ML20002C950

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Submits Further Support to Licensee 640326 Request for 24 Changes to Tech Specs Per Discussions W/Aec
ML20002C950
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 05/21/1964
From: Kettner R
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To: Case E
US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101150308
Download: ML20002C950 (22)
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Text

{{#Wiki_filter:., f,e % h ~,/ a c Aa CONSUMERS POWER COMPANY ' - t I ' U3 GENERAL OFFICES, JACKSON.MICillGAN s .k,b $ N[D May 21, 1964 R. F. ECITN ER _ * . t cp .g plasc7tm or Mtu2AR ACO ' W 3,1j Q : 1- _ g -s,-pa Iy e ~~ v o M,v\\;.- BIG ROCK POINT NUCLEAR PLANT -'q / N Yi G Q I_ tb ,{ o,, -{

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Mr. Edson G.' Case D %~ [.:/ Acting Director ? ./ Division of Reactor Licensing Qs W s ing on O N Tilo Cory

Dear Mr. Case :

Docket No. 50-155 Recent discussions with your staff have prompted this letter as further support to the March 26, 1964 Consumers Power Company request for certain changes to the Big Rock Point Nuclear Plant Technical Specifications. As previously stated none of the changes is believed to present significant hazards' considerations not described or implicit in the Final Hazards Summary Report (dated November 14, 1961) as amended. Please note that in the original document (dated March 26, 1964) the proposed changes were grouped as 47 different items or changes. To more easily associate this information with its appropriate reference in the original document,the same change numbers are used herein. Change No. 1. The definition proposed on March 26, 1964 for " power operation," 1.2.1, does not appear to cover the interval of time between criticality and the time the reactor-reaches 212 F. There-fore it is proposed that the definition be changed to read as follows : "1.2.1 Power Operation - is any operation other than shuidown or cold shutdown with the reactor vessel closures oolted in place." It is believed that this definition fully covers all conditions of ' power operation and is compatible with the other definitions proposed for this section. LMu- - G PMN9.M2

2 ' Change No. 6. The decay factors and counting efficiency factors have been analyzed to determine the proper pc/ml limit for halogens when counted at two hours after sampling, and using the I-131 counting efficiency. The two hour decay of an ' equilibrium mixture of halogens lgives a factor of approximately o.46. The ratio of the estimated weighted average counting efficiency of an equilibrium mixture of halogens to the efficiency of -I-131 gives a factor of approximately - o.53 Thus the equivalent value of the 35 pc/ml would be o.46 x o.53 - x 35 = 8.5 pe/ml. - It is proposed that 8 pc/ml'be used as the limit. Change No. 9. It is proposed that another sentence be added to specify load limiter operation. The proposed change would thus read as follows: "During turbine start-up and shutdown, and for short periods of time during normal operation, operation on speed control shall be permitted. During such opera-tion, the turbine load limiter shall be set to limit turbine output to correspond to the planned reactor output." 4 Change No. 11. For clarification, modify the proposed change to read as follows: " Condition

Pressure, Head off

$ 09 Head on - Relief Valve Settings + 1250 1470 1500 1800 2 1750 2080 [hange No. 12. Consumers no longer requests consideration of this change as recent physics testing experience has eliminated the re-quirement for same..Therefore this proposed change should be deleted from further consideration at this time. l 2 e c - y y y -w,,- w-

.~. 3 Changes'No. lk and 15 The following additionalidata on control rod ' drive testing :is furnished in support of Changes No.14 and 15 A. Scram Tbsting - Twenty months _ of operations at Big Rock Point following initial plant start-up have resulted in many scram tests ~on the control rod drives. These_ tests are summarized as follows: 1. During the acceptance testing prior to initial criticality, all drives were scram timed a minimum of 25 times each. The average time for all drives (to the full in position) 'for all scrams was_0 975 second. The slowest. time was 1.25 seconds and the fastest vas-0.82 second during this testing.

2.. Since September 27, 1962, the drives have been scram timed on ten different occasions.

