Table 6.1.1-2) references a number of spent fuel storage rack designs to verify the feasibility of storing approximately 1300 tonnes UO2 in the spent fuel pool as designed. This reference does not relieve this design from a full storage rack review for any site specific application.
The piping that connects directly to the fuel pool is seismically designed by the methods descrioed in Section 3.6 of SWECO-7601.
This includes portions of the piping in the fuel pool cooling and purification systems.
It does not include any valves or other components. All piping entering the storage pool has a vent hole drilled into it as an antisiphoning device.
This vent is located under the normal pool water level but above the spent fuel to prevent uncovering of the spent fuel.
In addition all piping peretrations into the storage pool are made at a height of at least 11 feet above the top of the
stored spent fuel to ensure that a water level adequate for shielding and cooling of the fuel is maintained.
The fuel pool bridge crane and the 130 ton crane are designed to remain on their rails, although not necessarily operative, during a seismic event.
Enveloping structures containing the purification equipment, backwash equipment, and fuel pool demineralizer are designed, constructed, and inspected in accordance with Branch Technical Position - ESTB No. 11-1 (Rv.
1), Design Guidance for Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Reactor Power Plants.
The design guidance given in this position provides reasonable assurance that equipment and components bsed in the radioactive waste management systems are designed, constructed, installed and tested on a level commensurate with the health and safety of the public and plant operating personnel.
The position also notes that instrumentation and controls associated with the waste management systems should be designed to a quality commensurate with their intended function.
The staff finds the above classification of structures, systems, and components acceptable.
3.3 Wind and Tornado Loadings
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All seismic Category I structures exposed to wind force will be designed to withstand the effects of the design wind.
(See Appendix A, Interface Information, Section 3.3 " Wind and Tornado loadings.") These structures include for this design thdse enclosing the cask pool and spent fuel storage pool areas.
The specified design wind has a velocity of 120 miles per hour based on a recurrence of 100 years.
The wind velocity will be transformed into pressure loadings on structures and into the associated vertical distribution of wind pressures and gust factors in accordance with ANSI A58.1-1972, Building Code Requirements for Minimum Design Loads in Buildings and Other Structures. The staff finds that the procedures utilized to determine the loadings on structures induced by the design wind siacified for the plant are acceptable since these proce-dures provide a conservative basis for engineering design to assure that the structures will withstand such environmental forces.
The use of these procedures provides reasonable assurance that in the event of design basis winds, the structural integrity of the plant structures that have to be designed for the
. design wind will not be impaired and, in consequence, safety-related systems and components located within these structures are adequately protected and will perform their intended safety functions if needed.
Conformance with these procedures is an acceptable basis for satisfying, in part, the require-ments of 10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
All-seismic Category I structures exposed to tornado forces will be designed to resist a tornado of 290 miles per hour tangential wind velocity and a 70 miles per hour translational wind velocity. The simultaneous atmospheric pressure drop will be assumed to be 3 pounds per square inch in 1.5 seconds.
The procedures that will be used to transform the tornado wind velocity into pressure loadings are similar to those used for the design wind loadings as discussed in ANSI 58.1-1972.
The staff finds the design of these structures acceptable in accordance with Regulatory Guide 1.76 Design Basis Tornado for Nuclear Power Plants which provides guidance on meeting the requirements of General Design Criterion 2, Design Bases for Protection Against Natural Phenomena of 10 CFR Part 50, Appendix A.
The forces cited conform in Regulatory Guide 1.76 to those cited for Region I in Table I, Design Basis Tornado Characteristics.
3.4 Water Level (Flood) Design The seismic Category I Structures will be protected from the potential damaging effects of flood water to the facility yard grade.
The hydrostatic and buoyancy effects of the flood will be considered in the design.
This is acceptable to the staff with the reservation that the probable maximum flood must be determined fur the site specific application and further that any dikes or other structures that may be required to protect the site will require a full review in any site specific application.
This is noted in Appendix A, Interface Information, Section 3.4, Flood Protection.
3.5 Missile Protection The purpose of the requirements expressed in Section 3.5.1.4, Rev.1, Missiles Generated by liatural Phenomena, of the Standard Review Plan is to assure that hazards due to missiles
. generated by the design basis tornado identified in Section 3.5 of the Topical Report are acceptably small so that they need not be included in the plant design basis and that appropriate desiga basis missiles have been chosen and properly characterized.
(See Appendix A, " Interface Information,"
Section 3.5, " Missile Protection.")
