ML18139C041
| ML18139C041 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/30/1982 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML18139C039 | List: |
| References | |
| NUDOCS 8209270361 | |
| Download: ML18139C041 (46) | |
Text
.e
\.
- l .
A
SUMMARY
OF INFORMATION IN SUPPORT OF THE HANDLING OF SPENT FUEL CASKS IN THE SURRY POWER STATION UNIT.NOS. l AND 2 FUEL BUILDING Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 SEPTEMBER, 1982 VIRGINIA ELECTRIC AND POWER COMPANY
!EZ ti8Jii~~gggjg~ ** *-.
9 PDR
't .
TABLE .OF CONTENTS PAGE 1.0 Introduction 3 2.0*Genera1 Information 7 3.0 Ana1ysis of Damage to the Spent Fue1 8 4.0 Ana1ysis of Damage to Equipment 17 5.0 Analysis of Radio1ogica1 Consequences 17 6.0 Criticality Ana1ysis 20 7.0 Summary 24 References 26 Appendix A, "Spent Fue1 Cask Drop Evaluation 27 for the Decontamination Building and Truck Loading Area, Surry Power Station Units 1 and 2 11 2
brh/0010S/2
,~ . ','
1.0 INTRODUCTION
- 1. 1 Purpose of the Request The Technica1 Specifications for Surry Power Station Units 1 and 2 (T.S.
3.10-A.13) pres~nt1y prohibit movement of a spent fue1 cask into the Fue1 Bui1ding unti1 the NRG has reviewed and approved a fue1 cask drop eva1uation.
The purpose of this document is to provide a summary of a*postu1ated spent fue1 cask drop and to provide justification for the de1etion of Technica1 Specification 3. 10-A. 13 and for the substitution in its p1ace of the fo11owing:
"A spent fue1 cask sha11 not be moved into the Fue1 Bui1ding un1ess the Cask Impact Pads are in p1ace on the bottom of the spent fue1 poo 1."
This.Spent Fue1 Cask Drop Eva1uation wi11 address the potentia1 consequences of a postu1ated cask drop in the Fue1 Bui1ding to satisfy the eva1uation criteria of NUREG-0612, Section 5.1. APPENDIX A to this* e,valu"ation addresses the potentia1 consequences of a postu1ated cask drop in the Decontamination Bui1ding and truck 1oading areas. These areas are 1ocated a1ong the trave1 path for a spent fue1 cask and a drop in these areas must be eva1uated as per*
the guide1ines of NUREG-0612.
1.2 Background and Scope NUREG-0612, 11 Contro1 of Heavy Loads at Nuc1ear Power P1ants," (reference 1) contains guide1ines which shou1d be satisfied to ensure the safe hand1ing of 3
brh/OOlOS/3
'\ . **
heavy loads including spent fuel casks. This Cask Drop Evaluation will address the NUREG-0612 guidelines for load handling operations in the spent fuel pool area. Various alternatives are provided in NUREG-0612 (Section 5.1.2) to assure that the consequences of a cask drop in the spent fuel pool area are within acceptable limits. Alternative (3) was selected for the Cask Drop Evaluation. Analyses of postulated load drops and the evaluation of potential consequences under Alternative (3) must meet the following guidelines:
- l. "Hot" spent fuel should be concentrated in one location in the spent fuel ~ool that is separated as much as possible from load paths *.
- 2. Mechanical stops or el~ctrical interlocks should be provided to prevent movement of the overhead crane load block over or within 25 feet horizontal (7.5 m) of the "hot" spent fuel. To the extent practical, loads should be moved over load paths that avoid the spent fuel pool and kept at least 25 feet (7.5 m) from the "hot" spent fuel unless necessary. When it is necessary to bring loads within 25 feet of the restricted region, these mechanical stops or electrical interlocks should not be bypassed unless the spent fuel has decayed sufficiently as shown in Table 2.1-1 and 2.1-2, or unless the total inventory of gap activity for fuel within the protected area would result in offsite doses less than 1/4 of 10 CFR Part 100 if released, and such bypassing should require the approval from the shift supervisor (or other designated plant management individual). The mechanical stops or electrical interlocks should be verified to be in place and operational prior to placing "hot" spent fuel in the pool.
4
3.
- e Mechanical stops or electrical interlocks should be provided to restrict crane travel from areas where a postulated load drop could damage equipment from redundant or alternate safe shutdown paths.
Analyses have demonstrated that a postulated load drop in ~ny location not restricted by electrical interlocks or mechanical stops would not cause damage that could result in criticality, cause leakag*e that could uncover the rue 1, or cause loss of safe shutdown equipment.
- 4. To preclude rolling, if dropped, the cask should not be carried at a height higher than necessary and in no case more than six (6) inches (15 cm) above the operating floor ievel of the refueling building or other components and structures along the path of travel."
The guidelines for performing the required analyses are contained in Appendix
. A to NUREG-0612, "Analyses of Postulated Load Drops".
The results of these analyses must assure that the following criteria are met:
- 1. Re leases of radioactive material that may .result from damage to spent fuel based on calculations involving accidental dropping of a postulated heavy load produce*doses that are well within 10CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than one quarter of Part 100 limits);.
- 2. Damage to fuel and fuel storage racks based on calculations involving accidential dropping of a postulated heavy load does not result in a configuration of the fuel such that Keff is larger than 0.95; 5
brh/0010S/5.
3.
Damage to the spent fuel pool based on calculations of damage following accidental dropping of a postulated heavy load is limited so as not to result in water leakage that could uncover the fuel (makeup water provided to overcome leakage should be from a borated source of adequate concentration if the water being lost is borated);
and
- 4. Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited so as not to result in Toss of required safe shutdown functions.
