ML20030A339
| ML20030A339 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 11/14/1961 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090340 | |
| Download: ML20030A339 (26) | |
Text
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V) t SECTION 4 Rev 1 (3/19/62)
REACTOR 4.1 REACTOR VESSEL 4.1.I General Description 4.1.1.1 The reactor vessel is designed, fabricated and tested in accordance with the requirements of the ASME Boiler and Pressure Vessel Code and applicable code case rulings.
Where the Code is not applicable, the design is evaluated from Navy Bureau of Ship Publication " Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components," 1 April 1958.
4.1.1.2 The general arrangement of the vessel end internal structure is shown on Drawing 197E853 and schematically illustrated in Figure 4.1.
The vessel is cylindrically shaped with full hemispherical top and bottom heads. The following are some gener al characteristics:
' TABLE 4.1 REACTOR VESSEL DESCRIPTION Length, overall 30 ft Inside diameter 106 in.
Wall thickness (excluding clad) 5-1/4 in.
Cladding thickness (min) 5/32 in.
Removable head wt 28 tons Total vessel weight, including head but excluding H O or internals 120 tons 2
Design pressure 1715 psia Design temperature 650 F Maximum operating pressure 1485 psig Operating temperature 593 F 4.1.1.3 The following are the penetrations in the reactor vessel:
TABLE 4.2 REACTOR VESSEL PENETRATIONS Number Diameter, Inches Location Coolant Water Inlets 2
20 Bottom Steam-Water Mixture Outlets 6
14 Shell Outlet to Shutdovm Heat Exchanger 1
6 Shell Access Ports in Top Head 3
10 Head Control Rod Drive Fenetrations 32 4
Bottom Liquid Poison Inlet 1
3 Bottom Emergency Core Spray Inlet 1
3 Shell Vessel Vents 2
3 Head In-Core Flux Monitor Penetrations 8
2 Bottom Instrument Nozzles 4
3 Shell Seal Leak Monitor 1
1/2 Head Flange ALDLO4 0%D
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~ REACTOR VESSEL SCHEMATIC
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.Secti::n 4 Page 3 3
4.1.1. 4 The closure head is bolted in place with 42 4-3/4" diameter s tud s.
The studs are stressed to tensions with a maximum permissible range of 10% maximum to minimum variation of each other by using hydraulic stud tensioners which eliminate all torsional stress in the studs.. The stress level in the studs is below the allowable 1/3 of yield stress at oper-
.ating temperature.. The seal used is a self-energizing "O" ring backed up by a duplicate "O" ring. A leak off between seals is used for detection of inner seal leaks.
4.1. 2 Materials 4.1.21 The vessel shell and heads are constructed of rnanganese molybdenum steel meeting ASTM SA-302 Grade B.
The head flange and vessel flange are of forging material meeting ASTM SA-336, with additional requirements of Code case 1236 which defines specific chemical and heat.
treatment requirements. The various nozzle forgigs are also constructed of this same material.
4.1. 2. 2 The vessel is clad using 308 and 309 type s,tainless-steel welding rod which, when deposited in a minimum thickness of 5/32", has a maximam carbon content not exceeding. 08%.
4.1. 3. 3 Stainless steel type 304 meeting various ASTM specifications, depending on whether it is plate, pipe, tubing or forgings, is used for flange covers, nozzle extensions, drive housings and other parts coming in contact with the reactor coolant.
~4.1.2.4 The closure head studs meet the requirements of ASTM-A193 with chemical composition meeting AISI 4340. These have a minimum yield strength of 120,000 psi. The nuts meet
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ASTM SA-194 with chemical composition meeting Alst 4340.
The spherical washers are of the same material.
J 4.1.2.5 The flange face's in the area of the "O" ring seating surfaces are weld clad.with Inconel A to provide a slightly harder surface than can be obtained with type 304 staaless steel.
4.1.3 Internal Supports i
4.1.3.1 Four sets of brackets are welded to the vessel internal wall to provide supports for the reactor internals. These are core supports, thermal shield supports, baffle supports, and inlet diffuser brackets. The core support brackets are located on the bottom head and consist of four plates welded to the base metal and are constructed of stainless steel type 304. These four brackets align and support the core support plate.
J 4.1. 3. 2
' The six thermal shield brackets are located below the active fuel section of the core. They are welded to the vessel base metal and are made of 304 stainless steel. The brackets support and align the vessel thermal shield.
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Secd:n 4 Page 4 4.1. 3. 3 Just above the 14" steam-water mixture outlets are eight brackets welded directly to the vessel cladding. These are to support the steam baffle and the emergency core spray sparger.
4.1. 3. 4 The only other attachments on the vessel interior are brackets which hold the inlet diffusers in place in the bottom head.
These als.) are welded directly to the vessel cladding.
4.1, 4 External Vesse1 Supports The vessel is supported by 12 brackets attached to the ex-terior vessel shell. Twenty-four hanger rods attached to these brackets will transmit the vessel weight to supports anchored to the building foundation structure. Individual supports will carry two or four hanger rods. There are eight stabilizing brackets attached to the exterior vessel shell. Four are located near the vessel bottom head, and four are located near the vessel support brackets. The stabilizing brackets allow expansion of the vessel, both in the radial and longitudinal direction, but prevent movement i
of the vessel central axis. These vessel supports are designed to adequately withstand the horizontal loadings on the reactor vessel from the reactive forces as may be developed with the rupture of any single line to the ves;el.
4.1.5 Radiation and Stress Effects 4.1,5.1 The maximum fast neutron flux, including neutron mergies of v.1 hiev and over, for 40 years' full power operation 19 (100% load factor, 240 hiwt) is calculated to be 3. 3 x 10 neutrons per square centimeter at the inside surface of the vessel wall opposite the core inidplanc. The maximum integrated neutron flux on the same basis, but including only neutron energies of I hiev and over is calculated to be 1. 3 x 10 9 neutrons per square centimeter. All reactor vessel 1
no zles anel penetrations are in areas where the integrated dose is well below those indicated.
The effect of this integrated neutt on flux on the mechanical properties of the reactor vessel may be predicted on the {)esults reported in a cummary report sponsored by the AEC(
as follows:
(1)
Rich ard, Wm. Kelleman, Alco Products, Inc., Schenectady, N. Y. -
"A S 2rvey of the Effects of Neutron Irradiation c*n the Impact and Other hiecl anical Properties of Pressure Vessel Steels for the Shi-2 Reactor,"
i AEC Research and Development Report UC-81, uAE No. 61, dated,
- April 1, 1960.
Section 4 Page5 REFERENCE REPORT DATAkII 5 x 1019 nyt Before 1 mev Tensile Strength, psi 80,000 123,000 Yield Strength, psi 50,000 122,500 Uniform Elongation, %
20%
10 %
Increase in Transition Temp.
+300*F
% Decrease in Maximum Absorbed Energy 50%
4.1. 5. 2 The material properties are expected to tend in the direction indicated in the above reference information. However, it is also expected that changes in material properties will be different from those implied above by the simple summation of neutrons, as a result of different irradiation histories.
It is believed that there will be large differences in material property changes among those from: (a) short " fast" irra-diations compared with long " slow" irradiations; (b) full fission spectra compared with partially shielded spectra; (c) stressed _during irradiation opposed to essentially stress-free during irYadiation; (d) properties during irradiation opposed to properties after long decay periods after irra,.
diation; (e) long time at temperature during operation opposed to short times at temperature and on the absolun value of temperature; and (f) material identified simply as ASTM, SAE, AISI or other symbols opposed to that material that has a definitely known fabrication history.