Because of the small amounts of volatile and gaseous radioac-tivity available for releases from the Interim Spent Fuel Storage Facility it is not necessary to protect the facility against tornado missile penetration but rather to preclude the gross failure and collapse of the building structures.
Design of the facility to accommodate the automobile would be one means of accomplishing this objective.
The applicant has committed to meet the requirements of Standard Review Plan 3.5.1.4, Rev. 1 by protecting against the selected missiles and their associated velocities identified in Spectrum II, shown below.
SPECTRUM II Velocity (m/sec)
MISSILE Mass (kg)
Dimensions (m)
Region I Region II Region III A Wood Plank 52
.092 x.289 x 3.66 83 70 58 8 6" Sch 40 pipe 130
.1680 x 4.58 52 42 10 C 1" Steel rod 4
.02540 x.915 51 40 8
0 Utility pole 510
.3430 x 10.68 55 48 26 E 12" Sch 40 pipe 340
.320 x 4.58 47 28 7
F Automobile 1810 5 x 2 x 1.3 59 52 41 The staff concludes that the proposed ISFSI design will meet these requirements and will provide adequate tornado missile protection for a facility of this type.
3.6 Seismic Design Appropriate seismic design procedures will be used to assure the ability of structures to withstand the effects of the Operating Base Earthquake (0BE) and the Safe Shutdcwn Earthquake (SSE).
(See Appendix A, " Interface Information" Section 3.6 and 3.7.)
The design value of the maximum ground acceleration is 0.3g for the SSE and 0.15g for the OBE. The response spectra input for the OBE and SSE applied in the seismic design of Seismic Category I structures (enclosing the cask pool and storage pool areas) and systems comply with the recommenda-tions of Regulatory Guide 1.60, Design Response Spectra for
Seismic Design of Nuclear Power Plants, and the specific percentages of critical damping values to be used in the seismic analysis are in conformance with Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants.
Modal response spectrum and time history methods will form the bases for seismic analysis for the elastic structures.
Procedures of combining modal responses are in conformance with Regulatory Guide 1.92, Combining Modal Responses and Spatial components in Seismic Response Analysis.
The finite element approach will be used to evaluate soil-structure interaction effects in a site specific application. This is noted in Appendix A, Interface Information, Section 3.7 Design cf Category I Structures, Item 1.
3.7 Design of Category I Structuras The principal codes to be used for the design of interim spent fuel storage facility structures are for concrete the ACI 318-71 code, " Building Code Requirements for Reinforced Concrete"; and, for steel structures the AISC Specification,
" Specification for the Design, Fabrication and Erection of Structural Steel." ISFSF structuras will be designed to resist various combinations, as applicable, of dead loads, live loads, and environmental loads including winds, tornadoes, OBE, SSE and load generated ruptures of high-energy pipes.
Acceptable environmental load limits have been discussed in Sections 3.3-3.6 of this report.
Beyond this design envelope site specific interfaces are denoted in Appendix A, Interface Information, Section 3.7 and will need to be addressed in any site specific application.
3.8 Comouter Proarams for Dynamic and Static Analysis:
In the Topical Report Appendix 3A entitled " Computer Programs for Dynamic and Static Analysis of Seismic Category I Structures" has not been reviewed.
Its inclusion is considered to be for information only.
However, the staff notes that the computer programs referenced have been cited previously in reports submitted to NRC. Where staff evaluation has been made in such cases, any future application may appropriately reference such evaluation.
3.9
==
Conclusions:==
Topical Report SWEC0-7601 deceribes the Stone and Webster Engineering Corporation design of an Interim Spent Fuel Storage
Facility and is intended to serve as a basic reference document.
The staff finds that sections:
3.2 Classification of Structures, Systems and Components 3.3 Wind and Tornado Loadings 3.4.2 Water Level (Flood) Design - Analysis Procedures 3.5 Missile Protection - Barrier Design Procedures 3.6 Seismic Design 3.7 Design of Category I Structures of Chapter 3 " Design of Structures, Systems and Components" are acceptable and may be referenced in future license applications.
The use of design criteria in SWECO-7601 as defined by appli-cable codes, standards, and specifications, the loads and loading combinations, the structural and seismic design and analysis procedures, the structural acceptance criteria, and the materials quality control and special construction techni-ques provides reasonable assurance that, in the event of winds, tornadoes, earthquakes and various postulated accidents occurring within the structures, these structures will with-stand the specified design conditions without impairment of structural integrity or the performance of required safety function. The staff has concluded that conformance with these criteria, codes, specifications and standards in the design of seismic Category I structures of an Interim Spent Fuel Storage Facility constitutes an acceptable basis for satisfying the requirements of Criteria 2 and 4 of the General Design criteria.