The following general guidelines (NUREG-0612, Section 5.1. 1) for handling of heavy loads in the spent fuel pool area were addressed in the VEPCO Six Month and Nine Month Report~ for NUREG-0612. (References 2 and 3)
- l. Safe load paths
- 2. Procedures
- 3. Crane operator training and qualification
- 4. Design of special lifting devices
- 5. Selection and use of lifting devices
- 6. Crane inspection, testing and maintenance
- 7. Crane design Although a cask drop is considered very unlikely, this evaluation assumes the cask is accidentally dropped (single-failure criterion) over the spent fuel pool.
6
2.0 GENERAL INFORMATION A 125-ton capacity Fuel Building Trolley is provided for handling spent fuel shipping casks. This trolley runs on fixed rails from the cask loading area in the Fuel Building, through the Decontamination Building to a truck loaqing
.area in the yard.
The cask loading area is located in the northeast corner of the spent fuel pool as shown on the Fuel Buildin~ arrangement drawings(Figures 2.1 and 2.2).
Spent fuel racks are excluded from the area under the Fuel Building Trolley rails. The trolley is fixed at the center of its span and can move.only in a
,north-south direction which prevents moving a cask directly over spent fuel.
No mechanical stops or electrical interlocks are necessary to prevent the movement of a *spent fuel cask directly over stored spent fuel.
The Fuel Building is a Seismic Class I structure~ Restraints are provided to prevent displacement of the Fuel BuUding Trolley from the rails during a seismic event.
The Fuel and Decontamination Building ventilation systems are designed to maintain a negative pressure for inward leakage. The ventilation exhaust from*
these buildfngs may be directed through a charcoal filter bank or bypassed.
directly to the monitored Ventilation Vent System. The Technical Specifications require that the Fuel Building exhaust be directed through the charcoal filters during refueling operations. The Decontamination Building exhaust would be remote manually directed through the filters upon receipt of 7
brh/001 OS/T
1.
the ventilation vent high radiation alarm.
No credit is taken for the charcoal filters in determining potential offsite doses.
The Fuel Building and Decontamination Building systems are also described in the Surry 1 and 2 FSAR in Sections 9.12* and 9. 14.
Two casks were considered for this cask drop evaluation, in order to envelope the range of casks (storage or shipping) which potentially could be used at Surry.* The two casks are the Transnuclear, Inc. Model TN-2100 (proposed design) and the Gesellschaft Fur Nuklear Service Model GNS-5. Even though these are botti storage casks, this evaluation bounds the range of cask drop analyses for all casks that could potentially be used at Surry.
This cask drop evalua~ion addresses the following areas:
3.0 Analysis of Damage to the Spent Fuel Pool 4.0 Analysis of Damage to Equipment 5.0 Analysis of Radiological Release 6.0 Criticality Analysis 3.0. ANALYSIS OF DAMAGE TO THE SPENT FUEL POOL 3.1 Assumptions The following assumptions were utilized for evaluating cask drop consequences to the fuel pool structure. (reference 4) 8 brh/0010S/8
- 3. 1. l e.
Wei"ght of Spent Fuel Cask Calculations were based on the largest casks (storage or shipping) proposed for use at Surry Power Station. Models TN-2100 (storage 120 tons) and GNS-5 (storage 115 tons) were included. Weights used represent fully loaded casks with lifting yokes and associated rigging attached.
- 3. 1.2 Drop Location The cask drop accident is postulated to occur at any point alqng the cask travel path in the Fuel Building. Since the cask will be carried at the minimum height necessary, it will fall less than 6 in.
El.I'"
outside *the fuel pool wall and will remain upright for all postulated drops. Once the cask center of gravity is over the fuel pool, it is postulated to fall into the pool. The most likely drop location is assumed to be directly above the cask loading area in the hortheast corner of the fuel pool as shown on Figures 2.1* and- 2.2. Although the cask could be dropped at any location along the travel path of the 125-ton Fuel Building Trolley, the most severe consequences to the pool would result from a drop straight down into the pool.
Calculations have assumed both an edge drop (cask at worst case angle) and a flat dr~p of the cask. For cas~ drops over the north edge of the pool, .the cask is assumed to fall toward the south into the pool at an angle. The worst case edge drop covers this case.
Since the cask lifti~g trunnions and yoke are maintained in a north-south orientation by administrative control, the cask would tip 9
__ bi:_h/OOlOS/9.
either north or south upon the postulated failure of those items. A
_drop under these conditions would be an angle drop; The cask is considered to drop straight upon the postulated failure of the ~rane 0
hook and wire ropes.
- 3. 1.3 Drop Height The top of the cask impact pad in the fuel pool cask loading area is at elevation 6 ft-6 in. The cask is assumed to drop from an elevation of 48 ft-4 in. From this elevation the cask will fall through 3 ft-6 in. of air and 38 ft-4 in. of water before striking the cask impact pad. The 48 ft-4 in. elevation will al-low the cask to clear the crane rail (elevation 47 ft-9 1/2 in.) for the fuel handl_ing mot~r-driven platform. The floor elevation at the edge of the fuel pool is 47 ft-4 in. A h~ight restriction of 6 in. maximum above the crane rail will be maintained by administrative control during cask handling operations~
3.1.4 Impact Area of Load The pool floor in the cask loading area is protected by two (2) cask impact pads. These pads are stainless steel platforms designed to absorb most of the impact energy of a dropped cask and limit structural damage to the pool floor. The cask impact pads, as shown in Figures 3.1, 3.3 and 3.4, are constructed of energy absorbing pipes between two steel plates.
10 br h/0 OJOS1l.O__
\ .
The impact areas at the fuel pool floor were determined by the geometry of worst case striking the impact pad with flat drop and*
edge drop orientations.
The worst case cask for the fuel pool floor analysis was determined to be the GNS-5 cask (115 tons) because its 6 ft-4 1/2 in. base diameter could engage the fewest number of pipes, thus delivering its load to the floor over a smaller area. The heavier cask, the.TN-2100 (120 tons) with a base diameter of 7 ft-11 in. would engage at least
- three pipes, thus distributing its slightly larger impact energy over a significantly larger impact area~
The impact area for the cask hitting the fuel pool walls was determined from the geometry of the TN-2100 cask since it transmits a greater energy than the GNS-5 cask. It was conservatively assumed that all the ~nergy from the cask hitting the bottom of the pool was transmitted to the wall as rotational energy. The TN-2100 cask was assumed to fall a~ an orientation causing the g*reatest acceleration into the wall after striking the pool bottom.
