ML20125C319

Staff Exhibit S-2,consisting of Undated Safety Evaluation Re Increasing Spent Fuel Storage Capacity.Proposed Mods to Spent Fuel Pool & Transshipment/Storage of Spent Fuel Acceptable

ML20125C319
Person / Time
Site:	North Anna
Issue date:	05/21/1985
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), Office of Nuclear Reactor Regulation
To:
References
OLA-1-S-002, OLA-1-S-2, NUDOCS 8506120009
Download: ML20125C319 (58)

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SAFETY EVALUATION BRANCH BY THE OFFICES OF NUCLEAR REACTOR REGULATION AND NUCLEAR MATERIAL SAFETY AND SAFEGUARDS RELATED TO INCREASING THE SPENT FUEL STORAGE CAPACITY AND THE STORAGE OF SURRY SPENT FUEL AT i

THE NORTH ANNA POWER STATION, UNITS NO. 1 AND NO. 2 VIRGINIA ELECTRIC AND POWER COMPANY AND OLD DOMINION ELECTRIC COOPERATIVE i

NORTH ANNA POWER STATION, UNITS NO. 1 AND NO. 2 DOCKET NOS. 50-338 and 50-339

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e TABLE OF CONTENTS Page I

1.0 Introduction 1-1 t

2.0 Evaluation 2-1 l

2.1 Design Criteria 2-1 i'

2.1.1 D'escription of the Spent Fuel Pool and Racks 2-1 2.1.2 Applicable Codes, Standards and Specifications 2-2 2.1.3 Loads and Load Combinations 2-3 2.1.4 Seismic and Impact Loads 2-3

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2.1.5 Design and Analysis Procedures 2-4 2.1.6 Materials 2-6 2.2 Criticality 2-8 2.3 Thermal Analysis 2-11 2.4 Accident Analysis 2-14 2.4.1 Postulated Storage Rack Analysis 2-14 I

2.4.2 Rack Handling and Installation 2-16 2.4.3 Cask Drop Accidents 2-18 2.4.4 Spent Fuel Pool Gate Drop Accidents 2-18 1

2.4.5 Fuel Handling Accidents 2-19 2.5 Radioactive Waste Treatment 2-20 2.6 Occupational Radiation Exposure 2-21 2.7 Industrial Security 2-26 4

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TABLE OF CONTENTS (continued)

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i 3.0 Transshipment of Surry Spent Fuel to North Anna 3-1 i

'i 3.1 Radiological Assessment During Normal Conditions 3-1 g

j 3.1.1 R'adiation Doses to Transport Workers 3-1 3.1.2 Radiation Doses to the Public 3-2 3.2 Accident Conditions 3-4 3.2.1 Vehicle Accident - Cask Not Ruptured 3-6

l 3.2.2 Risk to Driver and Escort in Transport 3-6 Vehicle 3.2.3 Risk to Response Personnel 3-6 3.2.4 Risk to Public 3-8 3.2.5 Cask Ruptured 3-8 l

4.0 Nonradiological Assessment for Transshipment of Surry Spent Fuel to North Anna 4-1 4.1 Risk from Vehicular Accidents 4-1 4.2 Risk from Air Pollution 4-3 4.3 Risk from Increase in Traffic Density 4-3 4.4 Risk from Possible Terrorist Act 4-3 5.0 Sumary 5-1 6-1 6.0 Conclusion Appendix A A-1 Appendix B B-1

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SAFETY EVALUATION BY THE OFFICES OF NUCLEAR REACTOR REGULATION AND NUCLEAR MATERIAL SAFETY AND SAFEGUARDS RELATED TO INCREASING THE SPENT FUEL STORAGE CAPACITY AND THE STORAGE OF SURRY SPENT FUEL AT l

THE NORTH ANNA POWER STATIOW; UNITS NO. 1 AND NO. 2 l

VIRGINIA ELECTRIC AND POWER COMPANY AND OLD. DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION, UNITS NO. 1 AND 2 1

DOCKET NOS. 50-338 AND 50-339 1.0 Introduction By letter dated August 20, 1982, the Virginia Electric and Power Company (the licensee) proposed modifications to increase the storage capacity for spent fuel assemblies at the North Anna Power Station, Units No. I and 2 (NAPS, NA-182). The initial licensed capacity of the spent fuel pool was 400 fuel assemblies, but in 1979 the fuel pool was reracked with high density fuel racks which increased the NAPS storage capacity to 966 assemblies. On August 17, 1979, Amendment No. 14 was issued to Facility Operating License NPF-4 for NA-1 allowing replacement of the fuel racks to accommodate 966 fuel assemblies. The NRC Safety Evaluation for increasing the NAPS storage capacity to 966 assemblies was published as a supporting document to Amendment No.14 for the NA-1 Operating License NPF-4., The NAPS 966 fuel assembly storage capacity is the configuration presently in place which allows NA-182 to operate until 1989 with full core discharge capacity.

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The modification proposed by the licensee's August 20, 1982 letter is to replace the presently-installed high density fuel racks with neutron absorber fuel racks which would increase the storage capacity of the spent fuel pool i

from 966 to 1737 fuel assemblies. The installation of neutron absorber spent fuel racks in the NAPS spent fuel pool would provide full core discharge capability for NA-182 until 1997. This storage capacity is based on replacing approximately 33 to 40 percent of the fuel assemblies in the reactor core during each refueling according to the core design parameter and fuel manage-ment scheme'being utilized. The refueling interval for each unit at NAPS is approximately 18 months. This expanded storage capacity would include the capability to accept a full core discharge from one of the reactor units at any time.

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On July 13, 1982, the licensee made application to the NRC for a license amend-ment for NA-182 which would allow the storage of up to 500 spent fuel assemblies from the licensee's Surry Power Station, Units No. I and No. 2 (Surry) in the NAPS spent fuel storage area. As early as the spring of 1986, the licensee will lose the ability for full core discharge capability at Surry 1&2. Both Surry Units 1&2 would have to shut down due to the lack of storage for conducting refueling operations in the fall of 1987 and spring of 1988, respectively.

Storing 500 Surry spent fuel assemblies in the proposed spent fuel pool storage area at NA-1&2 would provide adequate spent fuel storage capacity (full core discharge capability) for both NA-1&2 and Surry-1&2 through 1992.

Inherent in the licensee's amendment request of July 13, 1982, is the necessity for transshipment of 500 Surry spent fuel ass'emblies from Surry to NAPS.

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4 The proposed neutron absorber spent fuel racks to be installed at NA-182 are designed to accomodate either Surry or NAPS fuel assemblies. The spent fLel assemblies for NAPS and Surry are manufactured by the Westinghouse Electric Corporation (W), the nuclear steam system supplier (NSSS) for both power stations. The spent fuel assemblies shipped from Surry to NAPS are 15x15 i

fuel assemblies. The NAPS fuel assemblies are 17x17 fuel assemblies..The spent fuel is contained in long sealed tubes called fuel rods. A cluster of 204 fuel rods arranged in a 15x15 array make up each of the Surry spent fuel assemblies. Similarly, a cluster of 264 fuel rods arranged in a 17x17 array make up each of the NAPS spent fuel assemblies. A comparison of the physical dimensions of the Surry 15x15 and NAPS 17x17 assemblies is provided in Table 1 below.

Table 1

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Comparison of the Physical Dimensions of 15x15 (Surry) and 17x17 (North Anna) Fuel 15x15 17x17 (Surry)

(North Anna) i Overall Length 159.76 159.8

8 Overall Dimensions 8.426 x 8.426 8.426 x 8.426 UO Rods Per Assembly 204 264 2

Guide Tubes Per Assembly 20 24 j

Number of Grids Per Assembly 7

8 Active Fuel Length 144 144 Cladding Material Zircaloy - 4 Zircaloy - 4 Clad Thickness 0.0243(Nominal)0.0225(Nominal) i 1,

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The presently in-place high density racks or the proposed neutron absorber racks to be installed at NAPS are designed to accommodate either Surry or NAPS fuel.

In performing the structural / seismic analysis, the themal hydraulic analysis, and the criticality analysis for the existing and pro-posed racks, the licensee has used either the Surry or NAPS fuel charac-

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teristics that provide the most conservative results. Therefore, the staff's safety evaluation which follows for the licensee's proposed neutron absorber spent fuel racks is applicable to either the storage of Surry or NAPS fuel or a combination of both (maxima of 500 Surry assemblies).

The proposed fuel storage expansion program is limited to the replacement of the current NAPS storage racks with neutron absorber fuel racks. Neutron

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absorber spent fuel racks pemit greater storage capacity by storing the

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spent fuel assemblies in closer proximity to each other. -

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The proposed modifications at NAPS will not alter the external physical geometry of the pool or require structural modifications to the fuel building.

1 However, in response to NRC IE Bulletin 80-11, Masonry Wall Design, the licensee has replaced some masonry walls inside the NAPS fuel building.

