ML20214C836
| ML20214C836 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/31/1986 |
| From: | Fraser R NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20214C820 | List: |
| References | |
| NUDOCS 8602210290 | |
| Download: ML20214C836 (31) | |
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NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 REACTOR VESSEL UPPER INTERNALS TRANSPORT AND STORAGE SAFETY EVALUATION Prepared by: Robert G. Fraser Date: January, 1986 8602210290 860213 PDR ADOCK 05000282 P PDR Page 1 of 31 L-
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SUMMARY
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To augment the replacement of the Prairie Island reactor vessel upper internals a cask has been designed to reduce the radiation dose rates and to transport the irradiated upper internals package. Removing the upper internals as a complete assembly will allow for on site storage and potential reinstallation and reuse if necessary.
The recommended thicknesses of lead for the cask are as follows:
12 cm. for the bottom, 13 cm. on the side below the upper plate, 5.1 cm. on the side above the upper plate, and 3 cm. on the top of the assembly.
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21 PPAIRIE ISLAND NUCLEAR GENERATING PLANT UNITS 1 AND 2 g
II REACTOR VESSEL UPPER INTERNALS TRANSPORT AND
,; STORAGE SAFETY EVALUATION
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- 1. INTRODUCTION II. REACTOR VESSEL UPPER INTERNALS CASK DESIGN
- 1. Shielding Evaluation
- 2. Structural Considerations
- 3. Cask Fabrication III. CASK HANDLING AND TRANSPORT
- 1. Lifting Equipment Evaluati6n
- 2. Upper Internals Cask Drop Evaluation IV. CASK STORAGE
- 1. Seismic / Structural Evaluation
- 2. Fire Protection / Detection Evaluation
- 3. Wind / Tornado Evaluation
- 4. Flood Evaluation
- 5. Shielding Evaluation V. SAFETY EVALUATION
I. INTRODUCTION n!'
T Replacing the Prairie Island reactor vessel upper internals is being performed to eliminate the potential problems related to the control rod guide tube split pin and flexure failures. Replacing the complete upper internals assembly as opposed to replacing just the vulnerable components will significantly reduce the unit down time. This action also is consistent with NSP's ALARA policy since exposure to plant personnel will be less than other alternatives.
This safety. evaluation addresses aspects of removing the irradiated internals from the containment and transporting them to the Low Level Radwaste Enclosure for storage.
Events which could potentially degrade the cask structure are also evaluated.
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II. REACTOR VESSEL UPPER INTERNALS CASK DESIGN
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- 1. Shielding Evaluation Dose rates exhibited by the Prairie Island irradiated reactor vessel upper internals make it prohibitive to remove the internals from the reactor containment building without shielding. Different shielding materials were considered and lead was chosen as being the most practical. Design constraints require the contact dose rate at the exterior of the lead shielding be less than 1000 millirems per hour. Furthermore, the design constraints require that the amount of shielding be optimized to keep the weight of the shielded container manageable and allow use of the existing reactor containment building crane.
The radiological source was defined from a survey performed on the Prairie Island Unit 1 upper internals package and from industry data on typical control rod guide tube isotopic gamma activity distribution. Beta activities were not considered in the analysis, since they are of short range and would be absorbed by any shield which attenuates the gamma-rays.
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s I A dose rate of 3200 rem per hour was measured beneath theupperinternalsatthecenter,oftheuppeb' core plate. The source term for the' upper core plate was established by assuming the gamma-ray isotopic distribution for the plate is the same as the distribution of the lower portion of the control rod guide tube and the support pins. The upper core plate was then conservatively modeled as a slab with a uniform internal source to reproduce the measured dose rate at the bottom of the plate. As a result of the large difference in dose rates above and below the upper core plate, the entire dose at the bottom of the plate was attributed to the plate itself, with no credit given to the control rod guide tubes or the support columns above the plate.
The source terms associated with the 33 control rod guide tubes and the 50 support columns were modeled as six homogenized uniform cylinders each 4 incNes high stac,ked one on top of the other above the core plate.
The-activity for each isotopic energy for each section above the core plate was multiplied by the percent of the disintegrations associated with that energy and the total number of sources and then divided by the volume s
[ of the assumed cylindrical source to establish a uniform volumetric source. Anatten,uationcoeffkcient
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for the cylinders was determined by calculating the volume fraction of steel within the cylinders and multiplying the attenuation coefficient of iron by this volume fraction.
