ML19274E519
| ML19274E519 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/19/1979 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML19274E515 | List: |
| References | |
| NUDOCS 7903290167 | |
| Download: ML19274E519 (37) | |
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s SAFETY ANALYSIS REPORT FOR INSTRUMENTED CONTROL AND ORIFICING ASSEMBLIES
- 1. FOREWORD Core temperature and nuclear channel fluctuations have been observed at the Fort St. Vrain r.uclear generating station. While the fluctuations are irregular and complex, the average system parameters are not signifi-cantly affected. As part of a comprehensive diagnostic program to investi-gate and characterize the fluctuations, installation of in-core instrumentation is being considered. Installation of two in-core instrumentation packages can be accomplished by modifying two control rod drive assemblies. The safety evaluation of this proposed temporary use of instrumented control rod' drive assemblies during cycle 1 and/or cycle 2 is presented below.
- 2. PLANNED MODIFICATION .
It is planned to make temporary modifications to two control and orificing assemblies by removing one control rod from each assembly and adding an instrument package which occupies the space vacated by the re- . moved control rod. However, while the instrument package does occupy the space vacated by the removed control rod, the instrument package is not connected to the corresponding control rod cabic. Instead, the instru-ment package is isolated from and independent of normal operation of this remaining control rod.
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- The axial positioning of the instrument package is independent of the operation and positioning of the single control rod. It is intended that each instrument package will be manually inserted into and withdrawn from the core channel normally occupied by the removed control rod. Con-sequently, these operations are to be done at refueling conditions. Thus, at refueling conditions, with all control rods fully inserted, the instru-ment packages will be lowered into the selected regions of the core and secured. The instrument package will then remain stationary during subsequent reactor operation and testing.
The in-core instrumentation will consist of the following (the purpose of each instrument and its identifying item number on Figure 1 is also indicated):
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A. three thermocouples - to monitor helium temperature at three axial locations; item nos. 29, 30, and 31 on Figure 1 B. a self-powered neutron detector (SPND) with a compensating cable - to monitor the local / region flux level at the core midplane; item no. 27 on Figure 1 C. a fission couple - to monitor local / region flux level near the bottom of the core; item no. 24 on Figure 1 D. a fission chamber - to monitor region flux Icvel; located in bottom block of upper reflector; item no. 28 on Figure 1 E. a microphone - to detect changes in turbulence, i.e., flow; item nos. 25 and 43 on Figure 1. Item no. 25 (KAMAN microphone) is on C60 assembly S/N 20 and item no. 43 (GULTON microphone) is on CGO assembly S/N 43. These instruments will be securely attached at various axial locations to a sorport rod. The axis of the support rod will be maintained at a position near the centerline of the control rod channel by means of several
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spiderlike assemblies at various axial locations (Fig. 1).. The basic materials of construction are magnesium oxide, stainless steel, inconel and chromel-alumel. Joints will be welded. In addition to the above noted instrumentation, the two esntrol and orificing assemblies are provided with pressure transmitters located in the mechanism compartment. They are plumbed in a manner to measure the pressure differential across their respective orifice valves. One of the control and orificing assemblies is provided with a linear variable differential transducer (LVDT). It is also located in the mechanism compartment and is mounted at the upper end of the orifice valve drive mechanism. This device will provide infomation relative to the vertical movement (and/or growth) of the region's central column over which the control and orificing assembly is mounted.
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Other temporary modifications were required in order to pass the temporary in-core instrumentation leads through the primary and secondary closures. These included making two temporary ports in the primary closure piece of the control and orificing assembly, and the fabrication of temporary secondary closure plates containing four ports (see Fig. 2, in particular see views A and D). After exiting the temporary secondary closure plate, the temporary in-core instrumentation leads will pass between hold down plates through a notch provided _to accommodate their passage (see Fig. 2, view A). There are many pairs of refueling regions for which it is feasible to install the instrumented control and orificing assemblies. However, the following four pairs of refueling regions have been investigated as .A-1 prime candidate locations for the installation of the two instrumented control and orificing assemblies: 35 and 36, 4 and 35, 4 and 36, and 5 and 35 (Fig. 3). The effect on the power distribution, shutdown margin, A-1 rod withdrawal sequence, and maximum worth rod of installing two instru-mented control and orificing assemblies in each of these four pairs of A-1 regions has been investigated. - A-1. _. _ ._
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The relative axial positions of the various instruments, when installed inthNreactor,areshownonFigure4. During transport of the inst umented control and orificing assembly and prior to their insertion into the core, the instruments are withdr.'wn into the control rod drive assembly much Jike a withdrawn control rod. Following completion of all measurements requiring the use of the instrumented control and orificing assemblies, all temporary instrumentation is to be removed (at the hot servide facility) and disposed of. The drives will then be returned to their original operational condition; i.e., the removed control rod will be reinstalled and all nonfunctional ports required for the temporary instrumentation will be sealed. The instrumentation and the modifications to the plant to accommodate the instrumentation are shown on Figures 1 and 2.