4.1.5.3 The. "as-fabricated" reactor vessel material has the follow-ing n.echanical prope ties:
Yield _Stre 4gth, psi 56,900 Tensile S.rength, psi 84,200 Elongation in 2", %
23 The nil ductility transition (NDT) temperature of this material is less than 10*F with an average impact value of 63 ft-lbs at 10
- F.
The effect of neutron irradiation on this NDT tem-perature is calculated to be about 150*F increase with an integrated dose of 1. 3 x 1019 (1 Mev and over) neutrons per square cengmeter and about 200*F with an integrated dose of 3. 3 x 10 (0.1 Mev and over) neutrons per square centi-me te r.
4.1. 5. 4 The changa in mechanical properties (indicated by reference report data) are based on exposures at temperatures of about 200*F and for full fission spectra at very high neutron fluxes (normally of the order of 1013 neutrons /sq cm/sec). It is felt that the totai actual changes of the Big Rock Point reactor vessel.nechanical properties will be generally less than those preuteted by this reference data, due to exposure at higher temperature and lower incident flux, and due to the self-annealing effect of operatton at about 560*F metal temperatures.
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Op. Cit., Richard, Wm. Kelleman
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. Seeti n 4 Page 6 This "self-annealing" term, or annealing during operation at elevated metal temperatures above 500*F, has been used by ORNL Solid State Division to describe and hypoth-esize a delaying of changing of center physical properties due to irradiation at elevated temperatures; these property changes taking place slower at elevated temperatures but tending to reach what appears to be a saturated condition in a longer period of time compared to rates of change at lower temperatures below 500*F (From Annual Progress Reports of Solid State Division ORNL, over the last several years).
4.1. 5. 5 At Big Rock Point reactor's design pressure and temperature, the pressure produced reactor vessel membrane stress is about 38% of the unirradiated yield point stress. Calculations indicate that the temperature difference acros:: the reactor vessel wall at the. core midplane at 240 Mwt power will be abut 14 *F.
This temperature distribution would produce tangential and longitudinal stresses on the outside face of a'aout 2600 psi tension. The ASME Code allowable steady-state thermal stress is 10,000 psi. Combining this thermal stress with the pressure produced membranc stress at design pressure produces a stress of only about 44% of the unirradiated yield point stress at design temperature.
4.1. 5. 6 During startup, the transient thermal stresses are basically compression on the inner face and tension on the outer face, and tend to be neutralized by the gamma heating stresses.
During shutdown, the transient thermal stresses arc reversed, but the gamma heating is 6% or less of the gamma heating at normal full power so the additive stress combination is ab-sent.
4.1. 5. 7 To minimize the thermal stresses in the reactor vessel walls, the rate of change of power is adjusted during startup or shut-down to maintain a temperature rate change not in excess of 100*F per hour in the reactor vessel walls.
4,1. 5. 8
'Vith due regard to the above indications of a lack of complete, precise knowledge of the effect of neutron irradiation on the mechanical properties of the vessel, it is believed that the Big Rock Point reactor vessel is adequately designed against the possibility of brittle failure during its 40 year operation.
4.1. 6 Reactor Ineulation The reactor insulation is an all metallic insulation three inches thick. It is attached to the vessel by banding and is supported by small brackets welded to the vessel prior to final stress relief. The complete outside surface of the vessel is covered except for the drive housings located on the bottom head.
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- 4. 2 REACTOR CORE AND FUEL
Secti:n 4 Page 7 0
- 4. 2.1 Fuel and Core Description
- 4. 2.1.1 Dimensions and arrangement of the core are schematically shown in Figure 4. 2; the fuel bundle design is illustrated by Figure 4.3 and described in Drawing 141F852. Each fuel Endle is made of 144 fuel rods, which are stainless-steel tubes containing stacks of cylindrical UO2 Pellets in a 12 x 12 array with the necessary handle, base, intermediate wire spacers, fasteners and other hardware.
- 4. 2.1. 2 In the fuel bundle there are 132 standard fuel rods and 12 special rods; three in each corner, with smaller diameter and thicker jackets. The 12 special rods are used to mini-mize the local heat generation hot spot factor. The four corner tods are segmented in order to mechanically support the three wire spacers. The dimensions of the fuel pellets and jackets are summarized in the data sheet in Section 8.
All rods but the four corner rods contain 70 inches active fuel length, plus a plenum located at the top of 2-3/4 inches length. A coil spring and wafer in each plenum insures that the stack of UO2 Pellets is properly positioned within the fuel rod. The fuel rods are sealed on each end by welded end plugs.
The segmented rods have correspondingly shorter active fuel and plenum lengths.
- 4. 2.1. 3 The fuel rods are positioned at the top and bottom of the fuel bundle by 1/8-inch thick drilled tie plates, which are supported in turn by a cast handle and base, at top and bottom respect-ively. The end plugs fit into the holes of these members such that improper assembly is not possible. A total of 32 of the 144 fuel rods have threaded end plugs which thread into the base and are attached to the handle with nuts, which are safety-wired. The other 112 fuel rods have unthreaded end plugs which fit freely into thentie plates. All 144 fuel rods are held securely but with provisions for slight relative expansions by coiled springs at the upper end plugs.
- 4. 2.1. 4 Three wire spacers guide the fuel rods at intermediate levels of the fuel bundle. These spacers are formed of a double layer of crimped wire grids and a surrounding sheet band dimpled at the points of contact with the fuel rods. The spacers are positioned axially by the four segmented corner rods which fit through sleeves in the corners of the spacers.
- 4. 2.1. 5 Fuel channel design is shown in Drawing 141F759. The first (157 Mwt) core will contain approximately 46 stainless-steel channels, and the remainder of the 56 fuel bundles will have zircaloy channels. The number of stainless-steel channels loaded into the first core will be adjusted to assure that shut-down control requirements are met. The actual number will be established during initial criticality measurements.
- 4. 2.1. 6 The fuel channels surround each fuel bundle, thus potitioning each bundle, provide guide surfaces for the control rods and make possible flow orificing. The clearan<.e between the fuel i
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i Secd n 4 Paga 10 channels and the upper guide grid limits the between-channel flow to the required amount.
- 4. 2.1. 7 The fuel channels are individua11f attached to and supported by the support tubes by a flange-tyi.- bolted joint, and pre-vented from lifting out of the core by the weight of the channel-support-tube assembly, and by the b 3ams of the upper guide grid.
- 4. 2.1. 8 The wall thickness of the zircaloy channels is 0,100 inch, this t5.ickness being required to withstand the possible pressure differential between the inside and outside of the channels. The wall thickness of the stiffer stainleas-steel channel is 4075 inch.
- 4. 2.1. 9 Flow orifices are positioned at the connection between fuel channel and support tube, held down by hooks, and are locked in position by a quarter turn of the orifice.
The flow orifices provide flexibility for meeting various operating conditions and allow flow through channels to be varied across the core.
- 4. 2.1.10 It is planned to use two antimony beryllium neutron sources of concentric cylinder configuration located either at the core periphery or in the core interior.