The cask drop accident scenario and the cask loading area for Surry Power Station are shown in Figures 3.2 through 3.4. Under the postulated cask drop~ the cask falls from an elevation 1 ft. above the top of the spent fuel pool wall onto a cask impact pad. The cask impact pad absorbs most of the kinetic energy and prevents damage to the fuel pool floor. Depending upon the angle of fall the cask may theoretically then (1) remain upright, (2) tip in an east or north 11 brh/OOlOS/11 .
direction and strike the wall, (3) tip to the south and. continue to fall over or (4) tip to the west and strike the fuel racks. The angle drop is assumed to be in the north-south direction because the lifting trunnions will be administratively positioned in the north-south direction during cask handling, such that a failure of the trunnions or lifting yoke will impart tipping rotation to the cask during a drop only in the north-south direction and not east-west. *since the craMe hook cannot move in an east-west direction by design, and the trunnions will be oriented in a north-south direction, no tipping in the east-west direction is postulated. However, this analysis evaluates the damage to spent fuel stored in the spent fuel racks adjacent to the cask handling area as required by NUREG-0612.
3.1.5 Walls and Floor Slabs The floor and walls of the spent fuel pool are reinforced concrete, 6 ft thick. The walls of the pool are lined with 1/4 in. stainless steel plate. The floor of the pool in the cask loading area is lined with 1/2 in. stainless steel plate.
3.1.6 Environmental Forces Credit is taken for buoyancy and drag forces of the fuel pool water in determining the impact velocities for the dropped casks. The water level used for this analysis was conservatively assumed to be the low alarm level for the pool, elevation 44-ft 10 in.
12
~ brh/OOlOS/12
', (
- 3. 1.7 Loads The loads considered for this analysis included the heaviest spent fuel cask (Transnuclear TN.:.2100) expected to be used at Surry Power Station and lifted by the Fuel Building Trolley (125-ton capacity).
This spent fuel cask drop evaluation envelopes the consequences of.
droppfng the Irradiated Specimen Shipping Cask (11.3 tons), as well as all other spent fuel:casks (shipping or storage) expected to be used at Surry Power. Station *
- 3. 1.8 Material Properties The structural analysis considers a concrete compressive strength of 4,000 psi based on actual laboratory test specimen results. No credit is taken for the energy absorbing effects of the stainless stee 1 1i ner.
The design of the.cask impact pads uses the deformati6n of large pipes as the energy absorbing mechanism. The cask impact pads consist of 24 in. diameter, Schedule 160, stainless steel pipes sandwiched between 1 1/2 in. thick stainless steel plates. The technology for this form of energy absorption device (pipe defonnation) is well _established and is used in the design of pipe rupture restraints.
13
. brh/OOlOS/13 .
- 3. 1.9 Makeup Water Sources Norma1 makeup water for the spent fue1 poo1 is provided by the.
Primary Grade Makeup Water System. There are two Primary Water Storage Tanks with a capacity of 180,000 ga11ons each. The makeup capabi1ity from this source is 200 gpm.
Makeup water is a1so availab1e from the Refue1ing Water Storage Tanks which contain at 1east 350,000 ga11ons of borated water each. These tanks may be 1ined up to supp1y up to 600 gpm from the Refue1ing Water Recircu1ation Pumps discharge.
Emergency makeup water is avai1ab1e from the Fire Protection system at a rate of_ up to 2000 gpm. This water is not borated.
3.2 Methods of Ana1ysis 3.2. 1 Fue1 Poo1 F1oor The structura1 ana1ysis of damage to the fu~l pool mat included the mitigating effects of the cask impact pad. The two (2) cask impact pads are 1ocated in the cask 1oading area and ar'e designed to 1imit structural damage from an accidental cask drop. Based on the cask impact pad crush characteristics, the peak 1oads and energy over the approximate impact areas were determined for the worst case edge and f1at drops.
14 br h/0 O10S_fl4
Due to the large span (12 ft) to thickness (6 ft) ratio and the short accident duration time, shear failure was assumed to govern the analysis. Therefore, punching shear failure was investigated for the peak loads which occur just prior to stopping cask movement (for flat drop) and at the instant of total crush (for edge drop), just prior to delivery of the excess energy to the floor. In the case* of excess energy, an impact analysis was gone *for the fuel pool mat based on Stone &Webster Engineering Corporation (SWEC) Topical Report
- "Missile-Barrier Interactiori 11 (SWEC0-7703). This was to determine the possibility of scabbing and depth of indentation into the concrete.
3.2.2 Fuel Pool Walls The analyis of damage to the spent fuel pool .walls conservatively assumed that all enefgy from the cask hitting the bottom of the pool was transmitted to the wall as rotational energy, i.e., no energy is lost at floor imp~ct _due to cask pad interactio*n. The cask was assumed to drop to the bottom of the pool-at the orientation which could cause the greatest ~ngular acceleration into the wall. The applied impact energy was then calculated based upon the resultant velocity and the total mass (conservative). For this impact energy, SWE*C Topical Report .Missile-Barrier Interaction" (SWEC0-7?03) was used to predict concrete scabbing. Also, depth of penetration was determined based on the cask geometry, the angle of impact and the crushing resistance of concrete, neglecting the steel liner's resistance to perforation.
15 brh/OOlOS/15 .
e e For both the floor and wall analyses, the cask was conservatively assumed to be rigid, thus absorbing* no energy by deformation.
3.3 Conclusions This analysis satisfies criterion 3 of NUREG-0612, Section 5.1.
3.3.1 Fuel Pool Floor Due to the resulting impact load from a flat cask drop onto the impact pad, a shear plug in the fuel pool floor is not anticipated.