- 3. SAFETY ANALYSIS Specific information for the basis of the safety evaluation follows:
A. AP Propriately designed seals permit passage of the in-core instru-
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mentation lead s~~through the primary and secondary pressure boundaries
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yhile maintaining the system's pressure integrity. All pressure retaining members are designed, fabricated, and tested in accordance with the requirements of Section 3 of the ASME Boiler and Pressure Vessel
. Code for Class I components for the primary closure parts and Class 2 for the secondary closure parts and demonstrated to be acceptable in accordance with the existing pressure retention and leakage criteria.
B. The operability of the reserve shutdown system and the orifice valve and its drive are unaffected by the modifications.
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- C. The presence of the instrument package in the control. rod channel will have an insignificant effect on the normal helium flow through
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the channel. More specifically, calculations indicate that the flows through the channel containing the instrument package, through .he corresponding channel with the control rod withdrawn, and through the corresponding channel with the control rod inserted are essentially the same (within 10'.) . These conclusions result from a comparison of the relative flow resistances for these cases. The values of flow resistance for the instrument package and for the control rod were determined by means of full scale flow tests (see Appendix I).
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The two principal resistances to flow through the control rod channel are at the location where coolant passes through two 5/8-inch diameter holes in the guide tube inside the plenum elements, and at the exit of the channel where the coolant leaves the channel through a 1-inch diameter hole. Friction and other losses in the channel are small compared to these for both the regular control rods and the instrument support structure of the ICRD as is shown in Table 1. The largest flow resistance occurs when the control rod is in any with-drawn position. Control rod components are then opposite the two 5/8-inch diameter holes in the guide tube and require the entering flow to turn after entering. For the inserted control rod or instru-ment support structure, the resistance of the channel is less than 15% A-1 of the total resistance so the flow is litt.le affected by the presence or absence of these assemblies. The flow in the case of control rods or instrument support assemblies inserted relative to the flow through the channel in which there is a control rod in the withdrawn position is shown in the last column of Table 1. A conservatively large value of instrument support structure loss coefficient, 1.8, has been used in computing the flow resistance with the assembly inserted as explained in Table 1. A more comprehensive presentation in support of the above is given in Appendix 1. It is concluded that the use of the instrumented assembly of the ICRD has a minor effect on the flow through the control rod channel. ,_
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TABLE 1 c , .
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.. Loss Coefficients Based on Channel Area (I) Flow Relative to Channel with Withdrawn Entrance In-Channel ( ) Exit ( } Total Control Rod
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Control Rod Withdrawn 454 0 395 849 1
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Inserted 275 , 85 395 755; 1.06
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Withdrawn 275- 0 395 , 670 1.13
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Inserted
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64(4) 395 73di' 1.08
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Channel area 12.57 in
- 1) =
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- 2) Assuming Tg , = 600 F T = 1000 F ,
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- 3) Assuming T = 1000 F .,
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From Table II of Appendix I loss for rod based on area of 2.3 in is 1,8. For 4-in; diam hole resistance = 4) 2 W 2 (T 4 )I = 54 1.8(( 2.3 /
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Correcting for. temperature rise 54 /060+140)
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D. The technical specifications require that the reactor must be capable of achieving a shutdown margin of 1% (0.01 op suberitical) at a core average temperature of <80*F with no xenon and the maximum worth control rod pair withdrawn (LCO 4.1.2). The following +.hree pairs of refueling regions were considered as potential location , for the installation of the two instrumented control and orificin.1 assemblies during A-1 cycles 1 and 2: 35 and 36, 4 and 35, and 4 and 36. Additionally, A-1 regions 5 and 35 were considered as locations during cycle 2. The A-1 corresponding shutdown margin was calcutated (using the physics design code GAUGE; FSAR Section 3.5 Ref. 3) for each of these four different A-1 configurations involving pairs of instrumented control and orificing assemblies. To show that the shutdown margin is satisfactory for these cases with the maximum worth rod withdrawn, the summary in Table 2 gives the calculated shutdown margin at the middle of cycle 1 A-1 and at the beginning (BOC), middle '(MOC), and end (EOC) of cycle 2. The middle of cycle conditions were'found to yield the minimum shutdown A-1 margins for cycle 1. In each case, the shutdown margin was found to A-1 be greater than 2% (0.02 op) . It should be noted that the calculated shutdown margins of >2% in-clude the following conservative assumptions. First it was assumed that all the Pa-233 had decayed to U-233. Secondly, since the scram time of a single rod on a modified control and orificing assembly is anticipated to be approximately double what it is with two rods, it was conservatively assumed that these single rods remain withdrawn during a scram. Thus, in calculating the >25 shutdown margin, it was assumed that three regions were unrodded; namely, the regions containing the two instru . mented control and orificing assemblies plus the region associated with the maximum worth control rod pair. E. The criteria serving as the basis for establishing the control rod withdrawal sequence are given in Technical Specification LCO 4.1.3. Calculations with the physics design code GAUGE (FSAR Section 3.5 Ref. 3) have demonstrated the acceptability of using the standard rod withdrawal sequence when the instrumented control and orificing
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Table 2 g.1
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- Shutdown Margins - Cycles 1 and 2
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Regions Assumed to be Unrodded , Shutdown Margin (%) Ap
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Regions Maximum Worth e Containing Rod Pair Cycle 1 Cycle 2 ' Instrumented CSO Assemblies Cycle 1 Cycle 2 MOC* BOC MOC EOC 35636 37 18 2.4 2.1 2.1 2.1 4 6 35 34 34 4.2 3.3 3.3 3.3 4636 37 37 3.0 3.0 3.5 4.4 5 6 35 N/A 34 N/A 3.3 3.1 3.1 A-1
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- MOC conditions yield minimum shutdown margin during cycle 1.