These matters are discussed in Section 2 of this report. Seismic and tornado design provisions as stated in the NA-182 Final Safety Analysis Report (FSAR) are not changed as a result of the proposed modification.

In addition, these f!

modifications will not affect the leakage and shielding requirements spect-fied in the FSAR. Also, the spent fuel pool cooling and purification system

~i need not be modified to acconinodate the proposed increased storage capacity.

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44-+F Section 2 of this report addresses the radiological aspects of the licensee's 1

proposed modifications to rerack the NA-182 spent fuel pool with high density neutron absorber fuel racks for an increased storage capacity of 1737 fuel i

i assemblies.

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Section 3 of this report addresses the r-isk. to the health and safety of the-public and to transport workers engaged in the proposed transshipment of j

spent fuel from the Surry reactor site to the NAPS reactor site. This evalua-tion is con'cerned only with the actual transportation and does not consider i

thoseactivities(loading, unloading,etc.)occurringwithintheprotected 1

i areas of the reactor sites. Loading and unloading activities within the restricted reactor sites are addressed in the Surry and NAPS FSARs and the

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staff's Safety Evaluations for Surry and NAPS. Loading and unloading activ-l l

ities associated with the instant proposed modifications at NAPS are addressed i

j in Section 2 of this report. The proposed transshipment evaluation examines the radiological and nonradiological risks that could possibly affect the highway transportation of spent fuel from Surry to NAPS.

i Finally, the dates specified in this report for loss of full core discharge and refueling for NAPS and Surry are those stated in the licensee's letter l

dated May 8, 1984.

It is noted that unscheduled shutdowns of the reactor units and/or unknown conditions could impact and stretch out the dates so specified in this report. However, any slippage in these dates (short tem) j does not mitigate the necessity for the long tem actions requested by the j

licensee.

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I 2.0 Evaluation 4

2.1 Design Criteria 2.1.1 Description of the Spent Fuel Pool and Racks The NAPS spent fuel pool serves both reactor units and is located between the reactor buildings with its long axis running East-West. The pool is a n.:

concrete box, rectangular in plan view. The walls and floor are approximately six feet thick (thicker in some places) and heavily reinforced. A 3.6 foot thick wall separates the spent fuel area and cask areas. The inside dimensions of the pool' are approximately 42.5 feet deep by 57 feet long by 29.25 feet wide, exclusive of the new fuel area and cask area of the East and West ends, respectively. The pool is founded on rock. The tcp of the pool is at elevation (feet above mean-sea-level) 291.83; the bottom of the pool (inside) is at eleva-tion 249.33; grade is at elevation 271.0. The pool is lined with a continuous one-quarter-inch thick stainless steel liner plate which is anchored to the concrete and is designed for the underwater storage of spent fuel assemblies.

The spent fuel pool is designed so that at least 24 feet 1 inch of water is always maintained above the active portions of the spent fuel assemblies stored in the pool. A leak-chase-channel leak detection system is provided.

The existing fuel storage racks are to be replaced with 16 free standing poisoned racks. The capacity of the pool will thereby be increased from.966 spaces to 1737 spaces. The 10 by 12 cell rack is about 10.6 feet long by 8.8 feet wide. All racks are about 14.8 feet high. Racks are primarily constructed of Type 304 stainless steel.

Individual square cells are constructed of 0.090 inch thick stainless steel base plate.

Each base plate is attached to eight adjustable pedestals.

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a The proposed spent fuel storage racks are fabricated of Type 304 stainless steel, which is used for all structural components. The individual cells which make up each rack array are in the fonn of double wall boxes welded to each other with tie plates, so as to maintain a 10 9/16 inch cell pitch. This type of construction provides four compartments which are open to the pool and in which Boraflex neutron absorber elements are placed for criticality control. The Boraflex is positioned on each side of a fuel assembly placed within the cell. The components containing Boraflex are not watertight, thereby eliminating the potential for a pressure butidup within the compartment, for example by radiolysis of entrained water vapor.

2.1.2 Applicable Codes, Standards and Specifications The proposed racks are designed in accordance with the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and revised January 18, 1979 (referred to hereafter as the "NRC Position"). As the basis for structural design of the racks, the NRC Position pennits use of Section III, Division 1, Subsection NF of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME code). Buckling j

of rack components was considered and all buckling design stresses were found to be acceptable when compared with the ASME code as well as the American Iron and Steel Institute Stainless Steel Cold Fonned Structural Design Specification (AISI specification).

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The existing spent fuel pool was originally designed in accordance with the AmericanConcreteInstitute(ACI)BuildingCode,ACI318-63. Acceptance criteria and load combinations as set forth in the NA-1&2 FSAR were used in the analysis of the pool structure.

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l 2.1.3 Loads and Load Combinations

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Loads and load combinations for the design of the racks were found to be in agreement with the NRC Position. Loads and load combinations for the analysis of the pool structure were found to be in accordance with original NA-182 FSAR connitments and are acceptable.

2.1.4 Seismic and Impact Loads The seismic loads are based on the original design acceleration response spectra calculated for the plant at the licensing stage. This was based on a y

0.12 Safe Shutdown Earthquake (SSE) and a 0.06 Operating Basis Earthquake (OBE). Damping values for the racks were taken as 1 percent for OBE and 2 percent for SSE.

Impact effects due to fuel bundle / rack interaction as well as rack / pool floor interaction were included in the analysis. Added mass effects of fluid were applied to the racks in a conservative manner.

i A separate fuel assembly drop accident analysis was perfonned. A 2,500 pound object was postulated to impact the top of the rack from a height of 59 l

inches. The same object was postulated to drop 227 inches through a cell and f

impact the bottom of the rack. The analysis shows that the dropped object t

would deform the top of the racks upon impact.

However, the analysis further showed that the structural integrity of the racks and the spent fuel pool j-floor will be maintained.

t It was concluded that the seismic and impact loads are in accordance with NRC criteria and are acceptable.

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2.1.5 Desian and Analysis Procedures We have reviewed the modeling and analytical procedures used in the seismic analysis of the proposed spent fuel poison storage racks. Also, we have reviewed the methodology used in ana.lyzing the design of the spent fuel structure and liner for the new rack loads. The design and analysis procedures are provided below.

A.

Racks (1) A detailed static analysis model was prepared, from which e'ssential design characteristics were extracted for use in the dynamic analysis models described below.

(ii)Adynamicallyequivalentresponsespectrumanalysismodel wasestablishedbasedonthedatageneratedinstep(1).

Seismic analysis was then performed using response spectrum

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analysis methods.

The corresponding inertia forces at each mass point were statically applied to the detailed model created in step (1) and the stress analysis was then perfonned for various load combinations.

j (iii) A separate, non-linear dynamic analysis model was prepared in order to establish the maximum sliding distance for the rack under seismic excitation. A time history analysis was then performed to establish maximum forces acting on the racks under various conditions of friction.

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(iv) Tipping and subsequent fall back loads were computed using i

energy-balance principles. The maximum energy imparted to the rack is established by the analysis in step (ii). The validity of the results in step (ii) were verified by comparing the maximum base shear and base moment resulting from the two separateapproaches(steps (ii)and(iii)).

It was found, based on the above analyses, that rack component and weld stresses we're within acceptable limits as defined by the more conservative approach of both the ASME code and AISI specification.

B.

pool The spent fuel pool structure and liner were analyzed for the new

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rack loads. The criteria used for the original design of the pool, as presented in the FSAR and used for the original and subsequent analyses, were reviewed by the staff.

Thermal loads were included in the original structural design of the t

pool. Since the temperature of the pool will be maintained at 140 degrees Fahrenheit (*F) for normal conditions and 170 *F for accident conditions as originally designed, no additional thermal analysis of the pool structure was perfomed. This is acceptable to the staff.

It was concluded that the design of the pool is acceptable for the new rack loads when compared to current criteria for Category I structures. Therefore, i

the spent fuel pool is satisfactory for the proposed installation. Based on 2-5 1

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.i the above, we find that the proposed rack installation will satisfy the i

i requirements of 10 CFR 50 Appendix A, General Design Criteria (GDC) 2, 61 l

and 62 (as applicable to structures), and is therefore acceptable.

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2.-l.6 - Materials Rack structural materials are in confomance_with the requirements as specified in the ASME code. The spent fuel pool for NA-182 is fabricated of materials that will have good compatibility with the borated water chemistry of the spen't fuel pool. The corrosion rate of Type 304 stainless steel in j

this water is sufficiently low to defy our ability to measure it. Since all materials in the pools are stainless steel, no galvanic corrosion effects are 4

anticipated. No instances of corrosion of stainless steel in spent fuel l

poolscontainingboricacidhasbeenobservedthroughoutthecountry.(I)

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Boraflex has been shown to be resistant to radiation doses in excess of any anticipated in the NA-182 spent fuel pool.(2) The venting of the cavities I

containing the Boraflex to the spent fuel pool environment will ensure that i

no gaseous buildup will occur in these cavities that might lead to distortion

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of the racks. We have reviewed the licensee's justification for the lack of a materials surveillance program in the spent fuel storage pool. We conclude j

that the surveillance programs presently in place with identical material in other existing spent fuel storage pools will provide adequate infomation on deterioration of materials in these pools. We do not anticipate that such deterioration will occur.