4.2.1.11 T he relationship of core components is shown in elevation and plan in the Vessel and Core Arrange-ment Drawing 197E853, and schematically by Figure 4. 2.
- 4. 2.1.12 Fuel and core materials selections are summarized in the table below:
TA. P LE 4.'3 PRINCIPAL CORE MATERIA 1E Component Material Specification Remarks APED-FA231904 3.2 % 1.068% enrichment Fuel Pellets UO2 Fuel Jackets 304 SS ASTM A-269 Min. yield 75,000 psi i
End Plugs 304 SS ASTM A-276 Min. yield 75,000 psi Fuel Springs Inconel-X AMS-5698 Single Aged Handle and Base Grade CF-8 SS ASTM A-351 Tie Plates, Sparers 304 SS ASTM A-167 Spacer Wire 304 or 308 SS AISI 304 i
Zircaloy Channeh Zircaloy-2 ASTM B-352 Vacuum Annealed Steel Channel, 304 SS ASTM A-167 Annealed
- Flow Orifices 304 SS AISI 304 i
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i Section 4 Pago 11 4
4.2.2 Fuel Mechanical Design Considerations 4.2.2.1 Stress Analysis of Fuel Jackets. The tubing is cold worked Type 304 stainless steel, specified to have a yield strength at room temperature of 75,000-95,000 psi. The minimum expected yield strength at 700'F is 55,000 psi. Irradiation will furthe r increase the yield strength of the material.
4.2.2.2 The worst casas for which stress was calculated are:
a)
Beginning of life: transien' overpressure, overpower condition. Reactor pressure = 1700 pai, pressure within jackej = negligible, maximum heat flux = 450,000 Btu /
hr-ft.
b)
End of life: power generation = 125% rated power, inter-nal pressure exists due to gases originally present within jacket and evolved fission gases, external pressure =
atmo sphe ric. This could exist only in a loss of pressure accident.
4.2.2.3 Stresses due to external pressure were calculated assuming maximum O. D., maximum tube ovality, and minimum wall.
Thermal stresses, as well as mechanical stresses, were con-sidered.
- 9. 2. 2. 4 In calculating the worst intergal pressure, the fission gas pro-duction is taken as 1.35 x 10- moles per MWD, all applicable hot spot factors are applied when calculating the maximum fuel exposure, the release fraction is taken as 50% based on N.S. Savannah developmental fuel test results, and the volume avai:able for fission gas is assumed to be only the plenum volume.
4.2.2.5 Applying the maximum shear theory to each of the calculated maximum circumferential longitudinal, and radial stresses:
Maximum shear stress = 1/ 2 (SX~8Z) max = 24,100 psi Shear stress at yield point = 1/2 Sy, p, = 27,500 psi Thus, the maximum shear stress in the fuel jacket is less than that required to cause yield. In the event of sudden loading, as would occur in a loss of coolant accident, the actual stress values would not exceed those calculated; however, the yield stress would increase on the basis of information given by Timoshenko, Vol. II, 2nd Edition, 1951, page 425. Thus, the applied stress from sudden load-ing would be a smaller fraction of yield than with the case of static loading.
4.2.3 Fuel Vibration and Distortion Analysis 4.2.3.1 Lateral rod deflections are caused by (a) thermal bow, due to variation in temperature of one side of a rod with respect to the other, (b) flow induced vibration, (c) manufacturic.g tolerances, (d) axial loading from the coil springs at the roc ends.
Ssetion 4 Page 12 Rev 1 (3/19/62) 4.2.3.2 Vibration amplitude is calculated based on experiments of the fuel Research and Development (R&D) program which utili7ed an instrumentcd pr ototype of the subject reactor initial core fuel bundle with actual size fuel rods in flowing two-phase water and steam at reactor conditions of temperature and pressure; these calculations are consistent with reported experiments of Burgreen, et al. Transactions ASME, July,1958.
4.2.3.3 The most significant distortions are actually the original manufacturing var iations resulting from clearances, tolerances,
rod bow, etc. Distortions due to thermal bow and axial loading are calculated based on conservative assumptions of heat generation distribution and loading.
4.2.3.4 The resultant minimum inter-rod clearance is.060 inch, and the minimum fuel rod to channel clearance is.050 inch from the distortions due to thermal bow and axial loading.
4.2.4 Experimental Verification of Mechanical Design 4.2.4.1 Irradiation tests of several Dresden prototype fuel bundles, 24 R&D fuel bundle s, and 25 Fuel Cycle Development Progr am fuel bundles in VBWR with wire spacers and with jacket stresses comparable to or greater than Big Rock Point fuel stres ces are being carried out. It radiation of Dresden Reload fuel will give further experience with these same design features. The R&D test bundles have been irradiated to an average exposure of about 2500 MWD /T as of October 30, 1961, without indication of f ailur e.
4.2.4.2 Experience with the Big Rock Point dummy fuel bundle, a full size lead shot filled prototype used for handling tests, has proven the construction, clearances, and tolerances of the Big Rock Point fuel suitable for all required refueling opera-tion s.
4.2.5 Heat Trfusfer Design Consideration of a Fuel Rod 4.2.5.1 The crud buildup is not considered to be a problem in the heat transfer design consideration of a fuel rod. Experience, to date, has shown that a 6.... oxide film develops on the rod, but that scaling or fouling due to the buildup of crud does not ap-pear. Stainless steel clad fuel rods have been examined at two other boiling water reactors. The crud thickness on these rods is such that it cannot be measured by pre-and post-irradiation diameter measurements, and in some cases, the stainless steel metallic luster remains after irradiation. In the temperature gradient across the fuel rod, an ample margin is provided for the drop through the cladding. This margin represents only a very small percentage of the total allowable drop. Therefore, even if the drop through the cladding should increase by two or three times, it would have very little effect on the total temperature drop across the fuel rod.
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f Ssction 4 Page 12A Rev 1 (3/23/62) 4.'2.6 Vessel Internals i
4.2.6.1 Three major assemblics - the core support, thermal shield and top guide - provide support and alignment of fuel and guidance of control rods. These and other vessel internals are shown in Figure 4.3A (Page 12B).
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4.2.6.2 The core support, a circular plate 1-1/2" thick, rests near the bottom of the reactor vessel on four support pads which are welded to the vessel. It has four alignment brackets which engage the four support pads to accurately position the plate relative to the control rod drive penetrations in the vessel bottom head. A minimum radial cicarance of 1/8" exists where each control rod drive nozzle penetrates the plate.
Eighty-eight support tube adapters, bolted to the core sup-port plate, accurately align the base of the support-tube-and-channel assemblies relative to the control rod penetra-tion s. Large, evenly spaced holes through the plate permit coolant flow from beneath.
4.2.6.3 Six equally spaced pads support the thermal shield in the vessel. Two threaded I" diameter studs in each pad secure the shield, and three 1-1/4" diameter pins provide and main-t min alignment. Holes near the bottom of the shield operate in conjunction with a seal ring at its bottom edge to provide the proper amount of coola'at flow in the annulus around the shield.
4.2.6.4 A continuous ring and intermit, tent brackets near the top of the thermal shield provide a mounting surface for the top guide, which is aligned with the shield by two pins and is bolted in place. The top guide provides radial sup-port and upper alignment for the support-tube-and-channel t
assemblies, and by virtue of the clearance between the channels and the top guide beams controls the leahage flow I
which bypasses the fuel bundles. When the top guide is in the assembled position, offsets on its beams prevent vertical movement of the support-tube-and-channel assemblies. The top guide beams are locked in the assembled position by beam clamp assemblies which can be released, allowing the beams
!I to be individually lifted to a vertical position in order to free the support-tube-and-channel assemblies for removal. The beam clamps are designed such that no force is transmitted to the latches.