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. assemblies are installed in any of the proposed configurations for c,ycle 1 or cycle 2. The power distributions are well within the allowable limits set by LCO 4.1.3 in all cases. These calculations modeled explicitly the fact that one control rod consisted of boronated graphite compacts containing 40 w/o natural boron while the second control rod consisted of compacts conte.ning 30 w/o natural boron (see item H below) .
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'The power 1evels at which the candidate regions for instrumented control and' orificing assemblies become unrodded during the standard rod withdrawal se-quences of cycles 1 and 2 are summarized in Table 3 Once the singic rod A-1 of an instrumented control and orificing assembly has been withdrawn, it,
,of course, has no effect on the power distribution. When rodded, however, the presence of a single control rod in a region instead of a pair of rods does tend to increase the region's _ peaking facto _r (RPF) . _ _
_ _ _ ____ This effect is most significant for the case of installing the instrumented control and orificing assemblies in the adjacent regions 35 and 36 during cycle 1. Furthermore, the effect is most pronounced when the corresponding rod groups containing these two regions are fully inserted (at 5'18% power) . This results in about a 40-50% increase in the power in these two regions. However, these are low power regions (RPFs < 0.5) and the change is easily accommodated by adjusting the orifice valves. The power in the regions immediately adjacent to regions 35 and 36 also increases by 6 to 12% with the remainder of the core regions decreasing sli.ghtly (by 57%) or not at all (see Fig. 5). Although ti.e power distribution is somewhat different when instrumented CRDs are installed, the differences are' . well within the Technical Specification limits of LCO 4.1.3. When the control rod in region 35 is withdrawn (at ~205. power), only region 36 has a single rod and the differences are less. Similarly, in the case of one instrumented CRD in one of the peripheral regions (35 or 36) and one in region 4, the power distribution changes are smaller than those shown for regions 35 and 36.
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Table 3 _A_ - 1 Power Level at which Region is Unrodded* Region Cycle 1 Cycle 2 4 Subcritical Subcritical/ Critical 5 N/A Suberitical A-1 35 ~20% Suberitical 36 ~50% ~25%.
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- Assuming standard rod withdrawal sequence.
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- 11 During cyc1c 2, the only significant case is when region 36 contains one of the instrumented control and orificing assemblies. The other three candidate regions for instrumented assemblies become unrodded A-1 while the reactor is still suberit' cal, or just critical (see Table 3). A-1 As mentioned above, this effect '.s most pronounced when the single rod is fully inserted (at power levels <'25%) and results in about a 321, increase in the RPF of region 36. Again, however, the change is in a low power region (RPF = 0.56), the resulting power distribution is well within the Technical Specification limits of LCO 4.1.3, and the changes are easily accommodated by orifice valve adjustments (see Figure 6).
F. The calculated maximum control rod pair reactivity worth was likewise investigated and determined to be in compliance with the appropriate criteria of LCO 4.1.3. Thus, the standard control rod withdrawal sequences of cycles 1 and 2 are acceptable and do not need to be modified. G. Since the instrument package will extend the full length of the core isd since it has a low neutron cross section for ' capture, the effect of
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its prescnce on the axial and radial power _d_istributions is, negligible. _
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~The control rodi on the standard control and orificing assembly ~~ ' ~ - -
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H. of refueling region 35 consist of boronated graphite compacts containing 40 w/o natural boron. In the event that the pair of A-1 regions chosen for installation of the instrumented centrol and A-1 orificing assemblies is 35 and 36, the control rod installed - A-1 in refueling region 35 may have boronated graphite compacts A-1 containing 30 w/o natural boron while those of region 36 would A-1 contain 40 w/o natural boron. There is no significant differ- A-1 ence in the relative worth of these two types of control rods since they are both essentially black to thermal neutrons. The calculated worth of a single 30 w/o boron rod in region 35 is negligibly different from that of a 40 w/o boron rod, the calculated difference being less than 0.0001 AK.
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- In any event, as described above, the shutdown margin calculations s
-demonstrated compliance with technical specification limits while assuming this rod to be stuck in the fully withdrawn position.