The average time for all drives during this period was 0 96 second with the slowest' time being 1.25 seconds and the fastest time being 0 73 second. A tabulation of the-testing follows which demonstrates the excellent scram consistency of thest drives. Consumers considers that this consistent performance, plus the faat that the drives are-scramming in less than half of the allowable scram time, clearly demonstrate that a six mon +' L' testing interval is appropriate. Ns s -m~i

b TABULATION OF SCRAM TIMES' original Scram Average for'All Fastest Slowest-Drive . : Times ( Avg) - Serial No. L(Acceptance Test).. ~ Recorded Scrams Scram Scram Since Installation Time Tir. _ -58 .95 96 .88 .1.15 -65 96 1.01-94-1.15 '39. 92 - .97 91 1.05 66-94-93 .86 1.01 42-1.02 93 .85 -1.01 38 - 94 96 90 1.05 68 94 97 .85 1.17 -60; 94 98 92 1.15 ST: -1.03 1.04 95 1.20 59 -.98 1.00 90 1.05 49 1.08 - 1.00 .85 1.05 51-99 1.03 94 1.25 52 96 1.02 91 1.09 .46~ 94 1.00 94 1.08. 69. 94 1.00 94 1.05 50 1.00 1.00 92 1.10 48 92 95 92 1.01 -40 93 97 93 1.02 l 34 97 95 93 98 61 94 1.10 92 1.18 64 1.01 95 .88 1.01-41 945 1.02 94 1.05 { 35 951 95 .87 1.05 54 .89 .89 .84 95 43-90 92 .84 1.00 1-55 95 95 .87 1.06: j 67 97 95 90 1.03

37 96-97 92 1.01 j
53 95-97 91 1.05 62 96 99

.82 1.15 45 96 93 .86 1.00 47 94 93-73 1.20 44' 93' 93 92 95 -56 98 97 93 1.00 ~ - - - yn w w p ,-4--

F-5 B. Frietion Testing - Friction testing of the Big Rock Point control rf.1 drives consists of measurement of the minimum pressure, under -1 *le individual drive piston, required to move the drive from one index tube notch to another. This pressure is monitored at each notch and must overcome the combined weight of the control rod and its attached _ drive internals, friction within the drive mechanism, and friction among the control. rod and its four sur-rounding support-tube-and-channel assemblies. The Big Rock Point drive system has been friction tested 11 times to date, with one test accomplished prior to initial plant start-up. The friction readings dc not-present a precise pattern and tend to vary somewhat. However, the general pattern has been in the direction of a slight increase in friction over the past one and one half years (readings of 55 to 60 psig to readings.of 60 to To psig). Changes of this magnitude have not caused any jog-ging difficulties or any noticeable increase in scram times. on occasion,_several drives have shown friction above 80 psig. Plant operating procedure considers this to be above normal and requires investigation. ' Friction above 80 psig has always oeen evident from jogging response. Therefore, the additional re-quirement fov friction tests is duplicative and merely further confirms tainting knowledge. Again the scram times have been shown to be very insensitive to these friction changes. The following table is typical of drives that have shown friction increases and illustrates this scram time insensitivity. Consumers is confident that drive friction would have to in-crease beyond the point where normal drive withdrawal was pos-sible before scram speed could be significantly affected. Scram Scram Scram -Drive No. Friction Speed Friction Speed Friction Speed 50 58 1.00 To 97 75 1.05 65 60 96 To 94 T7 1.15 h2 57 1.01 68 .89 72 90 54 59 .89 66 90 85 92 56 58 98 To 1.11 88 1.00 By way of further illustration, the highest drive friction experienced has been on Drive No. 56 where recent measurements at the middle third of the stroke revealed friction above 100 psig. At the same time the drive could not be jogged through these hign friction notches, yet the scram speed was not noticeably affected. Subsequent disassembly of this drive plus removal and inspection of the support-tube-and-channel assemblies surrounding the control rod revealed the high friction to have been caused by roughening on one surface of one of the support-tube-and-channel assemblies by the upper and lower guide rollers on one blade of the control rod. Replacement of the surrounding support-tube-and-channel assemblies plus the control rod immediately returned the