For an edge drop, concrete scabbing is not expected and a maximu~
concrete indentation bf 0.7 in. over the impact area is not expected to tear.or perforate the steel liner.
Fuel Pool Walls Analysis of damag! to the pool walls, neglecting the ~itigating effects of the cask impact pads, shows that a cask drop would not cause scabbing, but penetration into the wall is possible. The
/maximum expected penetration depth is 4.3 in. over an impact length of 61 in. This would cause pool wall liner damage, but no significant leakage, _since no through-wall cracking is predicted.
Sufficient borated makeup water is available to ensure the fuel remains covered.
16 L-1.. tnn,nc- L1t:
e 4.0 ANALYSIS OF DAMAGE TO EQUIPMENT Fuel Pool Cooling System piping in the pool and piping trench at the northeast corner of the pool could be impacted by a cask drop. Damage to this piping will not result in draining of the pool. Fuel Pool Cooling discharge piping is accessible in the pipe trench. Temporary repairs would allow this sytem to continue operation. If necessary, temporary makeup to the spent fuel pool using unborated water would maintain cooling and shielding while repairs are made to damaged piping. There are no other safe shutdown systems or .components in the cask travel path.
This satisfies criterion 4 of NUREG-0612, Section 5.1.
5.0 ANALYSIS OF RADIOLOGICAL CONSEQUENCES In order td evaluate the radiological cohsequences of the postulated cask drop accident it was assumed that the gap activity for the fuel assemblies stored in the first three rows of spent fuel racks is released. This is assu.med in response to the guideline in NUREG-0612 that no 11 hot 11 spent fuel be stored within 25 feet of the Fuel Building Trolley load block.
The first 3 rows of racks encompass an area out to 28 feet from the load block. Therefore, this assumption is conservative.
5.1 Calculation of Offsite Dose Off-site doses resulting from the rupture of a .fuel assembly following a cask drop accident have been calculated (Reference 5). The calculations 17
e.
were performed in accordance with the requirements of Regulatory Guide 1.25 and Standard Review Plan Section 15.7.5. The following assumptions were used in the calculations:
- 1. All gap activity is released. Gap activity consists of 30 percent of the total Kr-85, ten percent of the total of other noble gases and ten pefcent of the radioactive.iodine.
- 2. Fuel fission produ~t invento.ries were calculated using the LOR2 computer code.
- 3. The effective pool decontamination factor for iodine is 100.
- 4. The radial peaking factor. is 1.2. Core-rated power is 2546 MWT.
- 5. No credit was taken for decay in route to the receptor.
- 6. The exposure of f~el ruptured is 45,000 MW-day/Metric Ton.
- 7. The atniospheric dispersion for the site boundary is 2. lxl0- 3 sec/m 3* {Reference 6).
- 8. No *credit is taken fqr termination or mitigation of radiological releases from the Fuel Building.
Figures 5.1 and 5.2 present the Total Body and Adult Inhalation Thyroid doses resulting from rupture of one assembly as a function of decay time 18 brh/OOlOS/18 . .
l following discharge.
Figure 5.3 has been derived from Figures 5.1 and 5.2 and shows the number of assemblies that must be ruptured to yield an off-site dose equal to 25 percent of 10 CFR 100 limits versus decay time from discharge.
- 5.3 Conclusions Offsite doses from the postulated accident will be maintained well below the limits of 10CFRlOO by*ensuring that fuel assemblies stored in the first three rows of racks, which are conservatively assured to be damaged from a postulated cask drop, and fuel assemblies placed in the cask, have been allowed sufficient decay time following shutdown. Based on the results of this analysis, all new fuel and spent fuel which has not decayed for at le~st 150 days will be excluded from the first three rows of racks, which include the area within 28 feet of the Fuel Building Trolley load block. The 28 foot separation limit meets the *requirements of NUREG-0612 Section 5. 1.2 Alternative (3) and is greater than the
- guideline distance of 25 feet. Spent fuel which has not decayed for at least two years will not be loaded into a spent fuel shipping cask. These provisions will be maintained through administrative controls.
The first three rows of racks contain nine racks providing a total of 324 spent fuel storage cells. As shown on Figure 5.3, at least 2000 assemblies must be damaged to reach 25 percent of 10 CFR 100 Limits, if allowed to decay for 150 days after shutdown.
This satisfies criterion 1-of NUREG-0612, Section 5.1.
19 brh/0010S/19
e 6.0 CRITICALITY ANALYSIS
/
A criticality analysis was perfonned for spent fuel assumed to be stored in the first three rows of spent fuel racks. The method of analysis is set out below (reference 5).
6.1 Method* of Analysis 6.1.1 KENO-IV Code The Monte Carlo Code KENO-IV was used to evaluate Keff for the criticality configurations considered in this analysis. KENO-IV is a three-dimensional, multi-group criticality code that solves the Boltzmann transport equation to determine Keff values. The code is available from Oak Ridge National Laboratory Radiation Shielding Information Center. Working cross sections were generated using NITAWL and the 123 group XSDRN library.
6.1.2 LOR2 Code The LOR2 generation and depletion code was used to calculate the fision and activation product inventory of exposed fuel. LOR2 is a modified version of ~he Oak Ridge program ORIGEN.
- 6. 1.3 Uncertainty and Benchmark Calculations I
The uncertainties in.Monte Carlo criticality calculations can be divided into two categories:
- ) 20 I
hrh/0010S/20
(1) Uncertainty due to the statistical nature of Monte Carlo calculations. (The v~riance of Keff values over many generations.)
(2) Uncertainty due to bias in the calculational techniques.
The variance of Keff values for many neutron generations is directly determined in the KENO-IV calculations. Its magnitude can be reduced by increasing the number of neutrons tracked. For fuel rack criticality calculations, the number of neutrons tracked is sele~ted to reduce the variance to less than 0.005.
KENO-IV has been extensively benchmarked for light water lattices.