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4.2.6.5 The support-tube-and-channel assemblies support the fuel bundles, guide the control rods, and provide a means for controlling the flow of coolant through the fuel. Ba sically,
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'Rev 1 (3/23/62) each assembly consists of a round cross-section (support) tube and a square cross-section diannel joined by a bolted flange.
For each channel, an orifice assembly is attached to the support tube at the elevation of the flange and is locked in place by i
turning. Positive protection against unlocking is provided by interference between the feet of the inserted fuel bundle and b-the orifice handle. Two spring latches on the orifice assembly also act to prevent inadvertent unlocking prior to fuel loading.
4.2.6.6 The baffle, located above the steam outlet nozzles, prevents steam slugging in the risers between the reactor vessel and the steam drum. It is composed of four segments hinged to j
a solid ring and is provided with rotating locking clamps which interlock adjoining segments. The locks are a positive type i
which may be checked manually by use of a drift. Major access to the core is obtained by unlocking the segments and rotating them upward.
4.2.6.7 The two inlet nozzles. located immediately above the core j
support, are equipped with diffusers to impart lateral and downward velocity to the coolant, thus distributing the flow and sweeping the bottom head. The coolant flows both above and below the core support and enters the bottom of the sup-port tubes through the support tube adapter. The adapter is designed to permit full flow up each channel and yet provide the support to the support tubes. The adapter as shown in Figure 4.3A is a machined casting with five support ribs connecting the inner core to the outer support ring. The space between ribs is clear fo,r coolant flow therefore most of the area of the adapter is actually flow area. The water then flows up the support tube through the orifice and then through the fuel channel.
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l-Section 4 Page 13 Rev 1 (3/19/62) 4.3 CONTROL RODS 4.3.I Control Rod Description 4.3.1.1 There are 32 control rods in the Big Rock Point reactor, which are used in both tha 157 Mwt and 240 Mwt cores. The control rod blades are 11-1/2" wide x 5/16" thick and each control rod contains 116 poison tubes. The construction of the control rods is de-scribed briefly as follows, and the parts are identified on Drawing Il4B5965.
4.3.1.2 Poison tubes are stainless-steel tubes,.175" O.D. x.020" wall, with welded end plugs and with 68" poison length of natural boron carbide powder of 707. theoretical density.
The poison tubes also contain steel balls, swaged in position at regular intervals to compartmentize the boron carbide and minimize the possible effects of densification or settling of the B C powder, 4
4.3.1.3 The poison tubes are contained in a structure con.psed of a central core and four sheaths which form the cruciform shape.
This cruciform, along with a handle, and a connector which contains the coupling to the drive, make up the control rod.
Holes of 3/4" diameter are placed in the sheaths to allow coolant to flow by the B C poison tubes.
4 4.3.1.4 The control rod is guided on the fuel channels and support tubes by rollers at each blade, top and bottom.
4.3.2 Control Rod Materials Materials used in the control r6d are summarized below:
TABLE 4.4 CONTROL ROD BLADE MATERIALS PART MATERIAL SPECIFICATION REMARKS Poison Boron Carbide APED-DP231709 707. T.D.
Tube Type 304 SS ASTM A-269 60,000 psi min.
yield stress End Plug Type 304 SS ASTM A-276 Ball Stainless Steel AISI 400 Series Sheath Type 304 SS AST!A A-167 Core Type 304 SS ASTM A-351 Handle Grade CF8 SS ASTM A-351 Connector Grade CF8 SS ASTM A-351 Latch Grade CF8 SS ASTM A-351 Roller Pin Haynes 25 APED-FA19698 Roller Ste11ite Haynes No. 3 l
't S:cti n 4 Page 14 1
4.3.3 Control Rod Stress e nd Distortion ' Analysis
- 4. 3. 3.1 The probable limit to the life of the control rod is internal 10 pressure build-up due to release of helium formed by B 7
(n, c() Li reaction.
,4.3.3.2 The pressure build-up and stress in each individual poison tube of esch control rod will depend on its integrated exposure, i
4.3.3.3 In order to give an indication of minimum life expected for any individual control rod, hoop stress in the worst tube due to internal pressure has been calculated as a function of time based on the following assumptions:
a) Internal pressure is present due to 1500 ppm volatile cen-tent in the B C (assumed tc6 he H O which subsequently 4
2 2 and,0 ), and helium whic1r
" dissociates completely. to H 2
is introduced during fabrication.
b) The control rod is inserted continuously in the highest flux region of the reactor (1.3 times average flux), being fully inserted for a fraction of each operating cycle and g
being gradually withdrawn at the end of each operating cycle. (The operating cycle is the time between reactivity additions--refueling or steel channel removal. )
i 4
c) Reactor is operating at. 8 load factor.
d) Of the He atoms formed, 30% are released from the B C 4
powder and contribute to the internal pressure within the poison tubes.
If a control rod is inserted in the highest flux region con -
i tinuously as described above, the resultant life, or time for the hoop stress in the worst tube to reach 50,000 psi (90%
of expected yield strength), is greater than 1 year.
g 4.3.3.4 The stress in the worst tube has also been calculated as a t
function of time for " normal operation. " In " normal operation" all control rods are used to control excess reactivity for burn-up and fission product poisoning such that the worst control rod captures 1. 3 times as many neutrons as the average control rod and the warst poison tube in the worst control rod captures 2.9 times as many neutrons as tFe average tube in that rod.
As-sumptions for helium release from B C, initial pressure in 4
tubes, and plant load factor are as given above. Re sultant
(
controleod life if limited by internal pressure is greater than 10. yea r s.
s 4
Sziction 4 Page 15 4.3.'3.5
-An analysis was made to de ermine whether temperature grad-
-ients could exist in the structure of the control rod sufficient to cause thermal distortions. It was calculated that even with the control rod bowed close to the fuel channel in the worst-expected tolerance condition (1/16" gap between control rod and fuel channel along their full length) there was sufficient natural circulation flow (with local boiling) to keep all surfaces of the i
control rod at essentially uniform temperature.
4.' 3. 4 Summary of Experience and Testing of Boron -
rbide Control Rods i
4.3.4.1 Four control rods containing boron carbide in stainless steel tubes were used in VBWR from May,1958 to November,1959.
j.
These control rods were always withdrawn during reactor opera-tion but received a considerable amount of gamma and neutron i
exposure. These rods operated _ completely satisfactorily and no bulging or distortion was found after use.
4.3.4.2 Eighty control rods containing boron carbide in stainless steel tubes were used in the Dresden reactor from June to October,1961.
A complete post-irradiation inspection of six representative control rods was performed in October,1961. The visual and dimensional inspection indicated that the control rods were in excellent condition. No damage or distortion of any kind was
- observed, i
4.3.4.3 Additional experience with B C control rods has been obtained 4
in VBWR, OMRE, University of Michigan Research Reactor Original B C blades for VBWR (later j
and other reactors.