6 friction to normal, again without noticeable changes in scram speed. The roughening is believed to have been caused by the apparent defective rollers. Investigation of the straightness of the support-tube-r ad-cPannel assemblies involved is under way. It can be postulated tnc- *oller defects on thir. blade could - have been caused by the core internals' vibration prcblem which existed before the installation of the flow distributor. Jogging ability of all drives vill be continually monitored,as previously, and during the next scheduled shutdevn an inspection vill be made of the replacement channels. The Jogging circuit is very sensitive to increases in drive friction,as mentioned previously, and serves to warn the operator that changes in drive friction are taking place. For a jog with-drawal, movement of the jog sviten in the withdrav direction energizes a time delay relay which gives the drive selected an insert signal of approximately 0 3 second. This is sufficient to lift the drive off its collet fingers. Immediately after this relay times out, a racond time delay relay gives the drive a withdraw signal of appcoximately 0.5 second, which is suf-ficient to move the drive down past the now open collet fingers. The settling circuit remains open several seconds longer to permit the drive to settle into the next notch. It can be seen that abnormal friction at a point on the drive stroke vill slow the drive movement sufficiently so the drive vill not withdraw to the normal jog signals. Such occurrences are noted by the operator and are a positive indication of drive friction changes. Cnange No. 16. Delete " cold" from the proposed third paragraph. inis paragraph should now read: During power operation, if reactivity and control rod motion data indicate a possible loss of poison from a control rod, the reactor shall be shut down and, if any correctiv.' action is necessary, shall remain in shutdown condition until such corrective action has been taken." Change No. 20. In the interest of conformity with other licensed boiling water reactcrs, the following is proposed for Section 5 3.h(a) : "5 3.h(a) The power coefficient shall meet the following re-quirement at all times during the reactivity life of the core: (a) The.ontrol 2 nd movement required to produce a given power change viR. tha reactor pressure constant and the reactor water at saturation temperature with this pres-sure shall always infer a negative power coefficient."

m ~ L f,C LAnalysis: ;The moderator void coefficient in the Big' Rock Point' reactor is expected to' be negative in all core sizes and at all-temperatures. . Experiments lhave been~ performed for the initial. fuel bundle types at-m - ambient temperature in all core leadings _ from the minimum critical con - figuration in both zircaloy and stainless steel channel' types to the full core 56-bundleLloading. It was shown thatithe moderator. void

coefficient was safely negative Lin all representative core locations

~and~vithin an order of magnitude of the predicted value. The moderator. void coefficient has'been' demonstrated'to be negative at. the.several operating temperatures by requiring control rod withdrawal-to increase power and steam flow..The operating ' void coefficient (the power coefficient) has been demonstrated to be negative in this manner

in core sizes 'up to 74 fuel-bundles.

R&D type fuel bundles will be used in two core sizes - 84 bundles and approximatelyhhbundlesforthe60kw/ldemonstration. The temperature' ccefficient of these bundles in the 84-bundle core has been calculated (and'shown.to be very similar to the initial fuel. The temperature coefficient of the 84-bundle core will be measured and it is expected f - that the reactor will demonstrate a slightly higher turn-around temperature i-than'previously measured-cores (due to fuel exposure-and increased core size). Demonstration of a small change'in temperature coefficient is sufficient evidence tbst the void coefficient will be changed by a small amount. Void coefficients are generally calculated bf a method similar to the [ following: The void coefficient (dik/k % void) is equal to the suu of terms' involving first the effect of a change of k of the bundle due ay a to void' change, second a term reflecting:the change of leakage due to l a void increase (always a negative contribution), and third the effect of increased control worth due to an increase of voids in the core j. (again always negative). Only one term-could add a positive effect, and l t ils a positive change of bundle reactivity with increased voids. The initial fuel type (3 2% enriched, 2 7 water-to-fuel volume ratio) . with its overall negative coefficient has a negative contribution due to this reactivity change. All bundles loadei subsequently and which are now scheduled to'be loaded demonstrate a negative centribution due e i to increased enrichment, lower water-to-fuel ratio, or both with the following three bundle exception: the 2.8% enriched, 2 96' water-to-fuel i ratio, Phasc II R&D c. ries. The three bundle contribution to the whole core, while still negttive, is entirciy negligible. The leakage term is always large and negative, diminishing only slightly in effect in a larger core. i The.last term reflects *.he negative contribution of control rods in-serted into the. core. The fuel types which have been loeded into l the' core since the initial 56-bundle core and those fuel types planned for loading are,without exception, higher in reactivity than the initial fuel ~ type and will show always a more negative or equal control contri- .bution relative to'the initial core. i = =