It has been ~hown to be typically accurate to within one percent in Keff by the Babcock &Wilcox study performed for the DOE using 123 group cross sections. Similar benchmark calculations have been performed by NES on KENO-IV for critical experiments on light water lattices carried out at Battelle Northwest Laboratories. Results confirm that the KENO-IV code, with 123 group cross sections, yields conservative results typicallr within one percent of the experimental value. No credit will be taken for this bias in order to preserve the conservative nature of the results.
- 6. 1.4 Calculational Assumptions The following assumptions were used in the criticality calculations:
- 1. The pool water contains 2000 ppm of boron.
21 brh/OOJOS/21
e.
- 2. The fuel assemblies have no burnable poison.
- 3. For exposed fuel, only stable fission products were included in the calculations. The calculated Keff therefore represents an upper estimate of Keff for the exposed fuel.
- 4. ~o credit is taken for structural material other than the stainless steel storage cells.
- 5. All stainless steel storage cells are assumed to be 0.085 inches thick.
- 6. Calculations ass*ume 17 x 17 array fuel assemblies. Calculations have been performed which show the 17 x 17 Westinghouse design to have a Keff value of 0.007 higher than the 15 x 15 Westinghouse design of the same enrichment.
- 7. It was assum~d that any fuel stored in the first three rows of spent fuel racks would be spent fuel and any fuel in the remaining racks could be new fuel with an enrichment of up to 4.1 w/o U-235.
6.2 Cri-ticality Analysis and Results A matrix of criticality calculations was performed to determine the limiting fuel parameters that would allow fuel to be stored in fuel racks, subject to damage from a cask drop accident, without resulting in criticality. The parameters varied in the analysis were:
22 brh/OOlOS/22 .
Fuel Assembly Pitch Fuel Assembly Deformation (crushing of tne fuel assembly)
Initial Fuel Enrichment Fuel Exposure The analysis consisted of reducing the assembly to assembly spacing to maximize Keff for various combinations of initial e.nrichment and fuel exposure. The graph-presented in Figure 6.1 was derived from the calculations performed. The combinations of exposure and initial enrichment presented in Figure 6.1 assure that the maximum obtainable Keff *occurs for the conditions of Figure 6.1 when t~e assembly to assembly spacing in th~ spent fuel storage rack has been reduced to approximately 6.9 inches. This spacing assumes contact between the
,_, fuel storage rack cans and a uniform reduction of the fuel assembly fuel pi.n pitch.
6.3 Cone 1usions Based on tha results of the criticality analysis spent fuel which satisfies the initial enrichment vs burnup*limits in Figure 6.1 can be safely stored in the spent fuel racks .which may be subject to damage from a cask drop accident. A11 new fuel, and spent fuel which does not meet these criteria, will be excluded from an area within 28 feet of the Fuel Building trolley load block.
23
____k_h/0010S/23
The 28 foot separation limit is conse~vative in comparison to the guideline in NUREG-0612 for ~xcluding fuel not meeting the requirements of Figure 6. 1. These measures will ensure that accidental dropping of the spent fuel cask will not result in a configuration of the fuel such.that Keff is larger than 0.95. This satisfies Criterion 2 of NUREG-0612, Section 5. 1.
7.0
SUMMARY
In view of the equipment design, inspectidn procedures, and operating precautions that will be in place during spent fuel cask handling operations at Surry Power Station, the likelihood of a cask drop accident is extremely remote. Nevertheless, this analysis has evaluated the
~-
consequences of a_ postulated cask drop accident and has determined that these consequences are acceptable as follows:
- 1. The cask impact pads will protect the fuel pool floor from damage during a cask drop accident.
- 2. Tipping of a cask into the fuel pool wall following _a drop will cause only minor structural damage to the liner.
- 3. Sufficient sources of borated makeup water are available to compensate for any expected amount of leakage.
24 hrh/nnlOS/?4
\ ,. e.
- 4. If tipping of* a cask into the fuel racks following a cask drop is postulated such that damage will result to any or all of the first three rows of racks, fuel stored in these racks and damaged by the cask will not achieve criticality or result in unacceptable offsite doses if the stored fuel has sustained sufficient burnup and decay following shutdown as shown in Figure 6.1. Fuel not meeting these criteria will be excluded from storage in racks within 28 feet of the 125-ton Fuel Building Trolley.
- 5. Potential damage to the Fuel Pool Cooling System.piping along the cask load path will not result in unacceptable consequences.
25
REFERENCES
- l. NUREG-0612, 11 Control of Heavy Loads at Nuclear Power Plants, 11 USNRC,_June 1980
- 2. Vepco letter to Harold R. Denton, USNRC-ONRR, from R.H. Leasburg, Serial No. 688, December 22, 1981.
- 3. Vepco letter to Harold R. Denton, USNRC-ONRR, from R.H. Leasburg, Serial No. 171, March 22, 1982. ,,
- 4. 11 Spent Fuel Cask Drop Evaluation, 11 Stone &Webster Engineering Corp.,
September 1982.
- 5. 11 Surry Power Station Units 1 and 2 High Density Fuel Storage Racks/Cask Drop Accident Consequences, 11 NES, Inc., September 1982.
- 6. Stone &Webster Engineering Corp. letter to Vepco, 11 Reanalysis of LOCA Radiological Consequences, May_3, 1977, Serial No. NUS 7634.
11
- 7. 11 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition, 11 NUREG-75-087, USNRC, Section 15.7.4.
- 8. 11 Assumptions Used for Evaluating the Potential Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors, 11 USNRC Reg. Guide 1.25.
- 26
__ _ b_r-h/0010S/26
r1,;;~ f..~j
':U**: :l**lt!
10"1,
-,E ,,,..::~:.'
I.
Q
. .)
'"'--~-- > OAV':'t;* Ti:;:LV
- lElt::1c1CLL.
Q H *1*,o*
[l-11*.4
\.
l _jF -.1_
('M* se)
N--
L..
A (FM*96)
L-B (FM-98)
L..