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t From the above evaluation, we conclude that the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the 40 year life of the plant. Components in the spent fuel storage pool are constructed of alloys which have a low differential galvanic J

potential between them and have a high resistance to general corrosion, s

lccalized corrosion, and galvanic corrosion. Tests under irradiation and at elevated temperatures in borated water indicate that the Boraflex material will not undergo significant degradation during the expected service life of 40 years.

We further conclude that the environmental compatibility and stability of the J

materials used in the expanded spent fuel storage pool are adequate based on the test data cited above and actual service experience in operating reactors.

r We find that the monitoring programs in place with identical material in operating spent fuel storage pools and the selection of appropriate materials i

of construction by the licensee meet the requirements of 10 CFR Part 50, Appendix A, GDC 61 (having a capability to permit appropriate periodic in-spection and testing of components) and GDC 62 (preventing criticality by maintaining structural integrity of components and of the baron poison).

Therefore, we find materials compatibility to be adequate and acceptable. We further conclude there will be no significant degradation of materials due to corrosion.

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l 2.2 Criticality The fuel storage racks consist of double-walled stainless steel tubes having a square cross section with an inner diameter of 8.875 inches, an inner wall thickness of 0.090 inch, a neutron absorber chamber of 0.095 inch width and a cover sheet of 0.029 inch thickness. A Boraflex neutron absorber 0.085 inch thick with a boron-10 loading of.025_gr_ams per square centimeter.

2 (gm/cm ) is placed in the absorber chamber. The storage containers are held on a 10.56 inch center-to-center spacing by the racks.

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The nuclear criticality analysis of the racks was performed with the Monte I

Carlo KENO-IV code with 123 group neutron cross sections prepared with the

1. !PX-NITAWL code package. This code is widely used for this purpose in the I

industry and is acceptable.

It has been extensively verified and the

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verification was confirmed for these analyses by comparisons with a series of experiments performed by the Battelle Pacific Northwest Laboratories specially designed to mock up storage rack configurations, including poison curtains between assemblies. The results show that the KENO-IV code conservatively overpredicts the effective multiplication factor (K effective)* by less than one percent.

No credit was taken for this overprediction.

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• K effective is the ratio of neutrons from fissions in each generation to the total number lost by absorption and leakage in the preceding i

generations. To achieve criticality in a finite system, K effective must equal 1.0.

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The analysis was performed under the assumption of fresh fuel of 4.3 weight i

percent U-235 enrichment containing no burnable poison in unborated water.

No credit is taken for neutron leakage from the racks and for structural material other than that in the fuel storage containers. Both normal and abnormal configurations were considered. Normal configurations included the p

reference configuration and those variations in rack dimensions, fuel parameters and fuel location pennitted by fabrication tolerances.

Fuel and j

cell parameters used in the analysis are shown in Table 2.2.

Abnormal configurations included the results of credible accidents, seismic events, i

i and malfunctions of the fuel pool cooling system.

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The results of the analyses showed that the effective multiplication factor for the nominal configuration is 0.935. Whenalluncertainties(95/95)are statistically combined and added, the result is 0.947. This meets our acceptance criterion (0.95) for this quantity and is acceptable. The effective multiplication factor decreases with increasing water temperature.

The results of seismic events are bounded by the eccentric location analysis and are included in the uncertainty band. Placement of assemblies in other than the designated locations is prevented by the structural design. A fuel assembly which lies across the top of the racks is isolated from the fuel in the racks by more than 12 inches of water and is thus neutronically decoupled.

We conclude that any number of fuel assemblies of design similar to the Westinghouse 15x15 or 17x17 design and having an initial enrichment of no greater than 4.3 weight percent U-235 may be safely stored in the proposed North Anna spent fuel racks.

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Table 2.2 Fuel and Cell Parameters for Reference Configuration Westinghouse Fuel Type 17x17 15x15 Fuel Enrichment, w/o 4.3 4.3 Fuel Rod 00, inches

_ ___0.374 0.422 Fuel Rod ID, inches 0.329 0.3734 Fuel Rod Pitch, inches 0.496 0.563 Number of F'uel Rods 264 204 Cell Pitch, inches 10 9/16 10 9/16 Cell ID, inches 8 7/8 8 7/8 Cell Wall Thickness, inches 0.090 0.090 Neutron Absorber Material Boraflex Boraflex 10 2

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Neutron Absorber B Loading, gms/cc 0.025 0.025 Neutron Absorber Thickness, inches 0.085 0.085 Neutron Absorber Width, inches 7.5 7.5 Neutron Absorber Length, inches 138 138 Neutron Absorber Chamber Width, inches 0.095 0.095 Cover Sheet Thickness, inches 0.029 0.029 2-10

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2.3 Thermal Analyi_s_

i We have previously found the design for the spent fuel pool cooling and refueling purification system to be acceptable',' as discussed in Section 9.1.3 of the NA-1&2 Safety Evaluation Report.

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The sper:t fuel pool cooling portion of the fuel pit cooling and refueling purification system removes residual heat from spent fuel stored in the shared NA-182 spent fuel pool. The spent fuel pool cooling system is composed of' redundant trains, each train containing a pump and heat exchanger. The redundant trains can be cross-connected so that either pump can provide flow through either or both heat exchangers. The heat exchangers l

are cooled by component cooling water, with service water available as an j

emergency supply of cooling water. The spent fuel pool cooling system heat

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exchangers can be cooled.by the component cooling water system associated with either or both reactor units.

The design of tha' storage pool is such that the fuel will always be covered with water. Because of the locations of fuel pool piping penetrations, the configuration of the pool and the use of siphon breaker vents, no incorrect operation or failure in the' fuel pit cooling and refueling purification 1

system could crain the fuel pool water level below elevation 285 feet 9 inches. At this elevation, there is still 24 feet 1 inch of water above the

' fuel. Makeup water is nonnally supplied to the spent fuel pool from the boric acid blender in the chemical and v'olume control system associated with either reactor unit. Assured make up can be supplied from the seismic Category I service water system or the seismic Category I fire protection water system.

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The licensee has provided an analysis of the maximum nonnal and abnormal spent fuel decay heat loads in the spent fuel pool resulting from the proposed increased spent fuel storage capacity. The nonnal decay heat load results from the maximum number of normal annual refuelings where approximately one third of the fuel asemblies in the core are removed to the spent fuel pool. A full core contains 15Lfuel assemblies. All. storage spaces in the pool are assumed to be full except for the spaces reserved for I

a full core offload. For the abnonnal decay heat load case, all spent fuel storage spa'ces are assumed to be full, including a full core offload. For the nonnal refuelings, the spent fuel is assumed to be removed from the core to the pool instantaneously at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown.

For the emergency full core offload, the fuel assemblies are assumed to be removed from the core at the rate of 20 minutes per assembly, beginning at 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after

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6 shutdown. The licensee calculated a nonnal decay heat load of 23.1x10 i

6 British Thermal Units per Hour (Stu/hr) and an abnormal heat load of 39.2x10 Btu /hr. These calculated heat loads result in spent fuel pool temperatures of 138.5 "F and 156.5'F, respectively, with one spent fuel pool cooling pump and two heat exchangers in operation. The pool temperature is also calculated to be 135"F (nonnal heat load) and 151*F (abnonnal heat load) if the spent fuel j

pool cooling system is assumed to be completely operational with both spent fuel pool cooling pumps operating.

l The licensee has also provided decay heat load analyses for nonnal and I

full-core offload refueling with 500 spaces in the fuel racks occupied by spent fuel transferred from Surry. The presence of Surry fuel increases the decay heat load by a small incremental amount. With 500 Surry fuel 2.

i assemblies in the NA-1&2 spent fuel pool, the licensee calculates a normal 6

decay heat load of 24.08x10 Btu /hr and an abnormal heat load of 6

40.08x10 Btu /hr. These calculated heat loads result in a spent fuel pool temperature of 140*F (normal heat load) and 160'F (abnormal heat load) with one spent fuel pool cooling pump and two heat exchangers in operation. With the spent fuel pool cooling system fully operational, pool temperatures would be reduced to 136*F (normal heat load) and 148.5'F (abnormal heat load).

i The license'e has stated that the decay heat load was calculated in accordance with the guidelines of Branch Technical Position (BTP) ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling". An independent calculation by the staff has verified that, in general, the licensee's methodology for calculating spent fuel decay heat is acceptable. However,

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for the abnormal heat load case, the licensee assumed that the full core offload would take place at the nonnal refueling time. For the more conservative case with full-core offload occurring 30 days after startup from 6

the previous refueling, the staff calculates a decay heat load of 44x10 Btu /hr. This staff value is approximately 10 percent higher than the licensee's calculated value for the abnonnal heat load. However, by letter dated April 10, 1984, the licensee verified that the higher heat load would result in an acceptable pool temperature of 152.6'F with the spent fuel pool cooling system fully operational. Thus the spent fuel pool cooling system is capable of maintaining acceptable pool temperatures with the proposed increased spent fuel storage capability.