Furthermore, power distribution calculations were performed which conservatively modeled a 30 w/o boron control rod in region 35 during the rise to power. The resulting power distributions were well within the technical specification limits of LCO 4.1.3 for both cycles 1 and 2. I. The remaining control rod in each of the two instrumented drives may be inserted or withdrawn in the normal manner. However, the scram insertion time for these two rods is anticipated to be approximately doubled. This characteristic is neither deleterious nor considered as a degration of reactor performance since, as noted in the preceding, these rods are not required to establish the shutdown margin necessary for compliance with Technical Specification requireme,nts. THIni: The slack cable assembly is adjusted to properly function with a single rod in the same manner as with a rod pair. The trip 1 point will be appropriately reset. J. Mechanical failure of an instrument package is analogous to the failure of a control rod. The bottom of the instrument package has been designed such that in the unlikely event of a failure the agrit to the control rod channel would not be blocked to prevent
-the flow of coolant gas (see Fig. 7) .
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-Nevertheless, conservative thermal and fuel performance analyses have been performed assuming the instrument assembly drops to the bottom of the_ channel and completely blocks the coolant flow while the plant ds operating at full power. These analyses show that, even under these imost severe conditions and if the failure of the instrument assembly awent unnoticed for 48 hours or more, the instrument package would mot melt and the impact on fuel performance would be negligible.
,Actually, if the instrument package should fall, the failure would result in anomalous readings which would give immediate indication of such a failure.
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K. Wpen installed, the bottom of the instrument package.will be about five inches above the bottom of the control rod channel. The conse-quences of the instrument package, which weighs 12 lbs, falling during full flow conditions and impacting on the bottom of the channel has ber.t evaluated and determined to be insignificant. The reflector block upon which the instrument package would impact has been designed and tested to withstand the impacts of regular control rods, with shock absorbers, weighing ~120 lbs falling from the withdrawn position. L. The probability of an instrument package failure has been minimized by careful design, the judicious choice of materials having high integrity at high temperatures, careful attention given to fabrication techniques (e.g., joints are welded), extensive inspection procedures to assure quality, pre-testing of the actual assembly and of mockups (see below), and by conservatively designing for large margins of safety. Additionally, the bottom of the instrument package has been designed to act as a catcher for small parts, should an unlikely occurrence of that nature happen. M. The support structure for the instrumentation is fabricated from Inconel 600. The stress on the support rod due to the maximum differential pressure drop across the catcher plus the static weight of the entire assembly has been conservatively calculated and found to be less than 105 of the tensile yield stress of the material at the highest anticipated operating temperature. The highest temperatures occur at the bottom of the assembly. Actually, the portion of the support rod which experiences the highest stress is at the top of the, assembly and is thus in a significantly cooler environment. N. Tests performed to confirm and support the mechanical design and fabrication of the instrument assembly included the following:
- a. The actual instrumented assemblies have been pull tested to 50%
of the tensile yield stress, followed by careful reinspection. The load placed on the assemblies during this test is a factor of five greater than the maximum anticipated tensile force.
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- _ 14 b, A mockup of the bottom three feet of the instrument support rod has been fabricated from production materials and tested at temperature to its failure point. This mockup represented all typical welds to the central support rod.
- c. A full scale model of the entire instrument package has been made of appropriate structural material, inserted into a simulated control rod channel and flow tested. Air was used as the test fluid and was drawn through the control rod channel so as to simulate the range of Reynolds number expected in the reactor.
- d. The ability to manually insert and withdraw the instrument assembly into the control rod channel has been demonstrated by testing. The control rod channel mockup for this test modeled the maximum possible amount of control block misalignment for each block. .
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- 4. IMPACT ON PLANT OPERATION ~ ~ ~
The presence of the instrumented control rod drives in the reactor will have essentially no impact on the normal operation of the plant. The only noticeable difference may be a somewhat different power distribution at power levels less A-1 than 50** and 25*4 of full power during cycles 1 and 2, respectively. This is easily accommodated by the normal procedure of adjusting the orifices as is done routinely during every rise to power. However, as can be seen from Table 3, if A-1 regions 5 and 35 are chosen for cycle 2, there will be no significant impact on A-1 the normal power distribution since these regions become unrodded while the A-1 reactor is still suberitical. , In any event, operation of the plant will be A-1 within all Technical Specification and design limits. As mentioned in Section 2, it is intended that each instrument package will be manually inserted into and withdrawn from the core. While it is expected that the instrumented control and orificing assemblies will spend
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- a relatively ahort time in the reactor, a conservative estimate of the dose rate from the lead wire bundle (and support rod) assumed three month's ex-posure at 70's power. The resulting anticipated dose rate one. foot above the lead wire bundle coiled in the external void space of the control and orificing assembly primary closure is anticipated to be 12 mr/hr. The expected dose rate to the hand holding the hotter portion of the lead wire .
. bundle is 100 - 200 mr/hr.