4 replaced) consisted of B C plates supported by a stainless 4
3 1
steel sheath which was not water tight. In service the B C 4
was washed away by circulating water. The University of Michigan reported that the oval tubes about 0. 75" x 0. 225" x 24" long which contained the B C swelled to about 1.1" thick 4
after 2-1/ 2 year's service. Swelling was attributed to water leakage into the tube, subsequent plugging of the leak, followed by a pressure build-up, assumed to develop from 112 and O2 t
gas generated by dissociation of trapped water. No other report has been noted of unsatisfactory operation of rods made of B C in tubes.
4 4.3.4.4 The lessons learned from the service failures related above (a) the-B C must be contained within a pressure vessel are:
4 to prevent erosion or excessive distortion, and (b) the volatile content of the B C must be strictly controlled and accounted 4
for in the design.
i
.i l-I.
1 i
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S;ctirn 4 Page 16 f
4.3.4.5 Experimental determinations of release of He formed by n,oc reaction with boron-10 in B C have been reported in KAPL-4 1403, "Effect of Irradiation on Hot Pressed Boron Carbide, "
'W. D. Valovage, November 15, 1955: in BMI-1406, " Radiation Effects on Boron Containing Compounds, " D. J. Hamman and P. Schall, January 6,1960; and by E. W. Hoyt and D. L.
Zimmerman of VAL. Based on these reports, a conservative design figure of 30% He release has been selected.
Gas evolution tests of B C powder from the production line for 4.3.4.6 4
Dresden replacement control rods were conducted. Typical results were that 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after drying a 500 ppm weight loss resulted when the powder was heated to 750F, and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after drying a 1200 ppm weight loss resulted when the powder was heated to 750F. (Dresden control rod tubes were there-fore loaded with powder as soon as possible after the powder was dried. ) The gas was identified by mass spectrometry as H, which could have been fo'rmed from a water -B C 2
4 Volatile content of B C for the Big Rock Point reaction.
4 control rods is specified as 500 ppm maximum weight loss when heated to 750F, and a maximum boric anhydride content is specified to make this possible.
4.3.4.7 A series of mechanical tests were preformed on prototype B C control rod components to confirm the adequacy of the 4
structural design of the control rod. These included burst tests of the poison tubes with welded end plugs, tensile tests of spot welds between the sheath and core, and the following special tests:
a) During tube swaging the tube O. D. was ' reduced about 0. 012" and the spacer ball was found to be embedded into the tube about 0. 002" to 0. 003".
b) Ball fixity tests showed that a force of 350 to 590 lb. was required to move the ball from its indented position through the tube.
l c) Swaged tubes were burst inside a control rod sheath to determine the effect on sheath dimensions.
,4 ect was minor, only 0,012" resultant sheath bulge.
I d) Two prototype control rods, a Dresden prototype and an 111/2" wide prototype, were operated in the test facility in the scram niode of operation to test the clearances, wear, etc., between the control rod and fuel channels (as well as
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Section 4 Page 17 Rev 1 (3/19/62) testing the drive). In a premeditated test the prototype Dresden drive-control rod was given 1000 hot scrams and performed satisfactorily. The 11-1/2" blade and drive per-formed satisfactorily for 100 cold scrams and hot tests are continuing.
4.4 CONTROL ROD DRIVES 4.4.0.1 There are 32 bottom-entry, hydraulically-operated control rod drive mechanisms spaced on 10 466-inch centers, each drive actuating an individual control rod. Reactor feed-water is used as the working fluid for both normal and scram Tha drive mechanisms are operations (Drawing M-122).
mounted vertically in the thimbles welded into the reactor bottom head penetrations. The lower end of each thimble terminates in a flange for mounting the drive mechanism and attaching hydraulic lines. The drive mecha.tism is at-tached to the thimble mounting flange, using metal "O" rings to effect a seal. Unheated reactor feedwater is supplied through the pipe conne:tions in the thimble flange to hydraulically actuate the drive.
4.4.0.2 The drive consists basically of a piston operating in a hy-draulic cylinder. Differential pr essur es across the piston are used to drive the rod into or out of the core. A ratchet-locking device is used to hold the drive at set positions when driving pressur es are absent. Unlocking of the ratchet is not required during insertion of the control rod into the core, but is necessary for withdrawal of the rod. Speed of insertion or withdrawal is controlled by a flow-contr ol device in the hydraulic circuitI 4.4.1 Drive Mechanism Description 4.4.1.1 Drive Piston and Piston Rod 4
The drive piston, mounted at the lower end of the piston rod, has both inside and outside seal rings, and operates in an an-nular space between the inner cylinder and outer cylinder.
The piston rod, or index tube, is a long tube of full stroke length mounting a coupling spud at the upper end for engaging the control rod coupling. Uniformly spaced grooves (approx-l imately three inches apart) in the outside of the index tube are engaged by the collet lock. Filtered reactor water is admitted to the inside of the index tube to limit the effect of the react 6r pressure to the cross-sectional area of the index tube.
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'i S:cti::n 4.
Page 18
- 4. 4.1. 2 Collet Assembly The locking mechanism is a simple, ratchet-type device utilizing a collet with six locking fingers which engage the index tube grooves. The locking groove is shaped to open the fingers when inserting the control rod. The fingers must be held in a retracted position to withdraw the control rod. This is accomplished by hydraulically raising the collet piston, which raises the fingers behind an unlocking cam surface. The collet is normally held in its lower position (free to engage the locking groeves) by gravity and a collet return spring. The fingers trans-mit the load, resulting from control rod and index tube weight, to the shoulder of the outer cylinder wall.
- 4. 4.1. 3 Inner Cylinder Extending up the center of the piston and index tube is an inner cylinder or column attached to the bottom flange of the drive. Water is brought to and from the upper side of the drive piston through this column. A series of small orifices at the top of the column provides a progressive water shutoff, thus cushioning the piston at the upper end of the scram stroke. This column mounts a small, station-ary piston which seals off the upper side of the drive piston from reactor pressure. This riston is also used as a positive end stop, and a series of spring washers is used to cushion the final impact. Inside the column is a tube containing the position indicator probe.
- 4. 4.1. 4 Flange and Outer Cy1inder The main drive flange has an upper sealing surface for making a static seal to the thimble mounting flange. A shuttle or two-way check valve in the flange directs reactor vessel pressure (admitted through the annular space between drive and thimble) or outside control pressure to the lower side of the drive piston. This arrangement assures that the pressure below the drive piston cannot drop substantially below vessel pressure and allows reactor pressure to be used as an alternate driving force for scram operation.
^
Because the drive is normally held at a pressure slightly greater than reactor, the valve's normal position is as holding the reactor port closed. The shown in Figure 4.42 outer cylinder of the drive is welded to the lower head of the drive. Two concentric tubes provide water passages i
S;ction 4
Pags 19 FIGU RE
- 4. 4
, r 7,s eCOUPLING SPUD s
c
' i l
f l
REACTOR WATER COLLET FINGER f
COLLET PISTON AE a
v' PRESSURE INLE eee'~e PRESSURE NLET 7
,.t
\\
'/
/
i I s'I k
s
- iw al/
fm s
i s
h l
SHUTTLE VALVE
/
f::: : __ d: :
'j~ T POSITION INDIC ATOR ~
if
'/f OUTER CYLINDER THIMBLE
'j i
/
I l
f 3 If f ;
DRIVE PISTON 3 t
/ '
- I e
'/
PISTON RING INNER CYLINDER l},f) lc CONTROL ROD DRIVE MECHANISM
i Szction 4 Page 20 Rev 1 (3/23/62) between the main drive flange and the lower head. The outer,
heavy-wall tube forms the reactor pressure containment. A secondary closure is provided at the lower head of the drive for ease of manufacture, assembly, and maintenance.