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i Based on the above, we conclude that the spent fuel pool cooling system meets the requirements of GDC 44 and the guidelines of BTP ASB 9-2 with regard to decay heat removal capability for the proposed increased spent fuel storage capacity.

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2.4 Accident Analysis 2.4.1 Postulated Storage Rack Analysis The proposed spent fuel storage modifications will provide storage locations for 1737 fu'el assemblies and two failed fuel canisters. Each fuel assembly will be stored in a double-walled storage cell of Type 304 stainless steel.

The annular spaces between the double walls of the cell contain B C (boraflex) 4 neutron absorber elements positioned at the rack height corresponding to the active fuel length of the fuel assemblies. The individual storage cells are welded into rack arrays with array sizes ranging from 9x9 to 11x11 fuel assemblies (9x9, 9x12, 10x11, 10x12, 11x11). This configuration maintains a cell pitch of 10 9/16 inches and prevents placement of a fuel assembly in a location other than a storage cell. As stated previously, the licensee has verified that K,ff of the storage fuel configuration is maintained below 0.95 for normal and anticipated abnormal conditions assuming unborated water in the pool and no burnable poison in any of the storage fuel assemblies. Structural and seismic analyses have been performed by the licensee to verify that the l

rack design is adequate to withstand norrel operating, seismic and accident load conditions.

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Postulated accidents censidered were:

(1) the possibility of the fuel handling bridge fuel hoist grapple becoming hceked en a fuel storage rack, and

'h (2) The accidental drcp of a spent fuel asseebly from C

l the highest pcssible elevatten during spent fuel

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handling ente the storage rack.

l For item 1 above, an axial upward force of 4,000 pounds was considered to be

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exerted on the rack. By letter dated Septester 13, 1983, the licensee verified that a lead limiting cevice is used to automatically step upward heist retien if a preset weight is exceeded. This lead limiter setpoint is remally set at 3,400 pcunds which conservatively limits any postulated upward force of 4,0C0 pcunds en the storage racks.

For ites 2 above (the accidental drop case) a 2,500 rcund weight, the maximun load permitted by the MA-1&2 Technical Specifications (TS) cver spent fuel, j

was postulated to drop on the rack from a height of Sg inches abcve the top of the rack. This postulated 5g inches in drop height is greater than the height at which fuel assemblies are normally raised during fuel handling cperations.

The results of the seismic and structural analysis indicate that the stresses in the rack structure resulting from the specified load cases are within

~

allowable stress limits for seismic Category I structures. The analysis of the accidental fuel assembly drop condition indicates no buckling or collapse of the storage cells or puncturing of the fuel pcol liner (leak tightness j

integrity of the pool).

In all cases analyzed, the value of K,ff did not exceed 0.95.

2-15 i

t 4

The spent fuel pool and the proposed neutron abscrber storage racks are designed to seismic Category I criteria. However, in response to IE Bulletin 80-11, Masonry Wall Design, the licensee advised the staff that 15 of the masonry block walls in the NAPS fuel building do not meet structural design require-ments and would be replaced.

By letter dated April 10, 1984, the licensee 3

verified that the supports for the replace!nent walls are designed to seismic Category I requirements. The siding panel assemblies attached to the Category I -

supports consist of sheet metal, rigid styrofoam and subgirts screwed together as a unit. These assemblies could be dislodged by a design basis earthquaks or tornado. However, the licensee verified that the largest panel assembly j

weighing approximately 650 pounds could not fall into the spent fuel pool with sufficient kinetic energy to result in impact on the spent fuel racks.

i

~

Based on fj abm, we conclude that the spent fuel storage facility and the proposed neutron absorber fuel racks meet the requirements of GDC 2 and GDC 62 with respect to seismic design considerations and prevention of criticality, and 1

the guidelines of Regulatory Guide 1.13, " Spent Fuel Storage Facility Design Basis," Rev.1, December 1975 and Regulatory Guide 1.29, " Seismic Design Classification," Rev. 3, September 1978 with respect to fuel storage design i

and design classification and are, therefore, acceptable.

S 2.4.2 Rack Handling and Installation The review of heavy load handling at North Anna is being conducted as part of 1

the ongoing generic review initiated by NUREG-0612. " Control of Heavy Loads at Nuclear Power Plants". The results of that review will be reported as part of the Multiplant Action Item C-10. The' evaluation provided herein is limited j

to the heavy load handling activities associated with the proposed spent fuel j

storage modification.

2-16

The replacement neutron absorber spent fuel racks will be offloaded from a truck and brought into the fuel building using the fuel building crane. Movement of the racks inside the fuel building will be done with a special temporary crane which will be inspected and load tested prior to use in accordance with American Nuclear Society Institute Standard (ANSI) 830.2-1976.

In addition, p

the design of the crane will be in accordance with the appropriate guidelines of NUREG-0612.

A special lifting rig, provided by the rack manufacturer, will be used for the positioning and installation of the new racks. The licensee states that the special lifting rig satisfies Guideline 4 of NUREG-0612, Section 5.1.1(4),

and will -incorporate remotely actuated positive capture devices which will preclude accidentally dropping a rack during handling.

In addition, the licensee perfonned an analysis to verify that the drop of a rack from the

~

highest lift elevation could not result in unacceptable fuel pool damage or i

loss of spent fuel cooling function. Based on our review, we have verified 3

that the methodology and conclusions of the licensee's analysis are acceptable.

By letter dated September 13, 1983, the licensee provided a description of the rack handling procedures and verified that racks will not be transported above spent fuel. Also, the licensee verified that the procedures used in the rack handling would meet the requirements of Guideline 2 of NUREG-0612, Section i

5.1.1(2). Crane operators will be trained in accordance with Guideline 3 of NUREG-0612, Section 5.1.1(3). As discussed in a licensee letter dated April 10, 1984, the operations involved in the assembly and disassembly of the temporary crane will be performed away from the stored spent fuel such that postulated accidents during these operations will not affect spent fuel assemblies.

2-17

Based on the above, we conclude that the use of the proposed cranes and load lifting devices for rack handling and installation meets the requirements of GDC 4 and GDC 61 with respect to protection of systems or components important to safety from load drops, and the guidelines of NUREG-0612. Section 5.1.1, with respect to safe load handling practices.

_.. ~.

2.4.3 Cask Drop Accidents The licensee states in a March 22, 1982 submittal addressing NUREG-0612 require-ments that 'the fuel building trolley (1-MH-CR-15) used for moving spent fuel casks does not move over stored fuel. The hook centerline is capable of movement to 5 feet 4 inches from the edge of the fuel pool. The west wall of the fuel pool separates the spent fuel cask storage area from the fuel pool. Only during the movement of spent fuel casks into and out of the fuel building are the casks

" ~ '

raised above the top of the fuel pool wall. The centerline of the casks during this movement can be no closer than 1 foot 10 inches from the outside edge of the pool wall. The trolley is equipped with eddy current brakes, dual load holding brakes and " dead man" controls.

In addition, the lift height is limited to one foot. These characteristics greatly reduce the likelihood of occurrence of a cask drop, obviating the need for consideration of the radiological conse-quences of an accident in which a dropped cask would impact stored fuel. There-fore, we conclude that an analysis of the radiological consequences of a cask drop accident is not required.

2.4.4 Spent Fuel Pool Gate Drop Accidents The licensee states in its March 22, 1982 submittal addressing NUREG-0612 issues that the Fuel Building Movable Platform with Hoists (1-MH-FH-13), which is used to move the fuel cavity gates, is designed to be mineuvered over the spent fuel 2-18

_y.,

i 4

pool, the fuel transfer canals, and the new fuel handling and storage area as required during fuel handling operations. The movement of the platform is not restricted by electrical interlocks or mechanical stops. Technical Specifications prohibit the movement of loads in excess of 2,500 pounds from travel over gi irradiated fuel assemblies in the spent fuel pit and the licensee has proposed that plant procedures be revised to prohibit the handling of loads in excess of 2,000 pounds over spent fuel.

In addition, administrative procedures will require that the top of the fuel cavity gates be secured to the top of the fuel pool wall by chains during movement of the gates to ensure that the gates, if dropped, will be prevented from tumbling into the fuel pool and damaging the 1

spent fuel racks. Therefore, movement of the spent fuel pool gate should not result in an accident that could result in offsite radiological consequences.