Shielding formulae for cylinder, disk and line sources, with appropriate buildup factors, were utilized in the analysis. Due to the complexity of the geometry and the multiple energies involved, a computer program was developed and applied to the problem.
The results for the various cases studied are shown in Table I. The table includes a description of the geometry used.
Bottom Thickness. - The first two cases are for the thickness of lead required on the bottom of the cask below the lower plate. The model used was a disk source for the lower plate with 6 cylindrical sources on top of the plate. The cylindrical sources were not assumed to attenuate each other, nor was the attenuation of the plate considered.
Thus, for the cylindrical sources above the first, a void was assumed between the bottom of the cylinder and the shield.
The results of these calculations indicated a thickness
- of lead of 11.09 cm. required in the center at the bottom, and 11.56 cm. at a position 10 cm. beyond the radius of the
'. lower plate. No side shield was assumed for the latter calculation. The thickness variation is most likely a result of the different methods used to calculate on-axis and off-axis disk sources, and in particular the interpolation and extrapolation required for the latter. The results are within 5% of each other, which is considered within the accuracy of the techniques used in this analysis.
As the computer results indicate, the bottom plate or disk dominates the effect of the control rod guide tubes and support columns between the upper and lower plates.
Top - Results for the top of the shield are shown in the next case studied. For this case the attenuation cf the upper plate played an important role in reducing the lead thickness required. The dose equivalent rate
~ at the top of the upper support plate at its center was determined and found to be 25,082 mrem /hr.
The upper plate, therefore, cannot act as a top shield by itself. A th ?ttaess of 2.8 cm. was calculated as beirg 'v.ce ary at the center of the top shield.
Side - Some side shield results are shown in the next three cases studied. The basic model of six cylinders to simulate the control rod guide tubes and support
' columns were used for cases 5, 6, and 7. Case 5 looked at a position 5.08 cm. above the disk (in reality, the middle of the lower plate). The calculation indicated a lead thickness of 12.8 cm. required on the side at this location. Similar evaluations were made at 40.64 cm. above the disk, i.e., at the top of the fourth cylinder, which resulted in a thickness of 13.08 cm.,
and at 152.4 cm. above the disk, which resulted in a lead thickness of 12.9 cm. Thus, all side shield results below the upper plate consistently indicated a thickness of about 13 cm. being required. These results indicate that the side shield should be thicker than the bottom shield which is directly below the disk. This is attributed to the mean free paths through the shield, which for the side shield, are smaller for a larger fraction of the disk.
The side shield calculations for the disk source were accomplished with an off-axis disk configuration, which assumes the shield is parallel to the disk and not perpendicular to it. It was assumed that the shield thickness for the side shield would be accurately obtained from this configuration. To check m s -
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on the effect of parallel vs. perpendicular shield, the disk was modeled as a small line source at its center, 1 inches high, i.e., the height of the lower plate, with the formulae for a line source
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with a side shield. The results are shown as cases 8 and 9, which yield thicknesses similar to the off-axis disk results.
A final calculation was made for the side thickness required above the upper support plate, with the attenuation of the upper plate included in the calculation. This is shown as case 10 in Table 1, where a thickness of 5.1 cm. is required on the side above the upper plate.
A summary of the results obtained is given in Figure 1.
Several conservative assumptions were included in the analysis. These are:
- 1) The source term for the plate was assumed to be uniform, with no reduction toward the edges, which in reality most certainly occurs.
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- 2) No credit was taken for the attenuation provided by intervening materials, except for the upper plate.
- 3) The lower plate, which is by far the most prominent element in the assembly, was modeled as a disk, so no credit was taken for the self attenuation within th'e plate itself.
- 4) The formulae utilized in the analysis provide conservative results.
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TABLE I GEOMETRY DESCRIPTION LEAD THICKNESS (CM.)
CASE NO.