In the highly unlikely event that one of _the_ instrument _ packages __ ___ should fail and fall ('S inches) to the bottom of the control rod channel, the retrieval of the assembly would be accomplished much in the same manner as would the retrieval of a failed control rod. Such an operation would involve the use of the fuel handling machine and the core service mani-
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pulator and special tools (FSAR Sections 9.2.4, 9.2.5, and 9.2.6). The core service manipulator attachments and special tools include: a five-inch hand, an eight-inch hand, a hook, a fuel element grapple, a fuel element pickup probe, a control rod removal tool, a cable cutoff tool, a fue.' element righting tool, and a core service vacuum tool. These
-devices were designed to perform non-routine salvage operations in the space within the Prestressed Concrete Reactor Vessel (PCRV) that is accessible through the refueling penetrations, and are capable of reaching to the core support floor.
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- 5. CONCLUSION 1This change does not constitute an unreviewed safety question and
-there is no increase in the prebability of or the radiological consequences of a system failure. The change does not create the possibility of a radiological accident different from that evaluated in the FSAR, and the margins of safety as defined in the bases for the Technical Specifications
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0 4
\
APPENDIX I AIR FLOW TESTS OF INSTRUMENT SUPPORT STRUCTURE OF 11E INSTRUMENTED CONTROL ROD DRIVES
.
o I @ N e
- e-, e
_ - -
- Appendix I
.
.
.
\ .
- 1. INTRODUCTION Flow tests of a model of the instrument support structure to be used in the Instrumented Control Rod Drives (ICRD)were performed using atmospheric pressure air as the test fluid.
The primary objective of the test program was to find the flow characteristic of the assembly in the control rod channel. Several alternative designs were tested and a design was selected which provided about the same flow resistance as an inserted control rod. Therefore, the flow through the control rod cha.nnel with an inserted instrument support assembly is within about 10% of the flow which would pass through the same channel with an inserted control rod.
.
e e
.
O
.. Appendix I
. .
2.2 JEnd Piece The end piece was made up of a cylindrical section and a conical - section. Several different hole patterns were used in the seventeen different configurations tested. Configurations tested are shown in Fig. 3 The end pieces used in the reactor are 3.805 inches in diameter, have four 1.06 inch diame'cer holes in the cylindrical section at 60" angles. In the conical section, 377 holes of 0.06 inch diameter are drilled in a 135 sector. The reactor end pieces have a reference area of 2.3 square inches._ _ _ _ _ _ _ , _ _ 2.3 knstrumentsandLeads The model instruments and leads were approximately the same size
~
and one-third the weight of the corresponding reactor components. The materials used are tabulated below: - INSTRU'ENT HATERIAL USED IN ICRD Thermocouples 1/16" Dia. Teflon Rod Weighing .0029 LB/FT Fucrophone 1/2" Dia. x 0.8", .031 LB Weighted, Wooden Cylinder Self-Power Neutron .049 Dia. Plastic Covered Stranded Copper Wire Detector (SPSD) Weighing .0031 LB/FT SPND Compensating Cable .049 Dia Plastic Covered Stranded Copper Wire Weighing .0031 LB/FT Fission Chamber 3/16" Dia. Vinyl Tubing Weighing .0084 LB/FT
.
4 5
.
.
Appendix I .
.
s
- 3. TEST RIG AND INSTRUMENTATION
,
3.1 Test Rig For testing, the model was installed in a 42 foot long simulated control rod channel (see Fig. 2). The channel was represented by two 4-inch I.D. sections of cast acrylic tubing.! A guide tube was also included in the installation. Gaps which allowed gas ingress and which modeled reactor geometr/ 1:ere provided. These are: a) annulus. formed by the lip seal at the top of the guide tube and a nominal 5.075 inch diameter duct,- b) annulus formed by the lip seal at the bottom of the guide tube and a nominal 4.77 inch diameter duct, c) two 5/8 inch diameter holes which are part of the guide tube design, d) a 0.1 inch gap between two adjacent pieces of 4-inch diameter tubing just as in the plenum element design. In order to gain insight as to the effect of cross flow within the core, a gap in the tubing was added in the vicinity of the uppermost spider for several selected tests. The installation arrangement was such as to allow flow both into and out of the gap. A ducting system connected the test rig to the blower.
.
Ambient air was drawn into the control rod channel through the gaps. After leaving the model, the flow passed through a control valve and a flow meter prior to being exhausted to atmosphere through a single 250 horsepower blower. Due to the nature of the ducting arrangement, pressures everywhere in the system were sub-atmospheric.
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Appendix I
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Appendix 1
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L) h i s D = 3.7'5 Except - Chnfig.No.2 Dia.. = 3.805 ,
= 3.988 7 I)7 #
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Some Covered with Tape 2 Four .928 Dia. Holes, Equally -D Spaced, + N x .928 ]; [DT C Note: Reference Area = l section - Where N is number of open holes in conica _
.