8 . Calculated void. coefficients have never been intended to describe exactly _a measured void coefficient due to the very difficult calcula-tion-involved. _ Experimental void coefficients have been shown, when measured, to be within an-order-of magnitude of the_ predicted value and always negative' for the ~ water-to-fuel ratio and enrichment range 7 used in other similar reactors. -7he wording in Section 5 3.k(a) was changed to restrict the requirement to saturated conditions; these are.the only meaningful conditions when specifying a power coefficient in a bWR. 4 Change'Nck 21.- Ib cover the special case of removal of a control rod from the reactor vessel, it is proposed that a footnote be added to Section 5 3 5(a) as follows: "* Removal of a control rod from the core by_means cther than normal control rod drive movement shall require that. the four fuel bundles surrounding the control rod first be unloaded and that the control rod be properly re-installed and checked out prior to reinsertion of the fuel bundles." 4 - Change No. 22. In the first sentence of Section 6.2.1, substitute-i the words "five decades above source level, without moving the detectors (approximately10-9 to lO-b% of rated power). Change No. 23 In Section 6.1.2.2 change the last sentence to read l ' as follows: "The detectors shall be gamma compensated ion chambers with a design sensitivity of at least 2.2 x 10-10 ]a amperes /nv." Analysis: The CICs originally installed in the plant and still being - used have a design sensitivity of 2.2 x 10-14 amperes /nv. The original - Big Rock Point Tbchnical Specifications erroneously reported the sensi-tivity as b x 10-14 amperes /nv. Si ~ sensitivity of more than 2.2 x 10-1gce the beginning of operations, aamperes/nv hasj Pr 'hermore this sensitivity is not contemplated as a future requirement. This is quite evident from the following information on overlap of the - various ranges of instrumentation taken with the 84-bundle core under f cold' conditions:

. _ = 9 ~5 -3 -1 ~9-lO'I 10 10 _1o 1 to 1oo 1ogo. -10 I _% Power 1 1-l ~ 'l l I I i i 2 Additional Log Count Rate Decades When -5 Decades 1 to 105 Cps 9%to10"N% N--> Chamber Is 4-Approx.10 d Log-N Period Amplifier 7 Decad s

  • --- Approximately 10"g% to 100% N 1

Picoammeter 9 De ades Approximately-125x10g%'to125%N ~ l: i . Change No. 2h. Change the last sentence of Section 6.1.2 3 to read as follows: "The detectors shall be gamma compensated ion chamberb with a design ~ sensitivity of at least 2.2 x 10-1h amperes /nv. The amplifier output shall be connected to the reactor safety system." Analysis: The above change as explained in Change No. 23 reflects the sensitivity of the CICs from their original' installation until the present time, and no change of detector sensitivity is contemplated in the future. The change is also more technically correct, since the CIC outpat is fed j . to the picoammeter, whose trip circuit is connected to the reactor safety system.' Change No. 28. Analysis: In the December 20, 1963 Special Report SR-6, " Reactor Safety l System Detailed Operating and Maintenance Experience," experience with Big Rock Point Reactor Safety System was described; and it was indicated-that Consumers believed the system would continue to provide the necessary protection for safe plant operation. Since issuance of "SR-6" and the' . relocation of the steam' drum level sensors sin December,1963, there have - i m-,e%~ ~