OR DOSE EQUIVALENT Bottom, disk plus 6 cylinders 11.09 1
on-axis Bottom, disk plus 6 cylinders 11.56 2
147 cm. off axis 3 Top, disk plus 6 cylinders 2.8 on-axis, top plate attenuation included Top of upper plate at center 25,082 mrem /hr 4
5 Side, disk plus 6 cylinders 12.8 5.08 cm above disk 6 Side, disk plus 6 cylinders 13.08 40.64 cm. above disk 7 Side, disk plus 6 cylinders 12.9 152.4 cm. above disk 8 Side, line plus 6 cylinders 12.97 10.16 cm. above bottom of line source 9 Side, line plus 6 cylinders 12.8 24.05 cm above bottom of line source 10 Side, disk plus 6 cylinders 5.1 above upper plate
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. SHIELDING CALCULATION RESULT 5
- 2. Structural Considerations The reactor vessel upper internals cask is constructed as a cylindrical, double-walled steel vessel with lead shielding sandwiched between the steel. Provisions are available for moving the cask using three lift locations 120 degrees apart. A removeable, gasketed, bolted cover with a center lift point is utilized to gain access to the cask interior. The upper support plate of the reactor vessel upper internals rests on the top ledge of the lower cask, supporting the weight of the internals. See Figure 2.
The cask will be moved within containment by the polar crane. The overall dimensions of the cask are limited by the 21 foot diameter containment equipment hatch.
The irradiated upper internals will be placed into the cask in the water filled refueling cavity. After the reactor vessel head is reinstalled the cask top will be bolted in place. The cask will be drained and then lifted by the polar crane to the containment hatch. From this point, it will be transported to the Low Level Radwaste Enclosure and will be stored as a spare part.
The internals cask design criteria were established to give reasonable assurance that the transport and storage of the irradiated upper internals assembly
can be performed in a safe manner. The following conditions / load combinations were evaluated:
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Condition Loads 1 Lifting the reactor upper internals dry weight, cask dead weight and one foot of water in the cask from the three
. lift points.
2 Lifting the reactor upper internals dry weight, cask dead weight and one foot of water in the cask from the four jacking points.
3 Lifting the cask top enclosure using the single lifting lug.
4 Cask seated on storage slab; reactor upper internals weight plus cask dead weight plus 2 psi internal pressure.
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5 Cask tipping on storage slab due to horizontal earthquake; reactor upper internals weight plus-cask dead weight with seismic acceleration.
The cask was designed in accordance with the requirements of ANSI N14.6-1978 "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials" and AISC " Specification for Design, Fabrication and Erection of Structural Steel for Buildings," November 1978 Edition as applicable.
The Prairie Island Response Acceleration Spectra were used for the seismic acceleration input. To further assure adequate margin of safety for lifting the cask the calculated load on any lifting lug was increased by 10 percent to account for uneven distribution of load. A cask drop accident analysis is not necessary. The load path to be used assures that a dropped cask would not adversely affect any safety related system when it is required to be operable. Movement of the' cask in and out of containment will be performed when the reactor vessel head is installed. Furthermore, it is not possible to transport the cask over top of more than one train of required safety related systems at any time. See Figure 3 for the' prescribed safe load path.
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r The allowable stresses in any load-bearing member of the cask, including plates, built up members, welds, etc.
shall be limited to the lesser of the following two criteria:
- 1. One third (1/3) of the material's minimum yield strength (Fy/3).
- 2. One fifth (1/5) of the material's minimum ultimate, tensile strength (Fu/5).
The above criteria shall apply regardless of the type of stress or combined stress calculated; tension, compression, bending, shear or bearing.
Non-load bearing members shall be designed using the allowable stress as given in the AISC Code.
- 3. Cask Fabrication All steel plates and structural shapes are accompanied by a certified mill test report documented to the appropriate ASTM specification. Carbon steel plates one inch thick or greater have Charpy Impact Test data at minus 40 degrees fahrenheit per ASTM A370. Furthermore, carbon steel plate one inch thick or greater has-been ultrasonically inspected in accordance with ASME Section V, with acceptance criteria per ASME Section III Class NB.
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All welding was performed by welders and procedures qualified to the ASME Boiler and Pressure Vessel Code Section IX.
All welds were subject to a final visual examination by personnel qualified to ASNT-TC-lA. Load bearing welds were additionally examined using the magnetic
. particle examination method.