Config. # of Holes # Holes(So. Reference In.) Area, Remarks Cylinder Cone No. 1.239 6 N/A .050 spine failed 1 N/A 1.120 2 4 .125 spine, cone holes added 3.268
-
4 3 3 1 1.916 4 4 _ End piece evaluation 0 1.239 4 Sim. Inst. added 5 1 1.916 6 4 Reactor T/C 1 1.916 4 Reactor Mic. 7 1 1.916 8 4 Reactor fission chamber 1 1.916 9 4 . Cages, ferrules added 1 1.916 10 4 Chan. inflow gap added 1 1.916 11 -4 Metered inflov 1 1.916 12 4 Metered outflow 1 1.916 13 4 4 3 945 14 6 ,
'
0 1.239 10 6 0 1.239 End piece mass doubled _ 16 0 0 1.239 17 6
..
.
FIGURE 3 - C0hFIGURATIONS w-% e
. . . Appendix I
. .
3.2 ,
,
Instrumentation Approx. The test measurements were: Accuracy
. air temperature at the exit of the cha .nel 2F
. channel total flow rate at exit i 1%
. crossflow gap flow rato i 3%
. static pressure difference across the 0.1 inch instrument assembly
, static pressure at the inlet to the < 1%
instrument assembly
. barometric pressure << 1%
Table 1 lists the equipment used to make these measurements.
.
0
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.
S S P e
Appendix I
.
.
.
\
4.0 RESULTS 4.1 lbdel Pressure Drop Pressure loss measurements across the assembly were made by two static pressure taps about 21 feet apart as shown in Fig. 2. Using these pressure losses and the measured channel exit flow rate in
-equation 1, the assembly loss coefficient was calculated. This coeffi-sient is based on the reference area given in Fig. 3 for each configura-tion.
.
The loss coefficient K for the assembly was defined as K = $E- (1) 9 K = assembly loss coefficient AP = pressure difference across the length of t'he assembly (due almost entirely to the loss across the end peice) q = velocity head through the end piece - based on the total flow rate through the channel and the reference area given in Fig. 3 for each configuration.
.
The minimum flow area in the control rod channel with the instru-
"
sent package inserted was located in the vicinity of the end piece, and it was found that the model pressure drop - flow rate characteristic could be correlated regsonably well to this area. This reference area was taken as the sum of the minimum annular area between the end piece and the channel wall plus +.he area of the holes in the
~ conical section measured normal to the surface. (The holes in the
.
A,pendix I
.
. cylindrical section were not used in the reference area).. Loss coefficients calculated from the data are tabulated in Table II. By basing the loss coefficients on the area at the end riece for each con-figuration, the loss coefficients for each configurat'on have all been reduced to the range K = 1.6 0.3 for open-channel Reynolds numbers greater than 30,000. The duct Reynolds number is based on the control rod channel cross section:
=
45 (2) Re spD when .
'
m = channel exit flow rate p = fluid viscosity D = control rod channel diameter = 4 inches' The duct Reynolds number expected in FSV control rod channels at rated flow is between 20,000 and 70,000, 4.2 Estimation of the Flow in the Control ~ Rod Channel The two principal resistances to flow through the control rod channel are at the location where coolant passes through the two 5/8-inch diameter holes in the guide tube inside the plenum elements, and at the exit of the channel where the coolant leaves the channel through the 1-inch diameter hole. Friction and other losses in the channel are . - - - - - -- --
._ . - . -
.. _ .. . . _ . .
1sma_1_ler than_ these for both _the regular.. control rods and the instrument support structure of the ICRD as is shown in Table III.1,. The largest flow resistance' occurs when the control rod is in the' withdrawn position. In the withdrawn position, the control rod's lower absorber is about 4-inches above the active core. Control rod absorber containers are then opposite the two 5/8-inch diameter holes in the guide tube and require the entering flow
.
..
/.ppendix I
.
*
.
to turn after entering. For the inserted control rod or instrument support structure, the resistance of the channel is less than 15% of the total resistance so the flow is little affected by the presence or absence of these assemblies. The flow in the contro) rod channel for control rods or instrument support assemblies rela +1ve to the flow through the channel in which there is a control red in the withdrawn position is shown in the last column of Table III. A conservatively large value of instrument support structure loss coefficient, 1.8, from Table II has been used in computing the flow resistance with the assembly inserted as explained in Table III. It is concluded that use of the instrumented assembly of the ICRD has a minor effect on the flow through the control rod channel.
.
.
.
O e
- 4 - Appendix I
.
.
, .
TABLE I LIST OF EQUIPMENT Identificat. ion Item Model Number Cenco Scientific Barometer 'TEC 13004 Meriam Micro Manometer H-14161 TEC 46035 Meriam Inclined Manometer, 10" Hg 40HE35 TEC 46023 BIF Universal Venturi 182 TEC 73100 Meriam U-Tube Manometer, 50" H 2O TEC 46036 Meriam U-Tube Manometer, 50" H O TEC 46025 2 Meriam U-Tube Hanometer,15" H 2O -- Meriam U-Tube Manometer,15" Hg -
--
Meriam Inclined Manometer, 2" H O TEC 46049 2 Meriam Inclined Manometer, 2" H O TEC 46015 2
.
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-
Appendix I . .
.
.
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.. _
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TABLE II - TEST RESULTS _ _ _ _ . _ . .
.
.