8.0 RADIATION PROTECTION The Stone & Webster Topical Report, " Interim Spent Fuel Storage. ___
Facility (SWECO-7601)", provides information on the methods for radia-tion protection, including facility design and layout, and equipment design.
Information relevant to the health physics program is to be provided by the applicant for each site specific plant.
(See Appendix A,
" Interface Information" Chapter 8, " Radiation Protection.") An estimate of occupational radiation exposure to plant personnel is included.
Shielding will be provided to reduce radiation levels. Ventilation is arranged to control the flow of potentially contaminated air.
Radiation monitoring is employed to measure levels of radiation in potentially occupied areas, and to measure airborne radioactivity throughout the plant.
The staff reviewed and evaluated the description and analysis of the radiation protection program in the Topical Report and responses to its questions.
The criteria used to determine acceptability of the program are that doses to personnel will be maintained within the established limits of 10 CFR Part 20, " Standards for Protection Against Radiation" and that features are consistent with the guidelines of Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as low as is Reasonably Achievable" USNRC, March 1977, to the extent that these guidelines apply to the design of a fuel storage facility.
SWECO-7601 includes extensive material regarding design features introduced for the purpose of assuring that occupational radiation exposures will be as low as is reasonably achievable.
SWECO-7601 describes means to keep external and internal radiation exposures to personnel, including both individual and total man-rem doses, as low as is reasonably achievable (ALARA).
Shielding has been designed to centrol radiation exposure such that, (1) doses to operating personnel and the general public will be less than those required by the applicable sections of 10 CFR Parts 20; and (2) doses to plant personnel will be ALARA, and the project will follow the guidelines of Regulatory Guide 8.8.
On the basis of our review, the staff has concluded that the radiation protection program will provide reasonable assurance that doses to personnel will be less than the limits established by 10 CFR Part 20, and maintained as low as is reasonably achievable, consistent with the guidelines of Regulatory Guide 8.8.
The radiation protection program is acceptable. Details are discussed in the following sections.
d
. 8.1 Radiation Exposures S&W management is committed to maintaining radiation exposures to its employees as low as is reasonably achievable (ALARA).
For the protection of employees, S&W subscribes to the philo-sophy set forth in the Nuclear Regulatory Commission Regulatory Guide 8.8.
Design guidelines used with regard to keeping occupational radiation exposures ALARA include:
shielding wall thickness established to meet design radiation dose rates; separation of radioactive and non-radioactive equipment to prevent unnecessary exposure during operation or maintenance of non-radioactive equipment; radioactive source increments shielded from each other; ease of maintenance or removal of equipment considered in cubicle design; shielding thicknesses computed along vertical shield surfaces opposite the most intense source.
Based on the information provided in the Report and the responses to our questions, we conclude that S&W intends to desian the Interim Spent Fuel Storage Facilit.y in such a manner that occupational radiation exposures will be ALARA, and consistent with the intent of Regulatory Guide 8.8.
8.2 Radiation Sources SWECO-7601 in Chapter 7 provides concentrations of crud and fission product nuclides in the spent fuel pool water, assuming storage of 12 PWR cores with exposure equivalent to 3 years at 3800 MWe, with correction for decay.
The resulting dose rate is calculated by assuming a uniform distribution in a disk 76 feet in diameter and 10 feet deep.
Dose rates at the pool edge, principally from Co-60 and Cs-137, are calculated using GAMTRAN to be less than 2.5 millirems per hour.
The dose rate contribution from the spent fuel assembly is calculated by QADMOD, in order to assure adequate pool depth to limit surface dose rates to design levels.
. The assu.nptions and procedures used by S&W in estimating source terms have been evaluated.
The resulting estimates are reasonable, and consistent with the acceptance criteria of Section 12.2 " Radiation Sources" of the Standard Review Plan.
8.3 Radiation Protection Design Features Shielding has been designed to protect operating personnel and the general public from radiation sources within the plant, including equipment and piping, while maintaining suitable access for operation and maintenance.
The objective is to limit radiation exposures to the general public, plant personnel, contractors, and visitors to levels that are ALARA and within the limits of 10 CFR Part 20.
SWEC0 has provided 6 radiation zones as a basis for classi-fying occupancy and access restrictions on various areas.
On this basis, maximum design dose rates are established for each zone and used as input for shielding of the respective zones.