y s 10 been no abnormal operations of the reactor safety system. In addition, the daily simulated sensor trip to check the logic system (in effect since August, 1963) has shown no abnormalities. During the last' plant shutdown, - overvoltage protection on the reactor protective system motor generator sets was installed. It was felt that overvoltage was the prime cause for - the two shorted transistors discussed in "SR-6." Because of the experience on Big Rock Point and other nuclear plants using similar type reactor - safety systems, Consumers has extended the interval of time between logic system checks from the daily to a weekly check. The regular monthly trip tests will.be continued. The equipment manufacturer concurs with the above.- The rewrite of Section 6.15(a) more specifically describes the monthly tests of.the reactor safety system. By tripping one channel at a time, the entire " portion.of the reactor safety system can be checked, except for the sensors and connecting leads. Change No. 29 The following additional information describes the core spray and emergency condenser instrumentation and tests to date in support of the request for the change in testing frequency. The core spray initiation sensers consist of four reactor water level - switches and four reactor pressure switches. There are two sets of these level and pressure switches on each one of the two core spray valves. Including the original acceptance test calibration of these devices, 8 calibrations have been performed on each level and pressure switch since August 1962. The as-found calibration.of the 4 level switches has been within I " of the desired set point of the switches on all 8 tests. The l as-found calibration of the-4 pressure switches has been within 5 psig of the 200 psig set point on all 8 tests. The level calibration censists of placing a temporary water column on the level sensors and then draining the water column and noting where the switch actuates. The pressure switch calibration consists of applying a pressure to the Mercoids and noting the pressure required to close the contacts. In afdf'lon, motor-operated valve operation is cblained by closure of the correct combination of initiation sensors. Each emergency condenser motor-operated valve is actuated by closure of two combinations of 4 reactor pressure switches. Including the original acceptance test calibration of these devices, there have been 8 calibra-tions run on each pressure switch since August, 1962. The as-found calibration has been within 110 psig of the desired set point of 100 psig above reactor operating pressure, which is 100 + 1250 or 1350 psig at present. The pressure switches are calibrated by applying a pressure to the ' sensors and noting the pressure required to close the contacts. In . addition,-motor-operated valve speration is obta'ned by closure of the correct pressure switch combinations. The above tests on the core spray and emergency condenser systems have - proven satisfactory in all cases, and it is our opinion ~that a yearly test on these systems would be more than adequste. It should also be noted that either system may be manually actuated from the control room. s -s y-,., .-.-w-

m 11' T Change >No. 32.- It'.is_ Consumers' opinion that a functional" check once a - year of the control rod withdrawal permissive circuits will-be more than ! adequate ~ to assure proper operation of/ the interlocks. Including the' e acceptance test: cheek-out of the permissive circuits on the manual' reactor. control system, 8 separate checks have been run with no abnor-malities on-the system.' In addition, the interlocks have always t . functioned correctly during normal' operation. The check of the circuits consists of simulating or;actually setting up conditions to test out the refueling, pico, accumulator, and mode switch interlocks. j Change No. 37. Big Rock Point experience with gamma and neutron dose; -rate measuring inr,truments shows very little need for monthly calibra- . tion. Difficulties experienced on' the instruments include the' follow-T ing: minor difficulties consisting of cleaning, tightening parts, repairing ~ wiring, and major troubles consisting of repairs to meters l damaged by dropping, replacing electrometer tubes, resistors, and trans-formers. The following is a list of instruments with the difficulties experienced to -date : 8 Cutie Pies CP-3 5 minor troubles and 8 electrometer tubes replaced. 1 3 Jordan Radguns. 2-minor troubles and 1 tube failure. 2 Neutron Survey Meters. No maintenance problems. l After repairs to instruments they are recalibrated prior to being made available for use and this practice will be continued. The usage' frequency of the above instruments varies widely.. However, there have been no occasions where a monthly calibration check indicated an abnormality that would not have been noted otherwise.- From monthly calibration data, the maximum drift'obtained for any 3-month period 4 { vould be I %. The radguns have built-in check sources, while all instru-3 l ments have battery checks. Change No. 38. The counting efficiency and decay concepts described under Change No. 6 are applicable to the calculations of stack release j of iodines. The same general procedures apply, i i r t I, r t-4 ,,--,.n-es.-,-~ ..e, e,,-r ,.e , ~ -, ,m-, --.n ,,--n,.-, + y, ,.,a,--,,e ers -- e