Bare Press Temp Dens t ty Press Drop Loen Coef Re Flow Este Cap Flav Rate confir (PSIA) (DEG T) (?E:f/CU FT) (PSTn) K 310-3 (LBM/SEC) (LEM / set') (1) 1/1 14.61 66 07353 411 1.44 37.6 .120 0 1/2 14.71 65 .07502 .166 1.28 25.7 .082 0 1/3 14.71 65 .07363 .554 1.52 43.8 .140 -- 1/4 14.71 65 .07085 1.432 1.52 67.1 .214 -- 2/3 14.71 64 .07464 .343 1.39 31.98 .102 -- 2/5 14.71 65 .07371 .577 1.46 40.1 .128 -- 2/7 14.71 65 .07103 1.566 1.46 64.9 .207 -- 3/6 14.69 65 .07292 .325 1.61 83.1 .265 -- 3/9 74.69 65 .07327 .278 1.597 77.4 .247 -- 3/10 14.69 64 .07389 .220 1.61 69.1 .22 -- 3/11 14.69 64 07433 .261 1.56 60.1 .191 -- 3/12 14.69 64 .07483 .101 1.55 48.0 .153 -- 3/33 14.69 64 .C7303 .078 1.57 41.9 .133 -- 3/14 14.69 64 .07499 .081 1.52 43.5 .13S -- 3/15 14.69 64 0753 .047 1.57 32.6 .104 -- 3/16 14.69 64 ' .07561 .011 2.07 L2.7 . 040 -- 3/17 14.69 64 .07567 .004 2.72 6.88 .022 --
. 3/18 14.69 64 .07548 .025 1.3?9 23.6 .075 --
3/19 14.67 58 .07619 .029 1.678 25.1 .079 --
- 3/20 14.67 51 .07632 .014 1.69 17.7 .056 --
3/21 14.67 58 .07636 .007 2.55 10.2 .032 -- 4/22 14.67 63 .07132 1.007 1. 64 8 84.2 .267 -- 4/23 14.67 63 .07198 .844 1.635 77.7 .247 -- 4/24 14.67 63 .07284 .642 1.622 68.5 .217 --
-
4/25 14.67 63 .07355 .476 1.62 59.3 .188 -- 4/26 14.67 63 .0742 .330 1.6 04 49.8 .158 ' -- 4/27 14.67 63 .07469 .224 1.61 41.1 .13 -- 4/28 34.67 63 .07512 .125 1.529 31.5 .1- -- 4/29 14.67 63 .07533 .081 1.638 24.6 .075 -- 4/33 14.67 63 .07552 .04 1.561 17.7 .056 -- 4/31 14.67 63 .07564 .011 1.316 10.1 .032 -- 4/32 .28 4/33 .2d
,
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Appendix I
.
.
s
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_ TABLE II (Cont'd)
.
Baro Press Temp Density Press Drop Lose Coef Re Flev Rate Cap Flow Rate Confit (PSTA) (Drc F1 (LRM/cU TT) (PSID) K x10-3 (tBM/src) (LEM/SEC)(1) 5/34 14.66 63 06862 2.241 1.8 76.2 .242 - 5/35 14.66 64 .07196 1.122 1.74 36.2 .179 - 5/36 14.66 63 .07414 473 1.61 38.4 .122 - 5/37 0 5/38 .24 6/39 14.67 69 .06985 1.262 1.73 90.2 .289 - 6/40 14.67 69 .07106 .934 1.73 78.3 .251 -- 6/41 14.67 69 .07192 .714 1.72 69.2 .222 -- 6/42 14.67 69 .07228 .624 1.69 65.3 .209 -- 6/43 14.67 69 .07267 .53 1.65 61.2 .196' ~ -
'
6/44 14.67 69 .07329 .375 1.64 31.8 .166 -- 6/45 14.67 69 .07366 .288 1.69 44.8 .144 -- 6/46 14.67 69 .074 .202 1.62 38.3 .123 - A/47 14.67 65 .07422 .148 1.67 32.4 .104 -- 6/48 14.67 69 .0744 .105 1.61 27.8 .089 ' -- 6/49 14.67 69 .07461 .055 1.49 21.0 .067 -- 6/50 14.67 69 .07474 .026 1.29 15.6 .05 -- 6/51 14.67 69 .C7481 .01 .89 11.5 .037 -- 6/52 14.67 69 .07484 .003 49 8.9 .028 --
. 6/53 14.67 69 .06996 1.226 1.74 88.8 .294 -
6/54 14.67 69 .06784 1.713 1.76 10.3 .329 - 6/55 14.67 67 .07065 1.118 1.74 85.4 .272 -- 6/56 .28 - 7/57 14.76 69 .07106 1.073 1.77 83.1 .266 --
. 8/58 0 8/59 .2 8/60 .28 9/61 0 9/62 . 2,8 10/63 14.66 64 .07119 1.100 1.84 83.2 .264 -
10/64 14.66 64 .07205 .873 1.86 74.1 .236 -- 10/65 14.66 64 .07252 .743 1.84 69.0 .219 - 10/66 14.66 64 .07341 .512 1.78 58.6 .186 --
'10/67 14.66 64 .07418 .318 1.8 46.1 .147 -
. *
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l - - TABLE II (Cont'd)
- - - -- !