Due to the magnitude of the load, a test of the cask at 150 percent of the maximum load is not feasible. The cask will be tested at 100 percent of the load and the load bearing welds will be reinspected visually prior to use on the irradiated upper internals.
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FIGURE 2 TOPHAT LIFTING LUG
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III. CASK HANDLING AND TRANSPORT
- 1. Lifting Equipment Evaluation A stress analysis has been performed on all special lif ting devices for handling the cask with irradiated internals in it. These devices, the reactor head lifting rig for inside containment and a three point spreader for outside containment, have stress levels which do not exceed ANSI A14.6 - 1978 "American National Standard for Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds or More for Nuclear Materials" limits.
Slings, used to mate the spreaders with the cask, meet the requirements of ANS1 B30.9-1971 " Slings" and have a factor of safety of five on their rated load. The load applied to the slings is well below the rated load of the slings.
All movements of the cask will be performed using plant heavy load control procedures which includes lifting equipment inspection, safe load paths, safe operating pr'actices and crane operator training.
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- 2. Upper Internals Cask Drop Evaluation An analysis has been performed to determine the radiation dose rate at the site boundary caused from a hypothetical dropping of the upper internals cask. The following assumptions were made for this evaluation:
- 1. The cask housing the reactor vessel upper internals assembly is totally damaged and no shielding credit is taken for the cask.
- 2. After the cask drop accident, the entire assembly, without the cask, is sitting outside the containment building for up to two days.
- 3. Only direct shine from the sources to the site boundary are considered. Airborne radiation contributions were negligible and, therefore, were not included.
- 4. The radiation source term data is consistent with the source term data used for the cask shielding evaluation.
Using the QAD-P5 1 Computer Code the off site dose rate was determined to be 4.9 x 10-4 mrem / hour which equates to a two day total integrated dose of 0.024 mrem. This total integrated dose is well within the 25 Rem whole body limit set forth in 10CFR 100. .
1 The QAD-P5 code is an expansion of QAD-IV which incorporates a technique for interpolating the results of neutron calculations by the moments method solution to the Boltzmann equation, additional source description routines, and an increase of the options on output. Interpolated moments-method neutron fluxes, energy depositions and dose rates may be calculated. The source of the computer code is the Radiation Shielding Information Center, Oak Ridge, Tennessee.
IV CASK STORAGE IN LOW LEVEL RADWASTE ENCLOSURE
- 1. Seismic Evaluation The exterior walls of the Low Level Radwaste Enclosure are 24 inch thick reinforced concrete designed to meet the Operational Basis Earthquake seismic criteria. The foundation supporting the two casks meets the same requirements as the enclosure by using the Operational Basis Earthquake seismic criteria. Both high and low water levels are considered in the loading conditions for the foundation. The foundation structure is designed so that a safety factor of three is met under the seismic condition.
The casks have been analyzed in their storage condition for overturning during a seismic event. It has been determined that the casks will not overturn during a seismic event with a factor of safety against overturning of 2.76. A seismic evaluation of the cask is deemed unnecessary because the loading conditions to lift the cask are more severe than the seismic condition loads.
- 2. Fire Protection / Detection Evaluation The fire protection system for the Low Level Radwaste Enclosure ties into the existing plant system and meets the plant design requirements. The fire suppression water supply is from the plant sprinkler header via a manually actuated deluge valve. This system also meets the NFPA standards.
The Low Level Radwaste Enclosure fire detection system consists of 5 flame detectors and one heat detector each of which will illuminate a common annunciator in the control room. The detection circuitry is arranged as a Class A circuit which is electrically supervised. A break in the circuit will be indicated by a " trouble" light in the control room. This detection circuitry is powered by a plant safeguard power supply.
The fire loading within the Low Level Radwaste Enclosure will be unchanged because the upper internals provide no additional combustible source to those indicated in the plant fire hazard analysis.
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- 3. Wind / Tornado Evaluation The Low Level Radwaste Enclosure.has been structurally designed to withstand wind forces in accordance with ANSI Standard A58.1 using the 100 year study period. The structural integrity of the casks and the structural integrity of the building, together with the designed margin of safety, provide reasonable assurance that tornados or tornado generated missiles will not significantly affect the encapsuling of the stored upper internals assemblies.