1 12 i. Change No. 42.- Rather than as proposed ~on March 26,.1964, change Section 7 4(a) to resd~as follows: "Detailsd written procedures shall be available prior to each' refueling operation." Change No. 45 Section 7 5 7 should 'be further modified to read: "'It shall be permissible to remove a control rod drive from the core when the reactor is in the shutdown condition and. the mode selector switch is locked on the " Shutdown" or '" Bypass Damp Tank" position while the drive is removed. The core shutdown margin of 0 3%Ak rf/kerr with the strongest e rod out of the. core'shall have been met prior.to the drive removal; and, in addition, the equipment shall be properly tagged. The control rod drive that -was removed shall with-out delay be replaced by a spare drive or the original drive shall be reinserted.'" . Analysis: This proposed change clarifie. the drive removal procedure and allows drive removal under the newly defined " Shutdown" condition. The existing Section 7 5 7 is incompatible with the definition of " Cold Shutdown," in that the latter requires all rods to be inserted. The

control blade must be withdrawn to provide a water seal on the drive nozzle, and then is delatched before the hydraulic system of the drive is isolated. When the mode-switch is in the " Shutdown" or " Bypass Dump Tank" position, a second control blade cannot be withdrawn, assuring that the shutdown margin is maintained during the drive replacement.

Change No. 46. It is proposed to add a new section entitled 8.2.4 to describe and define-the Phase II R&D Program natural circulation tests to be performed as follows: "8 2 4 Natural Circulation Tests 7 During Phase II of the B&D Program, selected tests may be conducted under conditions of natural circulation. These tests may be performed with either the 84-bundle or 60 kw/ liter core loading and within the range of variables as-specified in Section 8.2.1. In all cases, any mode of natural circulation operation will have been shown analytically to be within the following limits for the specific flow rate applicable to the given operating conditions. In addition, as established during Phase I tests, testing will be performed at increasing increments of power.to compare analytical calculations with actual conditions. The maximum operating power level vill be

13 ' that which corresponds to 100% where the - calculated limiting conditions 'are considered at 122% overpower. -Picoammeter trip set points shall be adjusted.as neceacary_at.each test step to limit power level to

permissible values.

8.2.4.1 Operating Limits Minimum Overpower. Burnout Rati 2 1*5 Maximum Heat - Flux. at Overpower, Btu /Hr-Ft 530,000 2 Maximum Steady State Heat Flux, Btu /Hr-Ft h34,000-Maximum Fuel Rod Power at Overpower, Kw/Ft 17 2 Maximum Steady State Fuel Rod Power, Kw/Ft 14.2 Stability Criterion: Maximum Zero-to-Peak-Flux Amplitude, Percent of Average Operating Flux 20 Maximum Steady State Power Level, Mvt That Permitted ~ By Other Opera-ting Limits Maximum Reactor Pressure During Power Operation, Psig 1485 -8.2.4.2 General Procedure a Phase II natura] circulation tests will begin from a normal forced circulation mode in the following general steps : a. Trip both recirculation pumps from an initial steady. operating condition, not to exceed 157 Mwt. b. After power coast down has settled out, increase power in approximately 20 Mwt st w,s, while measuring power level, flow rate and flux noise amplituue to confirm observance of specified operating limits. c. At selected operating points, perform Phase II tests such as rod oscillstor tests, d. Upon termination of natural circulation tests, restore orced circulation mode after reducing power level to 't obtained in Step b.above after tripping pumps." Further Analysis of Natural Circulation Tests A. Typical Cenditions The data presented in the March 26, 1964 request was for 122%overpowerconditions. Additional detailed analyses have been per- -formed-recerdly for the specific 84-fuel bundle core -configuration to be i% -_.,4--..