2.4.5 Fuel Handling Accidents The maximum loads which may be transported over spent fuel in the pool is limited to that of a single assembly. We have previously analyzed the radio-logical consequences of a fuel handling accident at NA-182. The pror;;ed consequences of a fuel handling accident are found in the NA-1&2 staff Safety Evaluation Report of June 1976, wherein the NRC staff stated that the calculated doses are well within the guidelines of 10 CFR Part 100. The proposed modifications for increasing the present storage capacity at NA-1&2 to 1737 fuel assemblies does not increase radiological consequences of fuel

~

handling accidents previously considered in the NA-1&2 Safety Evaluation Report. Therefore, the NRC staff concludes that the radiological consequences of accidents involving fuel handling accidents related to the expansion of l

storage capacity in the NA-1&2 spent fuel pool meets the acceptance criteria l-of Standard Review Plan Section 15.7.4 and the guidelines of 10 CFR Part 100, and therefore, are acceptable.

2-19

2.5 Radioactive Waste Treatment NA-1&2 contain waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes that contain radioactive materials. The waste treatment systems were evaluated in the NRC Safety Evaluation Report dated June 1976, and Supplement No. 2 to the Safety Evaluation Report dated August 1976. There will be no change in-the-waste treatment systems or in the conclusions given in Sections 9.0 and 11.0 of the evaluation of these systems as a result of the proposed modification. The staffs evaluation of the spent fuel pool c'leanup system, in light of the proposed modification, has concluded that any resultant additional burden on the system is minimal because most of the activity will decay away and stop leaking from the spent fuel between refuelings. As a result, the added fuel would contribute little or no additional radioactivity. The existing spent fuel pool cleanup system is adequate for the proposed modification and will maintain concentrations of radioactivity in the pool water to "as low as is reasonably achievable" in accordance with Appendix I to 10 CFR Part 50, and, therefore, is acceptable.

Our evaluation of the radiological considerations supports the conclusion that the proposed modifications to the ipent fuel pool etGNA-1&2 are acceptable because:

(1) The conclusions of the evaluation of the waste treatment systems, as found in the NA-182 Safety Evaluation Report (June 1976) and Supplement No. 2 (August 1976), are unchanged by the proposed modification to the spent fuel pool.

(2) The existing spent fuel pool cleanup system is adequate for the proposed modification.

i 2-20

..m..

2.6 Occupational Radiation Exposure The staff has reviewed the radiation protection related portions of the licensee's plan for the removal and disposal of the presently installed high density fuel racks and the installation of neutron absorber fuel racks. The licensee estimates that the occupational exposure for this operation will be p

approximately 14.0 person-rems *, This estimate is based on the licensee's breakdown of occupational exposure for each of the following phases of the modification: (1) rack removal / installation, (2) diver operations, (3) fuel shuffle, and (4) disposal of racks. The licensee considered the number of individuals perfonning a specific job, the average dose rate in the area where the job is being performed, and the worker occupancy time while performing each job.

The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel (26 J

feet). The minimum water depth over a fuel assembly while it is being transferred is approximately 9 feet, which results in a dose rate at the water surface of less than 50 mrem /hr.

• Person-rem is an expression for the sunination of whole-body doses to persons in a group. Thus, if each member of a population group of 1,000 people were to receive a dose of 0.001 rem or if two persons were to receive a dose of O 5 rem each, the total person-rem dose in each case would be 1 person-rem.

2-21 i

One potential source of radioactivity in the spent fuel pool water is radio-active activation of corrosion products (referred to as crud). There are two principal sources of crud in the spent fuel pool water. Crud in the reactor coolant system water mixes with the refueling water and enters the spent fuel pool during refuelings. Also, crud deposits on spent fuel assemblies can be shaken loose in the spent fuel pool during fuel handling. Crud levels _in the j

spent fuel pool are highest during refuelings and decrease continuously over i

the plant cycle between refuelings. Another source of radioactivity in the spent fuel ' pool water is fission products.

Fission products are released through defects in the spent fuel cladding. Once the fuel is removed from the' reactor vessel and is no longer being irradiated, the release rate of fission products from the fuel is greatly reduced. The licensee will use the installed spent fuel pool purification system at NAPS to remove the nonvolatile corrosion and fission product nuclides from the spent fuel pool water. The removal of these nuclides will help to maintain radiation exposure to personnel at low levels. The licensee has estimated that the increased fuel storage at NAPS will have essentially no impact on the concentrations of airborne radio-activity in the fuel building.

1 4

The licensee is currently considering two alternative methodologies for disposal of the present spent fuel racks. These are: (1) decontamination of the intact fuel racks; and (2) cutting up of the spent fuel racks for offsite disposal. The licensee will evaluate each alternative method using the following criteria before deciding on which disposal methodology to use:

1.

As low as reasonably achievable (ALARA) personnel exposure 1

2.

Minimization of waste volume f

3.

Cost effectiveness 2 '

Based on previous operations for the disposal of spent fuel racks at Surry and NAPS, the licensee estimates that approximately 1.2 person-rems will result from the disposal of the present spent fuel racks. ' Rack disposal I

dose estimates from other utilities have ranged from approximately 0.7 to i

5.2 person-rems, depending on the disposal method selected. Dose estimates T

within this rance for other utilities have been previously reviewed and 1

accepted by the staff. The licensee's rack disposal estimate of 1.2 person-rems is within the envelope of this previously accepted range, and is,

]

l therefore a'cceptable. Once the licensee's plans are finalized and available, the licensee is requested to submit for NRC approval the final method for present rack disposal.

i i

In preparation for fuel rack disposal, the licensee will hydrolyze each fuel rack

~

to remove as much contamination as possible prior to removal from the fuel pool.

f' After an element has been lifted out of the water, it will be washed down with i

-j demineralized water to remove any remaining loose contamination. When the i

,i licensee used similar decontamination techniques at Surry during the 1978 spent j

fuel pool modification, the resulting fuel rack exposure levels were generally less than 30 mrem with very localized spots having levels greater than 100 mrem i

The licensee will use divers to precisely place interface plates, where required, on the spent fuel pool floor and to visually confirm the proper placement of l

~

the fuel racks on the fuel pool floor and the interface plates. Following l

removal of the old fuel racks, the licensee will vacuum the floor of the spent I

fuel pool to minimize the amount of contamination which would possibly be f'

stirred up by the divers and by the rack installation process. The licensee I

l 2-23 I

,j will position the spent fuel racks currently in the pool for a configuration that will minimize doses to the divers. Health Physics personnel will perform radiation surveys of the areas where the divers will work and will be in constant voice comunication with the divers -in the pool. The licensee has estimated that the total dose to the diver (s) used for the reracking operation at NAPS

,j will be approximately 4 person-rems.

1 The staff has reviewed the licensee's spent fuel pool modification report, including a' description of the health physics practices which the licensee will implement for the spent fuel pool modification. Many of these practices are similar to those used by the licensee during the similar fuel rack-replacement performed at Surry in 1978. Based on our review of the licensee's report, the staff concludes that the NAPS spent fuel pool modification can be performed in

[

a manner that will ensure that exposures to workers will be as low as is reasonably achievable (ALARA).

i The spent fuel pool modification at NAPS will increase the spent fuel pool j

capacity from 966 fuel assemblies to 1737 fuel assemblies. The spent fuel assemblies (17x17 or 15x15) will in themselves contribute a negligible amount

' l i

to dose rates in the pool area because of the depth of the water shielding the fuel. The escape of gaseous or volatile fission products from the added spent i

fuel is expected to be negligible, therefore there will be essentially no impact on concentrations of radioactivity in the fuel building atmosphere. As more fuel is stored in the pool, there is a possibility that increased amounts of radioactive crud will be introduced into the pool.

In order to remove these additional impurities, the filters and demineralizer resins will have to be i

j changed on a more frequent basis. The licensee estimates that one additional l

2.

-1

-... = -. - _ _

1 filter and resin change per year will be required resulting in an additional 0.346 person-rem / year.

In 1981, the total personnel exposure from all fuel pool related activities at NAPS was 2.4 person-rems. Based on these figures, the total annual personnel exposure from all fuel pool related activities following reracking should be less than one percent of the total annual p

occupational radiation exposure at both units. The small increase in radiation exposure due to reracking should not affect the licensee's ability to maintain individual occupational doses at as low as is reasonably achievable levels and within 'the limits of 10 CFR Part 20. Thus, we conclude that storing additional fuel in the NAPS spent fuel pool will not result in any significant increase in doses received by plant personnel.

There will be personnel exposure associated with the loading and unloading of spent fuel shipping casks used in the transport of Surry spent fuel. The licensee estimates that there will be approximately 150 to 500 cask shipments over an estimated five to six year period. These shipments will result in a total personnel exposure from cask loading / unloading of approximately 28 to 84 person-rems (between 5.6 and 16.8 person-rems per year, based on a five year shipment period). The maximum estimated increase in annual worker dose from cask loading / unloading of 16.8 person-rem (based on 500 shipments over a 5 year period) represents a small fraction (approximately 2.0 percent) of the average annual radiation exposure at NAPS of 816 person-rems *. This

,1 i

~I

• The average dose is the average total dose for both units, and is taken from

" Occupational Radiation Exposure at Comercial Nuclear Power Reactors 1982,"

NUREG-0713, Vol. 4, December,1983.