Earo Press Temp Drnalty Press Drop Loss Coef Re T1ow Rite C.ip Tiov Rste Coeffr frSTA) (DFC F) (I BM/CU FT) (P$fD) K x10-3 (f rfjfSEC) (f.PM /N EC) (I) 10/68 14.66 64 .07459 .217 1.82 37.9 .121 -- 10/69 14.66 64 .C7502 .112 1.82 27.3 .087 -- 10/70 14.66 64 .07523 .065 1.56 22.5 .072 -- 10/71 14.6G 64 .07536 .033 1.40 16.8 .053 -- 10/72 14.66 64 .07545 .011 1.1 10.99 .035 -- 11/73 14.65 64 .07076 1.223 1.84 87.3 .278 Unmess e 11/74 14.65 64 .07169 1.198 1.84 86.9 .276 Unmeas 11/75 14.65 64 .07132 1.227 1.86 87.3 .273 Un=ess
' 1.82 86.4 .276 +.030 12/76 14.69 67 .07092 1.194 5
12/77 14.69 67 .07086 1.183 1.80 f.6.4 .276 t.021 12/78 14.69 67 .07074 1.158 1.81 85.3 .273 0 12/79 14.69 67 .07179 .573 1.77 75.4 .241 . 0 12/80 14.69 67 .07199 .884 1.8 75.4 .241 +.025 12/81 14.69 67 .07206 .884 1.80 . 75.4 .241 +.030 12/82 14.69 67 .07059 1.212 1.83 86.8 .277 0 12/83 14.69 67 .07036 1.205 1.82 86.8 .277 +.025 12/84 14.69 67 .07092 1.197 1.81 86.8 .277 +.030 12/85 14.69 67 .07109 1.057 1.72 83.8 .268 0 12/86 14.69 67 .07115 1.111 1.80 83.8 .268 +.023 12/87 14.69 67 .07122 1.109 1.79 84.1 .269 +.030 13/88 14.68 67 .07072 1.154 1.77 86.0 .275 0 13/89 14.68 68 .07046 1.125 1.78 64. 5 .27 .021 11/90 14.68 68 .07039 1.114 1.78 84.1 .269 .030 13/91 14.68 68 .0717 .775 1.76 71.1 .227 .030 13/92 14.68 69 .07121 .909 1.76 76.6 .245 .017 13/93 14.65 69 .07112 .891 1.76 75.8 .243 .028 11/94 14.68 69 .07118 .896 1.77 75.9 .243 .024 13/95 14.68 68 .07058 1.154 1.79 85.5 .273 0 14/96 14.69 69 .07327 .218 1.63 81.5 .261 0 14/97 14.69 69 .07205 .353 1.63 103.0 .330 0 14/98 14.69 69 .07114 .444 1.63 114.4 .367 0
- 14/99 14.69 71 .06916 .615 1.65 131.8 .423 0
.
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' ' - Appendix I
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~
. TABLE II (Cont'd)
Fress Drop tees Coef Re Flow state Cap Flow Rate 64ro treae Tesp Decelty K n10-3 (1.?'if 5FC) (f1M / $1'C) (I) r.efig (P'ff% (!37C F) ff 5'/.T IT) (f's t T*) 1.64 31.4 .101 0 1%f 3 re) 14.69 72 .0735 .332
.451 1.66 36.3 .117 0 l',/ s ol 14.69 72 .07312 0
#
72 .07212 .779 1.73 46.3 .149 15/102 14.e9 1.035 1.78 53.6 .173 0 15/103 14'69 72 .07118
.g .
.227 73 .06905 1.701 -
1.56 70.5 -- 16/104 14.69
.07331 .310 1.59 30.7 .099 --
17/105 14.67 74 74 .07257 .494 1.57 38.7 .125 -- 17/106 14.67 1.58 50.5 .263 -- 17/107 14.67 74 .07145 .858
.07072 1.095 1.61 56.3 .182 -
17/103 14.67 74 50TE: (1) + Denotes cas inflew *
- Denotes Gas Outflow
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- TABLE III Loss Coefficients Based on Channel Area ( ) Flow Relative to. Channel with Withdrawn Entrance In-Channel (2) Exit ( ) Total Control Rod
. . .
Control Rod Withdrawn 454 0 395 849 1 Inserted 275 85 395 755 1.06 ICRD Withdrawn 275 0 395 670 1.13' Inserted 275 64(4) 395 734 1.08
-
- 1) Channel area =
f(4) = 12.57 in
- 2) Assuming T = 600 F T = 1000 F h ,.
- 3) Assuming T = 1000% g out 2 4
- 4) From Table II loss for rod based on area of '2.3 in is 1.8.
= 54 For4-indiamholeresistance=1.8(2.3 )3 Correcting for temperature rise E
g 060 + 1460 54 = 64 ( 2 (1060) j ~
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