14 operated for the initial Phase II tests. This core' includes 18 develop-mental assemblies and the selective use of BgC poison rods-in 12 of these assemblies.. late following thermal hydraulic and stability data have been s' calculated;for_this core: operating under natural circulation. The limit- -ing case is seen to be based on the steady state burnout ratio: limit of 15 Reactor Thermal Power at Overpower, Mw 190 Reactor Pressure, Psia 1,250 2 Maximum Overpower Heat Flux, Btu /Hr-Ft 305,000 Minimum-Overpower Burnout Ratio 1 51 Fuel Rod Power at Overpower, Kw/Ft .7 08 Inlet Subcooling, Btu /Lb 38.6 Average.InletVelocity,Ft/See 1.75 Maximum Inlet Velocity, Ft/Sec 1.82 Stability Phase Margin at Overpower, Degrees 55.6 6 Recirculation Flow Rate, Lb/Hr 5 0 x 10 Differences in core-power distribution account for the differences between this case and the data previously cited. Whereas the minimum burnout ratio has decreased somewhat, the stability phase margin shows improvement. Each operational mode is evaluated to assure 'that all the applicable limits are met, as is shown for the above case. The analytical method incorporates significant inherent conservatism. The thermal hydraulic analyses are based on methods developed by General Electric Company and have been verified in the operation of many natural and forced circulation reactor systems. The stability studies are -based on methods developed as 'part of the Consumers R&D Program program.(1).and reported in the topical and quarterly reports of that The ournout ratio increases in natural circulation as power decreases, though the recirculation flow also decreases. Tb illustrate this, an example of three calculated power levels is given, including the limiting overpower case, one above, and one below. Flow - Lb/Hr Min B.O.R. 4.9xlOf 1.89 177 Mwt 190 Mwt 5 0 x 106 1*51 206 Mwt 5 1 x 10 1.29 The stability of the 84-bundle core was evaluated for two cases, i.e., that corresponding to the burnout limiting overpower case of 190 Mwt,1250 psia, and its corresponding nominal operating power ~ of -156 Mwt. The results are presented as frequency vs phase and magnitude in the accompanyird figures. (1}GEAP-3795" Consumers Big Rock Point Nuclear Power Reactor Stability Aralysis" by J. M. Case and L. K. Holland. r-- e

15 Power Level Gain Margin Phase Margin 190 Mwt 10.6 Db 55.6 Deg 156 Mwt. 10 9 Db 59 8 Deg ' Both operating. points demonstrate substantial margin and indeed indicate - little~ sensitivity to changes in operating power levels. B. Procedure for Initiating Natural Circulation Tests Phase II tests which involve natural circulation will begin a from a normal forced circulation mode in the following general steps: - 1. Trip both recirculation pumps'from an initial steady operating condi-tion of.157 Mat. 2. Power coast down will settle out at about 90 Mwt, while flow will drop to about 4.1 million pounds per hour. This condition is far from limiting with respect to any. conceivable operating limit. 3. Power will be increased in approximately 20 Mwt steps, while measuring the power level, flow rate, and flux noise amplitude t3 confirm con-tinued observance of specified operating limits before prvceeding to next higher power. h. At selected operating points, Phase II tests such as rod oscillator tests will be performed. 5 On termination of the tests, the forced circulation mode will be re-stored from a reduced power level. C. Recirculation Flow Coast Down Attached is a curve-for a typical case where both recircu-lation -pumps are tripped. The case is the nominal 1250 psia, 157 Mwt initial power situation referred to in the preceding procedure. Shown on the graph are the time responses of heat ' flux, recirculation flow rate, and minimum burnout ratio. These data reflect substantial information gained during the Phase I tests where pump trips were performed from various initial operating levelt and therefore reflect a high confidence level in their adequacy. D. Reactor Operating Limits Reactor operating limits as specified in the Tbchnical -Specifications, Section 8 3, will apply except for.the flow limit given which is applicable only to forced circulation operation. Any natural circulation mode of operation will be shown analytica_ly to be within the following limits for the specific flow rate applicable to the given s = operating condition. The maximum operating level will be that which corresponds to 100% where the calculated limiting conditions are con-sidered as 122% overpower. D -e