2-25

1 small increase in radiation exposure due to cask loading / unloading should

~

not affect the licensee's ability to maintain individual occupational doses i

l at as low as is reasonably achievable levels and within the limits of 10 CFR j

Part 20. Thus, we conclude that loading / unloading activities associated with spent. fuel shipping casks will not result in any significant increase in dose

,l received by plant personnel.

.t l

2.7 Industrial Security We have rev'iewed the proposed modifications to increase the storage capacity for 4

spent fuel assemblies with respect to industrial sabotage. We consider the proposed modifications in the spent fuel pool to have no effect or relevance to the security plan presently in effect for NA-1&2. Our conclusion is based on l

the fact that the spent fuel pool is designated as a vital area. As a vital

~

area, it is afforded the protection required by 10 CFR Section 73.55 to provide high assurance against successful industrial sabotage by both of the following:

(1) A detennined violent external assault, attack by stealth, or deceptive actions, of several persons with the following attri-butes, assistance and equipment: (i) well-trained (including military training and skills) and dedicated individuals, (ii) inside assistance which may include a knowledgeable individual 4

who attempts to participate in both a passive role (e.g., provide i

lt information) and an active role (e.g., facilitate entrance and exit, disable alarms and communications,' participate in violent i

i attack), (iii) suitable weapons, up to and including hand-held 2 i 1

r _. -.,, _.,,,,.

__..,m.,

,____,y

.-.,.,-._.-_,..g,,

automatic weapons, equipped with silencers and having effective long range accuracy, (iv) hand-carried equiprnent, including incapacitating agents and explosives for use as tools of entry or otherwise destroying the reactor integrity, and (2) An internal threat of an insider, including an employee (in any position).

In light of the above, the proposed modifications for reracking with high density fuel racks to increase the spent fuel storage capacity does not change the required level of protection nor the structural design of the external barriers of the pool against the threat of industrial sabotage.

1 e

4 a

I 4

2-27

3.0 Transshipment of Surry Scent Fuel to North Anna 3.1 Radiological Assessment During Nonnal Conditions This section deals with expected radiation exposures to transport workers and to the public during normal transport conditions and possible radiation risks resulting from transportation accidents.

i 3.1.1 Radiation Doses To Transport Workers _..

In its route plan, the licensee has proposed a primary route and four alternate routes designated A, B, C.,and D as provided in its July 13, 1982 request for spent fuel ' shipping route approval. Safeguards route surveys have been conducted by personnel from the NRC Material Transfer Safeguards Licensing Branch on all five routes and, based on the findings of these surveys, the routes have received safeguards approval (see letter from T. S. Sherr, NRC to R.H.Leasburg(licensee)datedJuly 28,1982). The licensee commits to

'~

having two persons in the transport vehicle through heavily populated areas plus one person each in leading and trailing armed escort vehicles. The radiation dose received by transport workers is estimated as follows.

l The dose rate in the cab of the transport vehicle was assumed to be 0.2 mrem /hr (USAEC Report WASH-1238, " Environmental Survey of Transportation of Radioactive Materials To and From Nuclear Power Plants," December 1972,

p. 40, Ref. 3). For each route, the amount of each type of road was deter-mined. Vehicle speeds were assumed to be 45 miles per hour (mph) on 6-and 4-lane highways, 35 mph on 2-lane (rural) highways, and 15 mph on city streets.

Escort vehicles were assumed to lead and follow, respectively, at a two-second interval from the transport vehicle. The distance between the escort vehicle and the transport vehicle would then be 130 feet, 100 feet, and 45 feet for I

these three types of highway. The dose rates at these distances were calculated 3.

.ve-+

-,--gw,,-.

from the equation in Reference 3, p.107. A smaller dose rate factor, K, was used to account for the longer cooling of the Surry fuel as compared with that assumed in Reference 3.

For these calculations, a dose rate factor of 390 was used. This value was obtained by multiplying the dose rate factor used in k' ASH-1238, Reference 3, by the ratios of total dose rates at 10 feet from the cask surface for fuel cooled 730 days and 150 days (See letter dated November 7,

.f 1983, from C. V. Parks, Union Carbide Corporation, Nuclear Division, to R. H.

Odegaarden, NRC; Reference 9). For compliance with 10 CFR 71.47(c), the radia-tion level ~of 2 meters (approximately 6 feet) must not exceed 10 mrem per hour.

To achieve this level, the dose factor must not be greater than 360. The Certi-ficate of Compliance for the TN-8L cask would further restrict the radiation level to 17 mrem per hour at 3 feet. This level corresponds to a dose rate factor of approximately 150. Thus, the dose rate factor used in these calculations is

'~

conservative with respect to the requirements specified in 10 CFR 71.47(c).

The total exposure to the two persons in the cab for all 167 shipments would be between 200 and 300 person-mrem, depending on the route. Total exposure to the escorts would be about 40 person-mrem, and the cumulative exposure for all trans-port workers would be of the order of 300 person-mrem, regardless of the route taken. '

The dose to each person in the cab is but a small fraction of the 500 mrem /yr dose restriction for persons in unrestricted areas specified by the National Committee on Radiation Protection and Measurements (NCRP), NCRP-39, 1971 i

and the International Committee on Radiological Protection (ICRP), No. 2,1959 and No. 26, 1977.

3-2 e

.. - - - - - -r

o l

3.1.2 Radiation Dose to the Public To calculate the dose received by members of the public, each route was divided into segments for each type of highway and each census unit. Population densities were calculated from 1980 census date and county areas provided in the U. S. Depart-ment of Connerce " County and City Data Book",1977. Based on data as stated in USNRC Report NUREG-0170, " Final Envir.onmental Statement on the Transporta-tion of Radioactive Material by Air and Other Modes", December 1977, p. 4-16, Ref. 4, the traffic densities (vehicles per hour) were estimated to be:

3 Highway Type Traffic Density

~

6-lane Interstate 2800 4-lane Interstate 1480 4-lane Primary 780 2-lane Rural 470 Highway speeds were again assumed to be 45 mph on interstate and 4-lane primary roads, 35 mph on 2-lane rural roads, and 15 mph on city streets.

On the average, two persons were assumed to occupy each vehicle on the highway at the same time 1

as a fuel shipment.. For all 167 shipments, the cumulative dose to persons on the highway traveling in the same direction as the shipment was calculated using Eq. D-22 (p. D-13 of Reference 4) to be about 200 person-mrem.

From Eq. D-17, (p. D-11 of Reference 4), the cumulative dose for all shipments to persons on the highway traveling in the opposite direction as the shipment was calculated to be a little less than 200 person-mrem. The cumulative dose to all persons living along the shipment route, based on Eq. D-8, (p. D-6, Reference 4), was calculated to be about 40 person-mrem.

3-3

The total cumulative exposure to the public for all shipments is the sum of exposures for those persons on the route traveling in the same direction, those traveling in the opposite direction, and those along the route not on the highway. The cumulative exposcre would be on the order of 400 person-mrem.

This dose would be distributed among many thousands of persons living along 157. '

l the route and traveling on the highways with the shipments. This cumulative exposure is negligible in comparison with the cumulative exposure that this population receives from natural background radiation.

Consider now the maximum exposed individual. Assume a person who lives along the highway and who is within 100 feet of the transport vehicle as it passes on each of its 167 trips. From Reference 3, p. Ill, an individual 100 feet from a shipment traveling 8.33 mph would receive an exposure of 0.00058 mrem.

~

Assuming that the individual postulated in this analysis is in a town where the vehicle speed is estimated to be 15 mph, he would be exposed each trip to i

(0.00058) x (8.33/15) = 0.0003 mrem per shipment, or a total of 0.05 mrem for all the shipments necessary to ship 500 Surry spent fuel assemblies. This I

exposure is negligible in comparison with the more than 100 mrem / year of natural background radiation. A person along other portions of the route would receive a smaller dose because the vehicle would pass more quickly than in the example provided above.

3.2.1 Accident Conditions From Reference 4, accident frequency is about one accident in one million kilometers, or 1.6 accidents per million miles.

For a round trip using the 3-4

.o longest route (177 miles one way), three vehicles would accumulate 1062 vehicle-miles.

(Note that only one vehicle, equivalent to 354 miles, would s

be transporting the spent fuel cask.)

In the 167 round trips to haul 500 assemblies, the total distance would-be 59,000 vehicle-miles for the transport vehicle and 177,000 vehicle-miles for all three vehicles combined.

j The expectation is a little more than one.Jn_.four (0.25) that one of the vehicles would be involved in a collision sometime during the proposed trans-I shipment and about one in ten (0.1) that the transport vehicle carrying spent fuel would 'be involved in an accident.