D ::::
. r.,-:.*:,.,
\;,! __________ _
EL E.'*10** J~-tJ A e, r r B-B
__,______ ~
i.,,
~lOTS roP. ,, .... Tt 7,°;~:L~ J 2 MINlt-'.!Jt,.I
~ MO'IH~,_L
-k" *1"- o'E:l\.CEPT A~ l-lOTEO.
S.l"[N'1' f\.lf.L lt.J 1\.1.M.M&LIU, "
"""'"'°"
-~=-
11:LQUl'lEO !M,r[ .....n. .. M11n.o AAO.ll
,...:o:.
\.t.'-lf..\. U.& S.P[NT Ml\.1NlA.1NED .l.T tL 45'-..c.:, .
.. UP?li:t 5,1(.'Tlo+,I C'l1 c.,1,,Tt VALVI.
STtt-11 T01!11. ~ l . TO I\.LLOW
~OTOR. oa.tV!.H .,i_-,.TtcRt-1.
- 5. V. L.- IWOICATE.S YEATICAL LAODU~
rvtl.
PIT 10*&'
H *.-- -* .':-~f~of,A"'ii7'"°~ rt;~!,io"V STOfJ! *
- I E..,l,6.Qc,£D
~~~~~~~~~~~~~- --*--:~,--. ~ ~ 1 - ~ ****-~ .. * - - - - --'-* - - * '
PLAN AtH*R.£:UCE DWG~:
AUt(ILI.\.RV; BLOC. ""ARG'T rl.PfL BUILDING LH 2 r.t,a-or-
{J:M*5A)
D-D l_. PLAN EL 41'-4' L B D I/~
CONTA.INMfNT
~i~
- ,-~
~1~
, WA.Ll l ~
J/ ~ ~
- ~...........+- :-ii- "'
\. L~DCEA10 EL -1,*.10*
- ~
'- FUEL BUILDING ARRANGEMENT Sheet l Figure 2. 1
- 1*
~ ,1/1 .j . i *:.'.
- S17't' 11 *.*r ----
7 11 7:!.
A
.'I
"~ .,J,,
,~ r.
- --r.::~i t (
1 :I
- \ ... )
-1,.*,**-C TO*:.<<*o* --* 'Ii
"°L'> <'**
'* -~r----_.. ~.-i I 17"~---u~~
- /
J.;l:--- -- :
=r- - __ !1! _--~,
>,ew <UEL C"NE :,
~ - ~~
,11 u '
1 11
'i
\. E 10\
i
-- r-*.
J.KJTU: .
loOl.C: -o*-
. (,.(N(;llAl. NOT[":> , *<<nt. ...... .,~.J-~'if\:t-4 FUtl 1R>>6'Ut c.>>UJ..-
FUll ~LOO
[ 7,0SW
" : ~ . - - ~~~1*-c:.
c-c ct F'JEl TC.t..N?.HR 1UN, El B'*!O" 1,1,1,.(;/A'TlQII-Dfl((.11!),J
~.:ll.'" ~fUT
[lc:.*-i.c*
\.
FUEL BUILDING ARRANGEMENT Shrr>t. ~
- e. e .
Crane Hook Row Row Row ,~Fuel Cask 3 , 2 1 2
83~'t'12'
~ (typ)
I*
.--~__....;.,i Lower Cask Impact Pad 1
L rn 21'-4"
_lJ t.
Upper Cask Impact Pad Cask Loading Area I
Spent Fuel
.I.
Storage Area PLAN EL 6'-10" FIGURE 3.1 iiI
?,_ * - ,,,. *-* v o - ....... - - . , , . - - . -.w.,-..........,..,. ____ ,, __ ,.. ... , *-- --...-- *--* - . * , ...* *r******,n *,-,* ~ .. - . ,., .. :.," ---~ ,,
~~
..... e
- I I
I :
i.
I
- r. 125 Ton Trolley .- .
El 47'-4" Nonnal Water Level Elev 45'-:4" .';
Spent Fuel Stotage Racks Spent Fuel Shipping ~r Storage Cask
- Cask"-Impact Pad El 4 1 -4 11
.I. .., c.. * ? * *
. >"*...... t *J.. p.....
r,' ti. I.
0
' -~
.t. 1.. * : . I>.~-,.
~
u !"'I LI u
..,.t,' ~
u uu
~.:;*,t I
ELEV*l-1.
FIGURE 3.2
125 Ton Trolley 47-4 11 iJ 4" SST
-*-. I r Pit Liner
- .: ._,_:_1 Spent Fuel Cask Lifting .Lugs 4 I -4 II ,
Q ..
~'&.&.,
,, ~. 1.'., -' Upending Lugs *
, d ...... I!. .(.
Lower Cask Impact Pad Upper Cask Impact Pad CASK LOADll\G AAEA ELEV 2-2 FIGURE 3.3
r==============================================r::::::::::===~========================================;-,
l*S>ILrl-lO'ltrl t>l i
I I
11 1 ,,.,,
12-12 IPARTIALI J
ICAll* l'*i*d' WlfH hC t,Jf l~lrC( ... N:10 00. A
,lat n.-r"""4 tvW'a.cr, ,,... **d o,A. re,,<<
I*
il
'I
!1
- i*i I*
- * ~ Mvu ftC u"n """ Oo\l\.N kollll-..Cl ""'""' f ,,0111 N: f A OHi.fi ii l;' !; I* ...... j . . . . Ill&\. M>llt, I , I *lM<WI AU ~Al 1,......,.IOOI.I
- 1. C,11,\"- """°'
V"r:0: HUl*l1*0
~*&.Ll N ,......,.,c11r10 "'flt J .....u r - l UC't. "t:1111 V(C '°IUl-fUO
!Ii ii ! 1 1*
!: *, I .,..,. c*"'
~~.~.: .**
4 0-C N ~ lOCAIINO 'k.On ,tJI( TO
('(M(II lllC C,, 'k,011
- s. ..Cit",.., o, 0\ I.,,,._
- 1,'l*f.111
'1. 100 c*** 1"""""Ct ...a f'Ol,l<)I. ...,.._,..