- 4. 4.1. 5 Position Indicator The center tube of the drive mechanism is a well containing the position indicator probe. This area is at atmospheric pressure. The probe mounts a series of hermetically-sealed, magnetically-operated switches, each of which indi-cates a discrete rod position. The switches are operated by a permanent magnet carried by the drive piston. The inter-vening walls are of nonmagnetic material, allowing each switch to be operated as the piston passes. Extra switches are provided at each end of the stroke to indicate limits.
4.4.1.6 A thermocouple is placed at the top of the position indicator probe of each control rod drive and is held in position by spring clips. The temperature readings from these thermo-couples provide an indication of the seal operating conditions and detection of flow from reactor to scram dump tank (via leaking scram valve).
4.4.1.6. I Melting of the low temperature solder used on the position indicator switches for similar drives in other reactors led to installation of thermocouples. Monitoring of these thermo-couples has shown the primary cause of high temperatures to be leaking of outlet scram v.alves, which can be adjusted as necessary to obtain tight shutoff.
4.4.1.7 Coupling The coupling is a semi-spherical collet-type device; see Figure 4. 5.
The coupling is attached to the drive by insert-ing the drive coupling spud, a six-fingered collet, into the control rod coupling socket. The collet fingers spring out into the spherical socket and a locking plug is moved down into the collet. This plug prevents the collet fingers from cloc ing. The collet fingers must be closed far enough to clear the socket for uncoupling. When uncoupling, the locking plug is lifted from above, using suitable -handling tools, or it may be operated through the drive itself.
4.4.1.8 Coupling Integrity Check An overtravel position is provided to permit checking the i
blade-to-drive coupling. This is accomplished by providir.g a blade.stop at the lowest normal withdrawn position. If the
(
l-Section 4 Page 20A
- I, Rev 1 (3/23/62) blade is coupled, both drive and blade will stop in this position
.when fully withdrawn. If the blade is uncoupled, the blade will stop, but the drive can withdraw a small additional distance--
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energizing a position switch warning the operator.
' 4.4.2 Materials of Construction 4(
4.4.2.1 The various materials used in the drive mechanisms have been chosen on the basis of long term engineering tests, together with
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operating experience from the Dresden installation. Extensive use has been made of 300 series stainless steels where high-strength or wear resistance is not required. Other materials,
- selected for strength, wear resistance or bearing character-istics, are summarized in the table below:
-4, i
k 4
2
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i.
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f I
1 Sacdon 4 Page 21 Figure 4. 5 BACKSEAT VALVE a
t p CONTROL' ROD T*
e El *l l*
I SPRINGS o,
,e
- r 7*l COUPLING SOCKET c
/
LOCK PLUG OOUPLING SPUD i
,* 5 7 <>
v~
l CONTROL ROD DRIVE il P00R ORIGINAL W
CONTROL ROD COUPL ING
t S2ction 4 Page 22.
a-TABLE 4. 5 MATERIALS OF CONSTRUCTION Sliding seals and bushings (high velocity)
G raphita r 14 Piston rings (low velocity)
Haynes 25 Springs Inconel X
Hard facing (used on collet fingers, cam surfaces Colmonoy 6
and piston ring wear surfaces).
Guide surfaces Hard chrome plate on 17-4 Ph SS (heat treated at H 1050).
Index tube 17-4 Ph SS heat treated at H 1050 4.4.2.2 Detailed procedu res for heat treating, plating and hard facing, as well as inspection techniques such as ultrasonic, x-ray, magnetic particle and dye penetrant examination are used to insure highest
_ quality components. The drive has been designed, and materials selected, to conform to the requirements of the ASME Boiler Code.
4.4.2.3 The in-service failure of 17-4 Ph SS in the Dresden reactor has led to a careful review of the remaining application of this material in the drives. The long-term corrosion tests continue to support the choice of this material when used with the speci-fied heat treatment. Performance of the material in the Dresden drives is being closely monitored. The first inspection of six (6) modified Dresden drives after plant operation of 460,000,000 kwh rs, revealed that all of these drives were in excellent con'-
dition and showed no evidence of earlie r material problems.
4.4.2.4 Results of the extensive engineering terts on the Dresden drives are largely applicable to the Big Rock Point drives; many parts are identical in function. Additional tests have been run over a 16 month period on a development model closely resembling i
the final production design. Normal and scram operation under widely varying conditions have been investigated to confirm design margins. Studies have been made on the effects of high tempe ratu re, misalignment and malfunctions of the hydraulic control system.
Components such as seals, buffei, coupling i
and position indicator, have been life-tested.
4.4.3 Testing of Drives Each Big Rock Point drive will be operated under normal and j
scram conditions at low and maximum reactor pressure before shipment. Site acceptance tests will include normal and scram performance measurements with the drives in place.
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SGctisn 4-Page 23 Rev 1 (3/19/62) 4.4.4 Hydraulic Control System 4.4.4.1 Normal _ operation of the drive mechanism is controlled by ex-ternal hydraulic circuits utilizing demineralized water. Se-lector valves are supplied for each control rod drive? The circuit is interlocked to prevent operation of more than one drive at a time except during scram. Flow is regulated by manually adjusted orifices to produce the desired position-ing speed in both directions of operation. Differential pres-sure acting across the piston provides the operating force for the drives. A pressure slightly higher than reactor pressure is maintained on the bottom of the piston when the drive is not operating to provide leakage flow to cool and flush the drive.
The drive system is shown by Drawing M-122.
4.4.4.2 To raise the rod (decreasing reactivity), water at a pressure approximately 200 psi above reactor pressure is admitted below the piston, while the water displaced from above the piston is discharged to the drive cooling circuit. When the rod has reached the desired position as determined by the operator reading the position indicator, the control circuit is shut off, thus equalizing the pressure in the drive. The rod and piston tube then lower by gravity until the next lock-ing groove above the collet reaches the collet. At that point, the collet fingers enter the locking groove, and tia piston tube is locked to prevent further downward motior..
4.4.4.3 To move a rod down, a momentar y upward pressure is applied, as described above, to unload the collet fingers. A " downward" pressure is then applied (the rpverse of the " upward" pressure) to hold the collet fingers open and to move the piston rod down to the desired position. The control velve is then closed to isolate the drive from the control pre ssure, which permits the 1
collet fingers to enter the locking groove and re-establishes normal pressure throughout the drive mechanism.
4 4.4.4 Scram or emergency insertion of the rod is accomplished by similar drive flow paths as normal raising of the rod except that the rate of control rod insertion is considerably faster.
4.4.4.5 Thirty-two gas-water accumulators, ne for each control rod drive mechanism, are the sources of the hydraulic pressure d
required for the scram at low reactor pressures, while a shuttle valve within each drive mechanism admits reactor water to the drive when the reactor pressure exceeds accumulator j
l
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S;ction 4 Page 24 f
pressure. The accumulators are charged with nitrogen gas to a pressure which will deliver enough water to scram a fully withdrawn rod at low reactor pressures. The charge on each accumulator is monitored continuously by a pressure switch. The water discharged from the drives during a scram is collected in a scram dump tank which is initially at atmospheric pressure. The accumulators and dump tank are isolated from the normal driving circuit connected to each drive by scram valves which are held closed by solenoid-operated pilot valves.
i-4.4.4.6 Scram is initiated by de-energizing the scram pilot valves which allows the scram valves to open, thereby connecting the drives to the accumulators and sciam dump tank. The large differential pressure between the accumulators and dump tank rapidly drives the rods into the core.