- 4. Flood Evaluation The reactor vessel upper internals casks, because of their construction and weight will resist hydrostatic forces as well as other effects associated with the maximum probable flood. The peak water elevation for the maximum probable flood is 703.6 feet (msl, 1929 adj). Since the cask is inside a closed structure, wave action need not be considered. The leak tight gasketed joint of the cask is at elevation 701.9 feet (msl, 1929 adj). The radiological hazard from potential flood water inleakage and the resulting out leakage of contaminated fluid is minimal due to the compressibility characteristics of the rubber like gasket.
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Long range advisory projects and three-day forecasts of river stage and crest are supplied by the National Weather Service for gage stations on the Mississippi River and its major tributaries. Gage stations covered included Hastings and Red Wing, up river and down river, respectively, from the plant site. Advisory and forecast reports are released over the Minneapolis, St. Paul local weather teletype circuit and received directly at a teletype station in the NSP system dispatch center. The basis to readily interpret these projects for specific locations between gage stations is developed in cooperation with the Corps of Engineers.
NSP has had extensive experience in using these advisories and forecasts to develop and use flood procedures for successful continued operation of many river-site plants subject to seasonal floods of varying
- - degree.
- Advance planning and preliminary arrangements for operation during floods would be used on the advisory reports of flood potential. Implementation of flood procedures would be based on the three-day forecasts of
< flood stage and actual flood stage at the plant site.
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- 5. Shielding Evaluation The reactor vessel internals casks will be stored as spare parts within the Low Level Radwaste Enclosure.
Entrance to this normally locked area is controlled through the plant access control point.
i For storage, additional shielding will be provided as needed to reduce radiation levels to acceptable values.
Areas directly outside the Low Level Radwaste Enclosure will be shielded to less than 0.5 mrem / hour. All handling, storage and monitoring will be performed using the plants existing Radiation Protection and ALARA programs.
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V. SAFETY EVALUATION The reactor vessel upper internals cask has been conservatively designed to reduce the dose rate of the upper
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internals to a manageable level for transport and storage.
The structural integrity of the cask has been designed to be within all applicable standards for both the seismic condition and the cask lifting condition. A heavy loads evaluation has been performed for movement of the cask within containment. There is no potential for damaging equipment required to maintain the plant in the safe shutdown condition. The lifting equipment for the cask has been demonstrated to have an adequate margin of safety for lifting the load. Assuming a cask drop were to occur outside of containment the two day total integrated dose at the site boundary is lower than the 10CFR 100 guideline by a factor of 10 6, For storage of the cask in the Low Level Radwaste Enclosure environmental parameters evaluated include; earthquakes, fires, tornados, and floods. A reasonable margin of safety exists against any environmental accident causing degradation of the cask. Shielding and exposure controls are implemented to assure that as-low-as-reasonably-achievable exposure is attained.
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l Safe replacement, transport and storage of the Prairie 1 Island reactor vessel upper internals is feasible and the replacement program:
- 1. Does not create a possibility for an accident or malfunction of a different type than evaluated previously in the USAR or subsequent commitments.
- 2. Does not increase the probability of occurence of an accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments.
- 3. Does not increase the consequences of any accident or malfunction of equipment important to safety previously analyzed in the USAR or subsequent commitments.
- 4. Does not reduce the margin of safety defined in the bases for any Technical Specification.
Prepared By: A Date: / - /3 /-ff, Reviewed By: A A G/) '
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Prairie Island Operations Committee Review Date: /-//- f6 Approved By: ,
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REFERENCES
- 1) C. M. Lederer, J. M. Hollander, and I. Perlman, Table of the Isotopes, Sixth Edition, John Wiley and Sons, Inc., 1967.
- 2) A. Foderaro, The Photon Shielding Manual, Second Edition, The Penn State Bookstore, 1978.
- 3) J. R. Lamarsh, Introduction to Nuclear Engineering, Addison Wesley, 1975.
- 4) T. Rockwell, Editor, Reactor Shielding Design Manual, D. Van Nostrand, Inc., 1950.
- 5) P. K. Stevens and D. K. Trubey, Weapons Radiation Shielding Handbook, Chapter 3, Defense Nuclear Agency, DNA-1892-3, Rev. 1, 1972.
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