I
,..,..-... r 0, ~~t* ('AU ...-,"Cl t-.o ,
. - ~
. ___ .... ___*u Ol f ~ I 11,000 f'O.#<OI *H'lkJC 1: :: n , 11 :;
- t6'1ff ~ I f it,MG, "'41t1Co1N Dl'lll\ IVl ..lotO "" AIO*IJJOI o,i '""""*
tra,~ND
"°°"1 ,...n ro M DI" f1* :Aou..,"'f v,,.."4.
- --**-- .._.u_g*.~----- _JJ_ UOOO \e 4 11111()0
- tu, lOAO AHf 0,111 fOllllA"'C(I VOi*
f.QCN-(\
PLAN Vflof f'U(\ l'lf l<"4* C* ,* *ll(A 1J f\l t" :Ii; OVIII t" :f lu 1<Att*f ***u* o.. .....\. * *
- o ro i 0"!l1 "'""" m t.t:r 0
~f ovUt 4',0° !f ll'(r(ll'(HCI o*u.... 1"*'*
'V'l,.f nAl " ' \.~* lOlf IMH,,..,_:,.-.
,UltOI PIA.l .....lo.fO 11441*,.M-.,.
~CLEAR SAFETY RELATED 11'. * ,OA CAT, l
,,_.,. DETAIL A. *.r ~'"';!-
. '. r*-~.
0
\o-...t* c~ . : . ' ~ , "-f.~*
DETAIL 9 IJ ~- .:.;...
\ll"'l'111tc*'1~*'J!..-.c, ,-c, l~ ....
ICAI( 1',(,(/' " **
- ~
J, .. ,..' **... * ,,
-.:.:\.)
11-11 t:*O ~~-*"'
K.Alt*f*l*O"
" Figure 3.4 FUEL CASK IMPACT PAD
10, Total Body Dose Rads/Assembly Ruptured*
1.0 0.1 0.01 0
- 0 0 1 -4+1-.-,--,--,-r_1-r1-i1--,~:--0--.----.,---,----,--,,---.-,,,'~'\O_O__----.--.-.---.---r.-,1,~rllO O0 Decay Time After Shutdown, Days.
FIGURE 5.1 - TOTAL BODY DOSE
., **- e 1000 100 Thyroid Dose, Rads/ Assembly
- Ruptured 10
. 1.0 0.1 .I 4 10 100 1000 Decay Time After Shutdown, Days FIGURE 5,2 - AD[LT THYROID DOSE
e.
100 100 More than 25% of Less than 25% of*
10 CFR 100 Limits Region 10 CFR 100 Limits Number of Assemblies Region to Yield 25% of 10 CFR 100 Limits 10 1.0 FIGURE 5. 3 -
0.2 10 CFR 100 LIMIT 10 100 1000 COMPARISON FOR CASK DROPS Decay Time After Shutdown, Days
- - e 2000 ppm Boron in Pool Water 30,000 Fuel Exposure K( 0.~95 (MWD/MT) *region 20,000 10,000 2.0 3.0 4.0 Initial Fuel Enrichment (w/o u235)
FIGURE 6.1 MINIMUM FUEL EXPOSURE VERSUS INITIAL ENRICHMENT TO PREVENT CRITICALITY IN DAMAGED RACKS
Ii
'(
APPENDIX A
.SPENT FUEL CASK DROP EVALUATION FOR THE DECONTAMINATION BUILDING AND TRUCK LOADING AREA SORRY POWER STATION UNITS l AND 2 Al.O INTRODUCTION The station spent fuelcask handling procedure is to set the cask down into the Decontamination Building during both cask receipt and shipment operations. There the cask is ~leaned and prepared for the next
- environment (fuel pool water or external ~lant environment) to which it wi11 be exposed. The Surry decontamination facility is further described in the Surry Units 1 and 2 updated FSAR, Section 9. 14.
This appendix will address the potential consequences of a postulated spent fuel cask drop in the Decontamination Building or truck loading area of Surry Un its 1 and 2. The building arrangements are shown on Figures A.1 and A.2.
27 brh/OOlOS/27 .
'I A2.0 CASK DROP ACCIDENT EVALUATION REQUIREMENTS NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" contains guidelines which should be satisfied to ensure the safe handling of heavy loads including spent fuel casks. Section 5.1.5 of NUREG-0612 contains guidelines for the safe handling of heavy loads in plant areas other than the Fuel Building or Reactor Containment which contains safe shutdown equipment. Since the Decontamination Building and truck
- lbading areas contain no safe shutdown equipment, these guidelines in NUREG-0612 do not apply.
NUREG-75/087, _"U.S. Nuclear Regulatory Commission Standard Review Plan,"
Section 15.7.5, contains guidelines for the evaluation of spent fuel
. cask drop accid~nts. Specifically, subsection I.2 requires a radiological analysis if a cask drop in excess of 30 feet can be postulated. The lowest elevation of the Decontamination Building is at elevation 6 1 10 11 and the greatest height for carrying the cask is at elevation 48 1 411
- Thus, a drop of 41 1 10 11 can be postulated in this ar_ea, requiring a radiological analysis. All other areas of the Decontamination Building floor are at elevation 23 1 011 and the truck loading area is at elevat*ion 27 1 0", such that the maximum postulated drops in these areas are less than 30 feet.
28 brh/OOJOS/28
e e A3.0 POTENTIAL TARGETS -IN THE CASK TRAVEL PATH Although no safe shutdown equipment is located in the Decontamination Building, there are two potential sources of radioactive releases beneath the cask travel path. T\'rn liquid waste tanks, the fluid waste treatment tank (DC-1K-2A) and the spent resin dewatering tank.
(1-LW-TK-10) are located at the 6 1 10 11 level. These tanks may contain a maximum of 7500 gallons of liquid waste. In addition, two resin shipping containers and nine (9) ion exchange beds are located at elevation 23' 011 in the middle bay. These items are assumed to contain a maximum of 4000 gallons of liquid waste. Both potential targets are assumed to contain degassed reactor-coolant at an activity level equivalent to that with 1% failed fuel in the core.