If an accident should occur, it would likely not be serious. An accident severity classification scheme and the relative frequencies of truck accidents for each category are given In Reference 4, indicating that 91 percent of all

~

accidents are in Accident Severity Categories I and II. These may be charac-terized as " fender benders", and would not be expected to involve the cask nor

,f hinder completion of the shipment. Accidents in Severity Categories III and i

i IV occur 8.6 percent of the time. These accidents would likely result in 1

considerable vehicular damage and possible personnel injury, but would likely not affect the cask or its tie-downs. Categories V, VI, and VII occur in 0.4 t

percent of all accidents.

If an accident of this severity were to occur, one would expect some damage (but not rupture) of the spent fuel cask, major vehicular damage, and serious injury or death of person (s) involved. An accident of Severity Category VIII is expected to occur in only 0.0015 percent 2

of all accidents. The probability of such an accident to the vehicle trans-porting the spent fuel is considerably less than one in a million (0.000001).

A Category VIII accident could conceivably result in a cask rupture.

3-5 j

o 3.2.2 Vehicle Accident - Cask Not Ruptured For an accident in which no damage is done to the cask or its tie-downs, the radiological effects would be virtually negligible. An accident involving vehicle overturn, fire, or the cask coming loose from its restraints could present a radiological risk during response and recovery operations, and these risks will be dealt with below.

I1 1

3.2.3 Risk to Driver and Escort In Transport Vehicle The most serious radiological risk would occur if the driver and/or escort i

in the transport vehicle were pinned in the transport cab in proximity to a loaded spent fuel cask.

It is expected that the maximum dose rate at one meter from the cask would not exceed 1 rem /hr. Assume also that the time to extricate the person (s) from the wreckage could reach 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The individual dose received under these circumstances would be 2 rem. This exposure would be far from life-threatening. Furthermore, the physical hann from radiation would likely be much less serious than the bodily injury suffered in the

. i

]

accident.

3.2.4 Risk to Response Personnel In the accident just postulated, the greatest radiation risk to response

.f personnel would accrue to persons involved in the extrication operations.

!l Although no one person would likely be in the 1 rem /hr radiation field ij for the entire 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> postulated for the extrication, one may consider this exposure as a bounding case. Under these circumstances, each of the rescue persons so engaged would receive a dose of 2 rem, the same as for the driver or escort.

.i 3-6 i j 4

For persons carrying out fire suppression and emergency medical treatment at the scene, one may assume that any one response person spends from 10 to 20 minutes within 10 feet of an accessible surface of-the cask. The dose rate i'

at 10 feet from an accessible surface of a loaded cask under the poetulated accident conditions would be of the order of 100 mrem /hr. Therefore, the response person might receive a maximum _do.se_of approximately 35 mrem._ This i i dose is small in comparison with the 500 mrem /yr dose restriction for persons i

4 li in unrestricted areas specified by the National Comittee on Radiation Protec-ii tion and Me'asurements (NCRR) and the International Comittee on Radiological Protection (ICRP).

During the recovery phase, the maximum exposed individual would likely be a rigger fastening chains to the lifting lugs of the cask in order to retrieve l~

it. For brief intervals only, this person may be closer than 3 feet. Assume

]

that the average exposure is equivalent to being at 3 feet for 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The O.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is considered to be relatively conservative time period since the fastening of chains is a simple procedure. The dose then received by the 4

1, maximum exposed individual would be 500 mrem. However, the licensee has 1

comitted to providing health physics surveillance at the scene, and the

!j rigger (maximum exposed recovery person) would likely be a licensee employee who is under the protection of the licensee's radiation protection program.

' i In that case, a radiation dose standard of 5000 mrem /yr would apply. This i

500 mrem maximum exposure is but one-tenth of the radiation standard for l!

i occupational exposures.

3-7

3.2.5 Risk to Public j

Assume that the licensee's Recovery Coordinator who accompanies each shipment will restrain onlookers from approaching the cask at the scene of the accident.

Nevertheless, assume a bystander were to remain within 25 feet of an accessible surface of a loaded cask for 0.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

The dose rate at that distance would y

4

9:

be approximately 15 mrem /hr. Under those conditions, the bystander would 1

i receive a dose of about 7 mrem. This dose is about one-fifth of the radiation received by an individual during a chest X-ray and would not have a detectable

~

health effect, i

3.2.6 Cask Ruptured In the unlikely event of an accident severe enough to rupture a cask con-taining spent fuel elements, the radiological consequences are described in

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Sandia National Laboratories Report SAND 83-0867, "A Preliminary Analysis of the Cost and Risk of Transporting Nuclear Waste To Potential Candidate Comercial Repository Sites," June 1983, Reference 7.

Reference 7 states "In some recent experiments, contents of a simulated shipping cask for spent fuel were forced out through an opening in the cask...

The opening was considerably larger than could result from an accident... However, the amount of material that could be forced out was so small that, if released under the worst possible meterological conditions and in an ultrahigh density urban area, no imediate fatalities would result.

Experimental evidence combined with conservative (producing the worst impact) assumptions indicate that only one delayed fatality would result." The population density here assumed was taken from Sandia National Laboratories Report SAND 82-2365, "An Assessment of the Safety of Spent Fuel Transportation in Urban Environs," June 1983, 3-8

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Reference 9, was that for the Manhattan borough of New York City -- a much greater density than that along the proposed transshipment route. Note also that the quoted analysis assumed spent fuel cooled'only 150 days, not the 2-year cooling of the Surry fuel prior to transshipment. This minimal consequence, coupled with the extreme improbability of cask rupture, results in a negligible risk from such an occurrencem.

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4.0 Nonradiological Assessment for Transshipment of Surry Fuel In this section, those occurrences are considered that would be encountered if the transport vehicle were hauling dogfood, television sets, etc. instead of spent fuel. These dangers include vehicular accidents, pollution from vehicle emissions, and possible hijacking attempts.

4.1 Risks from Vehicular Accidents The greatest risk to health and safety presented by the proposed trans-shipments lies not in the radioactive material being shipped but in the risk of death or injury from traffic accidents. The most recently available analysis of vehicular-accident fatalities and injuries is contained in Reference 7.

The shipment by truck of spent fuel from centralized reference points to Hanford, Washington was the basis of these data. The postulated

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shipments involved a total shipment distance of 320 million kilometers, or 200 million miles (Table 2, p. 9 of Reference 7). For these shipments, 37.3 nonradiological fatalities were predicted from accidents, 8.3 of which were to occupational workers (Table 26, p. 42 of Reference 7).

i The longest proposed route from Surry to North Anna is 177 miles, or 354 miles round trip. Assuming three vehicles per shipment, the 167 round trips necessary to complete the transshipment would entail 177,000 vehicle-miles. Using the fatality rates from Reference 7, the expected number of fatalities would be O.011, of which 0.0024 would be transport workers.

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Reference 8 also predicts injuries from accidents. For the proposed trans-shipments, these statistics would predict 0.14 total injuries, 0.005 of which would be to transport workers. The chances are thus 1 in 200 (0.005) that a transport worker would be injured in.a traffic accident.

I For verification, these estimates were checke.d.with those calculated from I

l the data given in Reference 3, p. 65. The accident rate cited for trucks i

carrying hazardous materials was 1.69 accidents per million vehicle-miles.

In 177,000' vehicle-miles, one would expect 0.30 accidents. The same source gives fatality and injury rates per accident of 0.039 and 0.51, respectively.

These rates would result in 0.011 fatalities and 0.142 injuries. The results of these estimates are:

Based on data from Ref. 3 Ref. 7 Total vehicular 0.009 0.011 l

accident fatalities Total vehicular 0.15 0.14 accident injuries l

' l The risks from the above estimated vehicular accidents are believed to be acceptably low.

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a-s 4.2 Risks From Air Pollution l

Pollutants emitted by the shipment vehicles can contribute to latent cancer fatalities (LCFs). Data in Reference 7 predict 0.7 LCFs in shipping spent fuel 200 million miles. For the 177,000 miles (longest route) for the trans-shipments, one would expect 0.00021 LCFs or about one chance in 5,000 of a j

single death. This risk is negligible when compared with the expected 1 in 5 (0.20) ' chance of a death from cancer during the life of a member of the public.

(Cancer Facts and Figures, American Chemical Society, 1981, p.7.)

t 4.3 Risks From Increase In Traffic Density The expected traffic density for various highway types was cited in Section i

3.1.2.

The increased traffic on the day of a shipment would be less than 0.01 percent for a six-lane interstate highway, where the traffic would be the greatest, and only a fraction of one percent on two-lane rural roads.

This small increase is inconsequential when compared with nonnal fluctuations in traffic flow and does not. appear to constitute an unreasonable risk to the safety of the public.

4.4 Risk From A Possible Terr'orist Act The gunplay that might take place in the event of a postulated terrorist attack on a spent fuel shipment could resull: in the death of one or more drivers,

'i escorts, and persons incidentally in the vicinity. These consequences cannot be minimized. However, so far as is known, the hijacking or sabotage of a

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spent fuel shipment has never been attempted. Accordingly, the probability i

1 of such an attempt cannot be quantified on the basis of historical data.