4.4.4.7
- When the scram stroke is completed, the accumulator pressure continues to hold the rods in the reactor. If the accumulator charging system does not function, the internal rod drive shuttle valve shifts and applies reactor pressure under the
- iston, until the dump tank is filled and system pressures are equalized. When the differential pressure decays to a point where the rods can no longer be supported, the weight of the rods are supported by the locking mechanisms which are engaged to hold the rods at the fully inserted position. An interlock in the control system prevents withdrawal of any control rod until conditions have returned t a normal, and safety circuits can be reset.
4.4.4.8 The hydraulic system is thus composed of a central system for driving individual control rods up and down, and a scram system i
for rapid insertion of all rods. The central system is supplied by two full capacity pumps.
- 4. 5 LIQUID POISON SYSTEM i
4.5.1 The purpose of the liquid poison system is to provide a means l
of rendering the reactor subcritical and holding it subcritical l
while it cools down, in the event that the control rods are g
i unable to do so.
lt 4.5.2 The system is not intended as a backup for reactor scram l
functions, however, as most transient conditions that bring about a scram are too rapid to be counteracted by liquid in-jection. Conditions under which it is necessary to resort to l
(
Secticn 4 Page 25 i
Rev 1 (3/12/62) liquid poison would be quite extreme; such as several rods sticking in the retracted position or in the event that all of the various indications of reactor power level in the control room should fail or provide such inconclusive informa-Jon that the operator must take extreme protective measures.
'4.5.3 The liquid poison system consists of a spherical pressur e vessel located above the main steam drum, which contains 850 gallons of sodium pentaborate solution (Na2B10016 +
9 H O). This solution is made up on the site by dissolving 2
stoichiometric quantities of boric acid and borax (6H BO3+
3 B oO 6 + 9 H O) in water at 2120 F, and is B07 ---> Na2 Na2 3
l 2
4 selected for its relatively high concentration of boron.
4.5.4 The discharge line from the tank is routed to alternate injection points into the bottom of the reactor vessel and into the recircu-lating pump suction lines (See Drawing M-121). A pressure equalizing line is provided between the steam drum and the tank, and a pressurization connection is provided from a bank of 2000 psig nitrogen bottles (Drawing M-107).
4.5.5 Normally, the tank is isolated from the nuclear steam supply system valves in both the discharge and the equalizing lines.
Also, the line from the nitrogen bottles is normally valved-off l
so that the tank is normally at atmospheric pressure.
4.5.6 The valves associated with the injection of poison are designed for very high reliability and leak tightness. Each valve consists of a sealed inlet fitting in which flow is normally blocked by a precision machined shear plug, and a trigger assembly in which a ram is forced out by an explosive charge to shear off the plug.
4.5.6.1 The explosive charge is provided with two electrically fired 4
squibs per valve, each of which is energized from a different d-c circuit. Two parallel circuits supply all of the seven explosive valves so that an external failure or internal breaking of a circuit due to staggered firing will not pr event all valves from being energized. Also, each individual line to the fourteen charges is fused so that a short circuit in an element during firing will not disturb the remainder. In the normal position, all elements are connected in series including the control relays 1
and each circuit is continuously monitor ed, a failure in any part of a circuit causing an alarm to be given in the main control room.
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4.5.6.2 T his type of valve, but with only a single explosive squib, has a demonstrated reliability of 99.967. The pressurization line and equalizing line each have two full capacity valves in parallel t,
Sacti:n 4 Page 26 Rev 1 (3/12/62) and the injection line has three half-capacity valves in parallel so that the mathematical probability of valve failure is infini-tesimal. The explosive, or squib, life is a function of environ-mental temperature. The squibs provided in the poison system are specially suited for high temperatures and can be stored indefinitely at 1700 F; the maximum expected ambient temper-ature in this zone is 120 F.
- 4. 5. 6. 3 The explosive valves used are manufactured by Conax Corporation of Buffalo, New York, who have produced over 50,000 units.
These have been used y edominately by the missile industry.
The primary modification for application at the Big Rock Point is in the use of all stainless steel pressure parts rather than aluminum.
4.5.7 As the solution will start to precipitate at 90 F, the tank temperature is maintained at 150 F by a dual element electric immersion heater, with automatic temperatur e control. The poison tank is insulated, so that upon loss of power to the heater s, the solution will not cool to satur ation temperature for approximately 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
4.5.8 Upon initiation of the poison valves admitting full 2000 psig nitrogen pressure to the poison tank, poison is forced into the reactor within a few seconds; however, if the primary system is at full operating pressure, the nitrogen volume will be insufficient for forcing out more than a few gallons of solutiom The driving force for the remaining volume is achieved from the static head due to the elevhted position of the tank and the head act oss the recirculating pumps. If both recirculating pumps are down, a valve on the injection line into the pump suction is closed by an interlock on the pump mctor breakers, and poison is forced directly into the reactor through a check valve. The primary purpose of nitrogen pressurization is to insure positive displacement of poison solution when the reactor recirculation system is static, such as during refueling, when there is no initial driving head to establish a siphon through
't the discharge dip tube in the poison tank. As sufficient pres-sunzed nitrogen is available to start flow into the reactor with all drum safety valves blowing (1870 psia), all other operating conditions are considered less stringent. For example, suf-ficient pressurized nitrogen is available to displace the entire poison solution from the tank without benefit of siphon for e,ystem operating pressures up to 350 psig. At any operating pressure with natural or pumped recirculation flow, a minimum of 9 feet friction loss through the reactor will establish siphon through the injection line and empty the poison tank even without benefit of nitrogen pressurization.
4 S2cti n 4 Page 26A Rev 1 (3/12/62)
- 4. 5. 8. I _
The top outlet arrangement of the poison injection connection was selected to avoid precipitation of sodium pentaborate in contact with external piping and valves. The resulting siphon required is 3 feet when the tank is full and 9 feet when it is empty. Integrty nf this siphon, once established, is assured by the following: (a) small size (3") of injection line, (b) in most cases internal pressure prevents air leakage into the line, (c) air in-leakage is improbable under any circumstances as the associated piping and flanges are designed for 2000 psig pressure.
4.5.9 Two spring-return switches are provided on the control console to initiate poison injection and both of these must be held closed to fire the explosive valves, thus insuring that operation of the system is a deliberate action.
4.5.9.1 In order to allow the operator the option of interrupting poison injection, once st" ted, a remote controlled air operated valve is provided downstream of the three explosive injection valves.
This valve will open on air or power failure. If the valve is closed, it can be reopened, and poison injection will continue.
4.5.10 Within the first five minutes, sufficient solution is introduced into the primary system to produce a boron concentration of 1300 ppm. This is equivalent to -16% ak/k, which is more e
than adequate to reduce the power level to zero by offsetting the effect of decreasing voids. Injection of the remaining solution raises the boron level to 2000 ppm, which produces
-25% Ak/k, and is sufficient tp hold the core subcritical even after it has cooled completely and with all control rods removed.