- A4.0 EVALUATION This cask drop evaluation will consider the following items:
(1) offsite doses (2) liquid releases.
A4. l Off site Doses Spent fuel to be shipped from Surry will have decayed a minimum of two years prior to loading into licensed shipping casks, or five years prior to loading irJto dry storage casks. The largest spent fuel casks that can be handled at Surry are dry storage casks such as the TN-2100 and 29 brh/OOlOS/29 .
(, If J 1"'*,-
the GNS-5 described in this report. Thes_e casks are typically designed to hold up to 24 unconsolidated or 48 consolidated PWR spent fuel assemblies. Accordingly, offsite doses resulting from gaseous releases from any fuel assemblies in the cask that may be damaged in the postulated cask drop accident will be well within l/4 of 10CFR 100 limits. Refer to Figure 5.3 of this report. The basic difference between the offsite dose assumptions for postulated cask drop accidents in the Fuel Building and the*Decontamination building is the absence of water to absorb iodine in the Documentation Building. Since the iodine dose rates have decayed by a factor of 5000 within 100 days of shutdown (refer to Figure 5.2), offsite doses from the postulated accident are enveloped by the results of the spent fuel cask drop analysis for the Fuel Building.
A4.2 Liquid Release When the cask is moved~rom the Fuel Building into the Decontamination Building, the cask voids will be full of fuel pool water, and the lid bolts will not be fully tensioned. Since the cask lid bolts are not fully tensioned during portions of the cask travel path over the Decontamination Building, it is assumed that the liquid inventory of the cask will leak out through the lid seal following a postulated drop in the Decontamination Building *.
30 hr
e.
The cask may imp~ct liquid waste tanks or ion exchangers during the postulated drop. The resin shipping containers will be removed from the cask travel path prior to cask handling. Reference A. 1 evaluates the consequences of liquid releases from these targets. This analysis assumes the maximu~ target release to be 7500 gallons, which is equivalent to the volume of the target tanks at the elevation 6'10". The liquid from the tanks is assumed to be degassed reactor coolant with an activity lev~l equivalent to that with 1%.failed fuel ih the core. Under the postulated cask drop, it is assumed that the cask penetrates the floor of the Decontamination Building, and that this amount of liquid is instantaneously released from the station.
The site ground water table has a continuous gradient toward the James River. A ~ell (well "D") is located at the Training Center approximately 1000 feet north of the release point. This well is about 400 feet deep and is*separated from the ground water aquifer by an impermeable layer of clay. The lined discharge canal is located approximately 150 ft. north of the Decontamination ~uilding release point. This *canal 'has an invert elevation of - 17.5 ft., which acts as a barrier to groundwater movement.
For these reasons, the ground water dispersion calculations consider that the water released from the Decontamination Building (basement elevation 6'-10") will disperse through the ground water aquifer along the shortest pathway to the river, a_distance of approximately 2000 ft.
The results of the analysis conclude that the releases at the river are less than 5% of the max_imum permissible concentrations (MPC) specified by 10CFR20.
31 brh/OOlOS/31 .
I.. <
The activitY of the fuel pool water in th_e cask is conservatively assumed to be the same as that in the liquid waste tanks. The maximum cask water inventory of the largest casks capable of being handled at Surry {dry storage casks) is approximately 2000 gallons. Since the assumed cask water contents are less than the postulated 7500 gallon release, including the cask water contents with the assumed release does not significantly change the above results. Releases at the river unqer these conditions would be less than 7% of MPC.
No fuel pellet damage is expected as a result of the postulated cask drop acc.ident. Therefore, even though some clad failure may be postulated, th.ere is no basis for assuming that any of the radioactivity in the fuel pellets is released to the water in the cask. Thus, the
. above assumptio~ for liquid releases is conservative for the postulated cask drop .accident at Surry.
A.5.0
SUMMARY
This appendix has evaluated the consequences of a cask drop accident in the Decontamination Building or the truck loading area at Surry Units 1 and 2. The analysis has demonstrated that the following criteria in NUREG~0612, Section 5.1, have been satisfied:
32 hrh/0010S../32
, \ ,,...,.
(1) Releases* of radioactive material that may result from damage to spent fuel based on calculations involving accidental dropping
- of a postulated heavy load produce doses that are well within 10CFR Part 100 limits of 300 rem thyroid, 25 rem whole body (analyses should show that doses are equal to or less than one quarter of Part 100 limits);
(2) Damage to equipment in redundant or dual safe shutdown paths, based on calculations assuming the accidental dropping of a postulated heavy load, will be limited, so as not to result in loss of required safe shutdown functions.
3~
brh/OOlOS/33
REFERENCES A.1. 1 "Radiological Consequences, Including the Ground Water Dispersion .
Effects, due to a Spent Fuel Cask Drop in the Decontamination Building, Rupturing Liquid Waste Tanks," Stone &Webster Engineering Corporation, Cale. No. UR(B)-054-0, August 16, 1982.
34 b~h/OOJOS/34
SPS UFSAR Figure 9.j4-l Loadin Area e
.".jA jA Q . -----.. ----'-*--*I°* *-*-
--' 11ll-.
PLAN-EL 47\4' C (>M*IOC)' --.:..- .
- .... -~~*
PLAN EL. 2f'o' ~ 2cl'O- . *---J . . . .,_,
C 11\ . :. ,-..: :-.<:~~
1 *"")"'-'--:---,:
r-.
- I ----**'*
. f';,c,,- -.- e
_./ ,.;.*.---*
- ~,~';~*.,*.c* **-"-'~::.r}>*/t . I til.HtRALNOTUi RU't.t:tt.>r:Ltwi!ll..,..._IP, j" . . .
- -- .-.~~b;- .*.{):~-,*'
. .'l 11\
C DECONTAMINATION FACILITY
\*LW*P*'f DC.*P.l A-A Figure A. l
-