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The following considerations would indicate that the probability of such an event is remote:

(1) Extensive safeguards precautions minimize the probability of success.

(2) Attempted sabotage, even if successful, would not produce serious radiological consequences (See Section 3.2.6).

(3)

Attempted theft and separation of plutonium or fission products, even if successful, would require complex equipment and time-consuming reprocessing.

(4) The size and weight of the cask and the intense radioactivity of its contents would strongly mitigate against the successful theft of the spent fuel.

On the basis of these considerations, the risk to transportation workers 1

and to the public from a possible terrorist attack is regarded as very small.

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5.0 Sumary The staff's evaluation supports the conclusions that the proposed modifica-tions to the spent fuel pool for the North Anna Power Station and the trans-shipment and storage of the Surry Power Station spent fuel to North Anna is acceptable because:

(1) The structural design and the materials of construction for the spent fuel modifications are adequate and meet the applicable design criteria.

(2) The installation and use of the high density racks will not result in any l

new means of losing fuel pool integrity or cooling water which has not already been addressed in the licensee's FSAR and the NRC SERs.

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(3) The installation and use of the high density racks will not create the possibility of a new or different kind of accident whose consequences would exceed those previously analyzed.

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l (4) The physical design of the high density racks will preclude criticality for any moderating condition with the limits imposed (for 15x15 and l

17x17 fuel assemblies).

(5) The spent fuel pool cooling system has adequate cooling capacity.

(6) The installation and use of the new high density racks can be accomplished safely.

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(7) The likelihood of an accident involving heavy loads in the vicinity of the spent fuel pool do not exceed radiological consequences of fuel

]1 handling accidents previously evaluated and are well within the guidelines of 10 CFR Part 100 and are acceptable.

i (8) The small increase in radiation exposure.due to the installation and use of new high density racks should not affect the licensee's ability to maintain individual occupational exposures as low as is reasonably achiev'able (ALARA) and within the limits of 10 CFR Part 20.

1 (9) The small increase in radiation exposure due to the storage and loading /

i unloading of additional spent fuel should not affect the licensee's t

ability to maintain individual occupational exposures as low as is l

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reasonably achievable (ALARA) and within the limits of 10 CFR Part 20.

i i i (10) The proposed modifications to the spent fuel pool for increasing spent 1

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fuel storage does not mitigate the required level of protection from industrial sabotage.

(11) The cumulative exposure to all transport workers engaged in normal transport conditions is well within the 500 mrem /yr dose restrictions for persons i

in unrestricted areas as specified by the National Connittee on Radiation 1

Protection and Measurements and the International Connittee on Radiation 1

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Protection.

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(12) The cumulative exposure to the public for normal transport conditions i

is negligible in comparison with the cumulative exposure that this l

population receives from natural background radiation, (13) The probability of a severe accident to a vehicle transporting spent fuel g

which could conceivably result in cask rupture is considerably less than one in a million.

(14) The do~se rate to the maximum exposed individual engaged in rescue and recovery operations from a severe accident is but one-tenth of the radiation standard for occupational exposures.

(15) The dose rate to the public (bystander) of an accident is postulated to

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be one-fifth of the radiation received by an individual during a chest

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X-ray and would not have a detectable health effect.

(16) The risks from estimated vehicular accidents (non-radiological) are believed to be acceptably low.

(17) The risk to transportation workers and the public from a possible terrorist attack is regarded as very small.

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6.0 Conclusion l

Based on the considerations discussed above, we conclude that:

(1) there is reasonable assurance that the he51th and safety of the public will not be-endangered by operation and transport in the proposed manner, and (2) such activities will be conducted in compliance with the Comission's regulations m

and.will not be inimical to the connon defense and security or to the health and safety of the public.

Principal Contributors:

L. Engle, Project Manager, DL/NRR L. Bell, Accident Evaluation Branch, DE/NRR W. Brooks, Core Performance Branch, DSI/NRR R. Fell, Meteorology and Effluent Treatment Branch, DSI/NRR C. Hinson, Radiological Assessment Branch, DSI/NRR j

0. Rothberg, Structural and Geotechnical Engineering Branch, DE/NRR E. Sylvester, Auxiliary Systems Branch, DSI/NRR B. Turovlin, Chemical Engineering Branch, DE/NRR J. Long, Advanced Fuel and Spent Fuel Licensing Branch, DFCMS/NMSS G. Terry, Advanced Fuel and Spent Fuel Licensing Branch, DFCMS/NMSS F

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APPENDIX A CHRONOLOGY OF RADIOLOGICAL REVIEW REGARDING SPENT FUEL EXPANSION s:

AND STORAGE OF SURRY POWER STATION SPENT FUEL AT THE NORTH ANNA POWER STATION NOTE:

Documents referenced in this chronology are available for public inspection and copying for a fee at the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the Public Document

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Rooms located at Board of Sup~ervisors Office, Louisa County Courthouse, Louisa, Virginia 23093 and the Alderman Library, Manuscripts Department, University of Virginia, Charlottesville, Virginia 22901 1

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July 13, 1982 Letter from R. H. Leasburg (licensee) to H. R. Denton, NRC,

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for a license amendment to permit the storage of 500 Surry spent fuel assemblies at NAPS.

i July 13, 1982 Letter from R. H. Leasburg (licensee) to Robert F. Bernett (sic).

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July 28, 1982 Letter from Theodore S. Sherr, NRC, to R. H. Leasburg (licensee).

August 20, 1982 Letter from R. H. Leasburg (licensee) to H. R. Denton, NRC, for a license amendment to modify spent fuel storage to 1737 fuel assemblies.

October 21, 1982 Letter from R. H. Leasburg (licensee) to H. R. Denton, NRC,

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transmitting additional information on spent fuel pool heat loads.

June 16, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, transmitting responses to NRC questions.

July 19, 1983 Letter from R. Clark, NRC, to W. L. Stewart (licensee) requesting information regarding the transshipment of fuel from Surry to North Anna.

July 25, 1983 Letter from R. Clark, NRC, to W. L. Stewart (licensee) requesting additional information regarding spent fuel storage expansion.

September 13, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, transmitting responses to NRC request for additional information.

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October 28, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, transmitting responses to NRC request for additional information.

November 10, 1983 Letter from J. R. Miller, NRC, to W. L. Stewart requesting g.

additional information regarding spent fuel storage capacity.

November 23, 1983 Letter from W. L. Stewart to H. R. Denton, NRC, advising when NRC request for additional information will be provided by licensee.

December 6, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, providing additional information regarding spent fuel capacity expansion.-

December 6, 1983 LetterfromW.L. Stewart (licensee)toH.R.Denton,NRC, providing additional information regarding increase in spent fuel capacity.

December 14, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, correcting typographical errors in the licensee letter dated December 6,1983.

Decent :r 14, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC,

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providing clarification for neutron absorber spent fuel rack.

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O December 29, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, providing additional information regarding spent fuel capacity expansion.

December 29, 1983 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, providing additional _inf.ornation of proprietary nature regarding neutron absorber racks.

April 10, 1984 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC,

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providing additional information regarding spent fuet expansion.

May 8, 1984 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC, providing additional information regarding available spent fuel storage.

May 18, 1984 Letter from W. L. Stewart (licensee) to H. R. Denton, NRC,

.providing additional information regarding low level radwaste.

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APPENDIX B REFERENCES k?

(1)

J. R. Weeks, " Corrosion of Materials in Spent Fuel Storage Pools,"

BNL-NUREG-23021, July 1977.

(2)

J. S. Anderson, Brand Industries, Inc., Reports 748-2-1 (August 1978),

748-10-1 (July 1979), 748-30-1 (August 1979).

(3) USAEC Report WASH-1238, " Environmental Survey of Transportation of

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Radioactive Materials To and From Nuclear Power Plants," December 1972.

i (4) USNRC Report NUREG-0170, " Final Environmental Statement on the Transportation of Radioactive Material by Air and Other Modes", December 1977.

(5) Rand-McNally Road Atlas,1982.

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(6)

U. S. Department of Commerce, " County and City Data Book",1977.

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(7) Edwin L. Wilmot, Marcella M. Madsen, Jonathan W. Cashwell, and Davis S.

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Joy, Sandia National Laboratories Report SAND 83-0867, "A Preliminary Analysis of the Cost and Risk.of Transporting Nuclear Waste To Potential Candidate Commercial Repository Sites", June 1983.

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(8)

R. P. Sandoval, J. P. Weber, H. S. Levine A. D. Romig, J. D. Johnson, R. E. Luna, G. J. Newton. B. A. Wong, R. W. Marshal, Jr., J. L. Alvarez, and F. Gelbard, Sandia National Laboratories Report SAND 82-2365, "An Assessment of the Safety of Spent Fuel Transportation in Urban Environs,"

..___ June 1983.

(9)

Letter dated November 7,1983, from C. V. Parks, Union Carbide Corporation, Nuclear Division, to R. H. Odegaarden, U.S. N.R.C.

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