For example, maxi'num calculated design radiation levels in Zone II, where operating personnel are expected to be working as much as 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, will be less than 2.5 millirems per hour at any point in the zone.
In actual practice, as has been the case with nuclear power plants, dose rates would be expected to be well below maximum calculated values.
Radiation protection concepts directed to keeping personnel exposures below regulatory limits have been used throughout the design.
Shielding design and radiation zoning were generally based on maximum fuel storage conditions.
To the extent practicable, major sources are in individually labyrinthed, shielded cubicles.
Pipes and ducts are routed through high-zoned, low access areas when practicable; shielding is provided for piping and penetrations.
Shield wall thicknesses are determined using basis shielding data.
The principal calculational method employed Point Kernel intergration, as described in RP-8A, " Radiation Shielding Design and Analysis Approach for Light Water Reactor Plants".
The staff considers the assumptions used in these shielding calculations to be conservative, and the models and codes used acceptable.
The objectives of the ventilation system are to ensure that maximum airborne radioactivity concentrations in the plant are less than the values given in 10 CFR Part 20, Appendix B,
. Table I, Column 1, and are ALARA within plant structures and in controlled areas on the site.
Subsystems shall be arranged for flow from areas with low potential for contamination to areas with a high potential for contamination.
The objective of the area monitoring system is to indicate, record, and alert personnel in the facility monitoring area in the event of abnormal gamma radiation levels in plant areas where higher radiation levels can arise.
Local audible and visual alarms and local readout capability will be provided to warn personnel.
Eight locations are identified for placement of area monitors, as well as the range and sensitivity to be provided.
The airborne radioactivity montiors are placed at the fuel pool area exhaust vent and cubicle area exhaust vent.
They provides samples of the radioactive particulates and gases in those respective areas.
Activity levels are displayed and recorded at the facility monitoring area; high activity level is indicated by audible and visible alarms.
The instrumenta-tion and/or action criteria for these monitors are in confor-mance with the guidelines of Regulatory Guide 8.8, and are acceptable.
8.4 Dose Assessment The estimate of annual man-rem exposure is based on conserva-tively assumed radiation sources, design shielding, calculated average dose rates, manpower levels, and occupancy factors.
Based on expected dose rates and occupancy times, S&W estimates average annual occupational radiation exposure due to all phases of plant operation to be about 31 man-rems.
The bases for this exposure estimate are reasonable, and consistent with the acceptance criteria of Section 12.4 " Dose Assessment" of the Standard Review Plan.
8.5 Health Physics Progr<
The details of the Health Physics Program will be provided by the utility in any application to construct and operate an individual independent spent fuel storage facility.
(See Appendix A, Interface Information, Section 8.)
. Conclusions The staff finds the radiation protection program acceptable and that sections:
8.1 Radiation Exposure 8.2 Radiation Sources 8.3 Radiation Protection Design Features 8.4 Dose Assessment of Chapter 8 Radiation Protection may be referenced in future license applications for independent spent fuel storage installations.
13.0 QUALITY ASSURANCE S&W's QA program governing safety related activities during design, procurement, and construction of the ISFSF is described in Topical Report SWSQAPl-74A, " Stone & Webster Standard Nuclear QA Program."
This QA topical report which commits S&W to comply with the requirements of Appendix B to 10 CFR 50, and to follow guidance prescribed by the staff therein has been found to be an acceptable prcgram by the staff for S&W's scope of work concerning nuclear power plants.
Accordingly the staff finds that SWSQAPl-74A is also acceptable for safety related activities of the ISFSF during design, procurement, and construction and modifications.
In order to provide for continued and reliable operation of the ISFSF, any moderation of the safety related systems, component or structures subsequent to the design phase will be controlled by measures comen-surate with those applicable during original design.
(See Appendix A,
" Interface Information," Section 13, " Quality Assurance.")
Conclusion The staff finds that Section 13.1 " Quality Assurance During Design and ccostruction" of Chapter 13. Quality Assurance is acceptable and may be reis enced in future license applications for independent spent fuel storege installations.
APPENDIX A " INTERFACE INFORMATION" Appendix A includes the infnrmation of Table 1.5-1 of SWECO-7601 -
Table 1.5-1 " Interface Information" has been examined by the staff.
It's outline of interface information required fcr licensing and construction of an ISFSF of the type described in this Topical Report including applicant responsibilities and interface requirements is adequate.
Conclusion The staff finds Table 1.5-1 " Interface Information" acceptable for reference in a license application for an ISFSI.
.