4.5.11 All components of this system can be tested for proper cperation except that actual opening of the electrically actuated explosive valves would allow poison solution to enter the reactor.
4.6 BIOLOGICAL SHIELDING 4.6.1 Shielding Design Criteria 4.6.1.1 The principal radiations from which plant personnel must be shielded are neutron and gamma radiation from the reactor during operation, fission-product-decay gamma radiation from the nuclear fuel, and the gamma radiation of the nitro-j gen-16. Other sources of radiation include oxygen-19, neutron-capture activation pr oducts of reactor materials, and activated corrosion products kom the entire system.
Li S2ctirn 4 Page 26B Rev 1 (3/12/62) 4.6.1.2 The design of the shielding is based on the assumption that the maximum permissible exposure rate is five rems per year,
as well as applicab'e 13-week dose limits. The target weekly exposure rate limit is taken as 100 mrems/ week. In carrying out the above, the following maximum dose rates are estab-lished for the designated areas indicated on Drawing A-52.
4.6.1.3 Zone I Areas where access is not controlled:
- 0. 5 mrem /hr.
Such areas include the Control Room and adjacent areas, and outside areas around the process build-
.ings. Thus, exposure in such areas for a 40-hour week will not contribute more than 20% of the work-ing limit dosage of 100 mrems and probably will average less than 10%. This will reserve most of the permissible radiation exposure of plant person-nel for radiation zone entry.
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i Saction ~4 Page 27
- 4. 6.1. 4 Zone IA - Same as Zone I except access is controlled.
- 4. 6.1. 5 Zone II In certain cases where extended occupancy may occasionally be required: 1. 5 mrem /hr.
4.6.1.6 Zone III Reactor Enclosure and Turbine Building areas requiring periodic entry for sampling, inspection, auxiliary equipment maintenance, etc. : 15 mrem /hr.
- 4. 6.1. 7 Zone IV Infrequently entered areas: over 15 mrem /hr.
- 4. 6.1. 8 In the unlikely event of a " maximum credible accident" which might result in fission products being dispersed within the en-closure, the control room is sufficiently shielded to reduce the exposure to less than 0. 5 rem during the first eight hours.
This will enable personnel to occupy the control room to take emergency action to limit the magnitude of the accident.
4.6.2 Access Control Criteria
- 4. 6. 2.1 The plant area is enclosed by a peripheral fence and divided by internal fences in a manner such that plant personnel can enter or leave in an uncontrolled manner only those areas designated as Zone I (See Drawing A-52).
Areas designated as IA, II, III, and IV, which are considered as controlled access areas, can be entered only by passing through either the access control room or.a normally locked door or gate. Routine access to or egress' from e11 controlled areas is made through the access control roor..
Only in the event of an emergency or the per -
formance of certain locally controlled maintenance work will access or egress be permitted by any other route.
4.6.2.2 Personnel monitoring equipment and protective clothing storage are provided as part of the access control facility.
4.6.3 Shield Arrangement and Size 4.6.3.1 Ordinary concrete has been used almost exclusively for shield-ing of the primary system and auxiliary systems which contain radioactive material. A small amount of heavy concrete is used in the shield wall around the turbine because of space limitations, and lead jackets have been placed around the con-densate demineralizers in order to minimize the number of remote valve operators.
4.6.3.2 Shield arrangement and thicknesses are shown on the Equipment Layout illustrations, Drawings M-100 through M 104, inclusive.
I Section_4 Page 28 Rev 1 (3/23/62)
- 4. 6. 4' Shielding Functions and Accessibility 4.6.4.1 Shielding is provided to assure adequate radiation protection to personnel during plant operation and to isolate specific com-ponents that require either immediate access for operation or maintenance whue adjacent radioactive equipment is in operation.
4.6.4.2 All areas and rooms are accessible during plant operation, how-ever, personnel entry may be restricted into most areas desig-3 nated as Zone IV. It is expected that the immediate area around the turbine which is designated as a Zone IV will be entered from the unshielded side for limited inspection of the turbine during operation.
4.6.4.3 Access to "high radiation areas" is governed by administrative procedures and controlled by appropriate marking, locked barriers, locked doors or alarm doors.
4.6.4. 3.1 Locked barriers, appropriately marked, control access to areas in which the general radiation level range is 100 mr/ hour to 1 R/ hour.
4.6.4.3.2 Areas in which the general radiation field is expected to remain at a 1 s ;l higher than 1 R/ hour for a period longer than 30 days i
are r *ovided with locked or alarm doors. Entry through these doors will be controlled by a single master key which will normally be in the possession of the Shift Super
',or.
4.6.4.3.3 Access to "high radiation areas" requires the pn mission of the Shift Supervisor. No person not qualified to monitor radiation will be permitted toenter any "high radiation area" unless accompanied by a person so qualified.
3 4
4.6.5 Specific Activities of Coolant o
4.6.5.1 During plant operation, the specific activity of the reactor coolant loop is predominately due to the equilibrium Nitrogen-16.
q 4.6.5.2 Average activities in the various regions of the external coolant loop based on plant operation at 240 Mwt, and based on the 6. 5 Mev gamma energy of N-16 are tabulated below:
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Rev 1 (3/12/62)
TABLE 4.6 SPECIFIC ACTIVITIES IN PRIMARY COOLANT SYSTEM EQUIPMENT 3
gammas /cm
- sec.
6 Risers
- 1. 6 x 10 Steam Drum 9.4 x 10 (Water only-i uncorrected for voids) 6 Reactor Vessel
- 2. 5 x 10 4
Steam at Outlet Nozzle
- 7. 8 x 10 Downcomers 5.6 x 10 5
Recirculating Pump
- 4. 7 x 10 5
Pump Discharge Piping
- 3. 8 x 10 4.6.5.3 The specific activity of system leakage and in the coolant loop after shutdown varies widely, depending upon the decay time and relative quantities of corrosion products and fission prod-uct leakage into the water. The activity is expected to range 4
from 0.02 uc/ml to a high of 6 uc/ml shortly after shutdown; the basis of 6 uc/ml is an assumption of 1000 leaking fuel rods.
4.6.6 Biological Shield Cooling 4.6.6.1 A coolie.g jacket is provided at the inner face of the reactor shield structure. The coolant flowing through the jacket removes i
the major portion of heat lost by conduction and radiation from the reactor vessel and the heat generated within the shield due to energy absorption. The jacket is water cooled with a design in-(
let water temperature of 68 F; cooling water is supplied from the closed loop reactor cooling water system. In the event a leak should develop, it will be possible to convert to air as the cooling medium.
4.6.6.2 The cooling jacket is a carbon steel, annular tank divided into eight segments. It extends vertically from a point opposite the bottom of the reactor vessel to an elevation just below the reactor supports. There is a two inch annular water filled space between the inside and outside faces of the tank. Water enters the jacket i
at the bottom and leaves at the top.
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Page 29A Rev 1 (3/12/62) 4.6.6.3 The maximum expected temperature within the shielding is 110 F with temperature gradient of 13 F per' foot within
.tructural portions of the shielding. The maximum thermal l
gradient occurs within the inner 6 inches of the shield and is
'approximately 80 F per foot. Complete disintegration of the inner 6 inches of the concrete opposite the core can occur without affecting the structural elements.
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