L 161 1

Minimum' Overpower Burnout 1 Ratio 15 2

MaximumHeatFluxatOverpower, Btu /Hr-Ft 530,000 2 Maximum Steady State Heat Flux, Btu /Hr-Ft 434,000 Maximum Fuel. Rod Pcwer at Overpower, Kw/Ft 17 2 Maximum Steady State Fuel Rod Power, Kw/Ft .1h.2 Stability Criterion:. Maximum Zero-to-Peak Flux Amplitude, Percent of Average Operating Flux 20 -Maximum Steady State Power Level, Mwt That Permitted by Other Operating Limits Maximum Reactor Pressure During Power Operation, Psig 1485 Set point changes in plant instrumentation are essentially only the picoammeter-set points which will be adjusted to be compatible ' with the new operating. power level dictated by any of the above operating limitations. E. cHazards' Considerations A brief evaluation concerning the varitas aspects of the hazards' analysis and the maneuvering analyses concerned with operation of the teactor on natural circulation has been made. The following statements can be made with respect to the various areas involved: 1. Maximum Credible Accident - The maximum credible accident analyses. (MCOA) conducted for the plant are all based on an initial power level of 240 Mwt and an initial pressure level of 1500 psia. Since the-proposed operation will be at 157 Mwt and 1250 psia, the effects of the MCOA from that operating condition are less severe than the. effects for.the case analyzed and reported previously. The two most important variables in the MCOA analyses are power level and pressure level. 2. Loss of Pumping Power - Since the proposed testing is to be performed at a natural circulation condition there is no need for the usual investigation of the effects of a loss of pumping power accident. The recirculation pumps will be deactivated at a lower power level and the reactor will be brought up in power to the appropriate level while in the natural circulation mode. 3 Plant Transient Performance - The most important parameters which in-fluence plant transient performance are the void reactivity feedback coefficient and the inherent pressure rate following a sharp steam shutoff. The void coefficient was estimated to be 7.8p/%, slightly higher than the 7.4d/% case studied for 240 Mwt operation. However, the pressure rate of the plant for this proposed operating condition is about 37% less than the pressure rate for the 240 Mwt case due to the lower steam flow rate in the proposed condition (610,000 vs 972,000~1b/hr). - Therefore, the' transient response expected from the various turbine trips and load rejections are less than those determined and reported for ' normal operation.

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3 17 Change No. h7. The planned 44-bundle, 60 kv/l core is shown on the . following figure, along.with the proposed source locations. Low power ~ physics tests to be performed on this core arrangement prior to power operation will include removal of fuel surrounding each source to - demonstrate that most of the neutrons being -seen by the start-up instrumentation chambers are fission neutrons. These physics tests will also be utilized to develop control rod patterns which give the required visibility during critical approaches. Although not requested with the ' changes proposed on March 26, 1964, the following proposed changes (designated 48 and 29) are now appropriate based on results'of recent'8h-bundle core low l power. testing. Change -No. h8. In Section 7 3 2(e) delete second sentence. Analysis: Section 51.6 of the Ibehnical Specifications states that the sources sha'l be placed on core coordinates 02-59 and 09-52 with the Bh-bundle core providing that the required 3 eps at a 3:1 signal-to-noise ratio can be ebtained with the sources in this position. Testing performed following loading of the reactor to 84 -bundles has confirmed the fact that adequate count rate and signal-to-noise ratio can be obta' ned-with the sources in these positions, and the reactor is presently operating with the sources in core coordinates 02-59 and 09-52. Since'the visibility requirements presently incorporated in the Technical Specifications under Section 7 3 2(e) were initiated due to the necessity of relocating the sources to a less favorable location for the 56-bundle core, it is now appropriate to delete the second sentence of 7 3 2(e). As discussed with your staff on previous occasions, the visibility re-quirements imposed,even with the sources in the 02-59 and 09-52 positions, force us into control rod patterns which require rod interchanges after criticality is obtained. Also the rod patterns for criticality frequently result in the reactor going critical with higher notch worths than we vould prefer. Consumers Power Company feels that the reactor can be more safely operated if the second~ sentence of Section 7 3 2(e) is deleted. L ' Change No. h9. The 60 kw/ liter core will require relocation of the sources to the core positions as described in' Change No. 47 It'is pro-posed that the following visibility requirement sentence be added to Section 8.2.2.2 as follows: d --c

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'l 18 " Critical approaches with this reduced size core shall _ require that whenever k,ff is greater than 0 970, any increaseinreactivity"f".(wheref='k eff ) greater eff than 0.001 shall result in a fractional increase in count rate ( C ) f not less than 25% of the fractional change - in reactivity

Analysis: Ycni vill note that we propose that a 25% fractional increase requirement be imposed.

It is most evident from recent critical testing that'is is not practical to expect to be able to see a 50% of theoretical response in any thing as complex as a rodded BWR core. It is believed at this time that the 25% requirement can be met without undue ' difficulty; however, we are quite certain that a 50% requirenent would prevent opera - tion since we are also limited by notch worths and rod worths. Yours very truly, REK/v1 Robert E. Kettner Attach. Director of Nuclear Activities 1 1 e

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