ML19338D673
| ML19338D673 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 09/22/1980 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19260G298 | List: |
| References | |
| NUDOCS 8009230640 | |
| Download: ML19338D673 (200) | |
Text
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,.-ENCLOS RU E :1':-
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(- u CALVERT CLIFFS UNIT I I H .- CYCLE-S REFUELING' LICENSE- , .;
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,3 _ _ _ _ _. . - . _ . CALVERT CLIFFS UNIT I CYCLE 5 REFUELItG LICENSE AEFDENT d y Table of Contents-1 Section
- 1. Introduction and Summary
- 2. Operating History of the Reference Cycle !
- 3. General Description
- 4. Fuel System Design
- 5. Nuclear Design 6.- Thermal-Hydraulic Design ~
7.. Transient Analysis
- 8. ECCS Analysis
- 9. Technical Specifications
, 10. Startup Testing
- 11. References
' Appendices A. Asymmetric Steam Generator.Transien stection. Trip Function
- 8. Method for Calculating Space-Time' Scram Reactivities i
C.. ' Description of Model Used:to Simulate NSSS Behavior During Steam Line
' }
Ruptur'e Event, ' D., Prototype Assembly: Description
.E. . ~ Description of Modified Assemblies
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~ 1. INTRODUCTION AND
SUMMARY
This report provides ' an evaluation ' of design and performance for the operation .of Calvert Cliffs Unit I during its fifth fuel cycle at. full rated power of 2700 MWT. All planned operating conditions remain the same as those for Cycle 4. The core will consist of presently operating D, E, and F assemblies, fresh Batch G assemblies and selected C-E/EPRI test pins. Plant operating requirements have created a nead for flexibility in the Cycle 4 termination point, ranging from 10,600 MWD /T to 11,600 MWD /T. In performing analyses of postulated accioents, determining limiting safety settings' and establishing limiting conditions fer operation, limiting values of key parameters were chosen to assure that expected Cycle 5 conditions are enveloped, provided the Cycle 4 termination point falls within the above burnup range. The evaluations of the reload core characteristics have been examined with respect to the Calvert Cliffs Unit I Cycle 4 safety analysis described in Reference 1, hereafter referred to as the " reference cycle" in this report, unless otherwise noted. This is an appropriate reference cycle because of the similarity' in the basic system characteristics of the two reload coIes. Specific core differences have been accounted for in the present analysis. In all cases, it has been concluded that either the reference cycle analyses envelope the 'new conditions or the- revised analyses' presented herein continue to show acceptable results. Where dictated by variations:from the reference cycle, proposed modifications to the plant Technical Specifications are provided and are justified by the analyses reported herein. e t
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p. I ~ :2. OPERATING ~ HISTORY'0F THE REFERENCE CYCLE- .
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u.. - Calvert' Cliffb Unit' .I ~ is; presently ~ operating in its fourth fuel cycle ' 4 utilizing Batch B,. D, E Land F . fuel ( assemblies. Calvert ! Cliffs Unit I , Cycle.-4. began l operation. on July 10, .1979 and reached full _ power on July
~ 20. ; The LCycle f4;startup testing was: reported to: the NRC in-Reference ~ 2.
The: reactor operated. until: about ' October 15, w'ith all ~ core paramet'ers~ in
. good ~ agreement:with calculational predictions.
After : October 15,a progressive' distortion 'of ' the power distribution' was observed, with : power shifting to = the - lower central region of - the. core. 6 This condition. was related to air ingress intial the primary coolant' through j the make-up water and the .lon demineralizer resin ' flush system. . This-
? ~ produced an- oxidizing condition which resulted in crud deposition in the-
., . upper peripheral region of the core.
d As a precautionary measure to maintain adequate margins, the power level was reduced to.'50% on- November .8. A physics testing program was .l initiated and carried out at various power levels' to develop a better understanding of the mechanisms involved, and - to ensure that the ' fuel performance was not adversely affected. The ~ plant was shutdown on January 15, 1980 ; for - TMI plant modifications. During ' this shutdown, a primary coofantchemicaltreatmentwascompletedaimedatremovingthecrud. g After: returning .to power in February, a series of ' physics measurements was-againi performed. to verify -the sur: cess of the chemical = treatment.
~
1 Reactivity f evel,. l , reactivity : coefficients', - and power distributions indicated - t'h at ' the : core characteristics had. returned to normal.- Since-l Marchf6 uthel plant ' has; operated at full power. . Continued monitoring of the plant} confirmed ithat the crud event did not . result in' any permanent - 1 alterahlon of t any Ohysics parameter c and ;no evidence . of a change in fuel
- performance fromIthe pre-crud l period was !found. From' March to the present,.
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- thU plant $ operation L has been ~;' normal ..' at L essentially: full . power.: (The.
_ i 'M evolution iof' the core- reactivity,- reactivity i . coefficients, _ power - q y '.distribbklonsjand.' peaking l factors have followed .the' calculated' predictions
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It is'. I presently estimated that Cycle 4 will terminate on or about October
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17, 1980..- However,; flexibility ?in this date is necessary because of uncertainties in'- the future station' capacity factor. The~ Cycle 4 -
-- termination : point 'can vary ) between 10,600. MWD /T and .11,600 MWO/T _to
[. accommodate the plant schedule. As. of_learly; September 1980,- the Cycle 4 burnup had reached 10,400 MWD /T. - Initial' criticaliti of Cycle :5 is expected 'to occur on or about ' November - b 28, 1980. I L I i. ! -i ? h r-I i l' , \- a r i L L
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i3.0 GEERUDESCRIPTION
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f Th[ Cycle 35.' core %illf consistJ of .the: number and typesi of. assemblies: and .
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1 fuel batchesLas described in,Tabli 3-1. ?The primarpl change. hoJthe; core,Ein; 3-
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, _ ) Cyclel5 f:is . the j removaliof 11 Batch [B 1 assembly,i71 : Batch D . assemblies,4 and - .
120..! B"atch ~ E . assemblies.1 lThese] assemblies. will: be replaced by140 ' eatch G
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. (3.65). w/o i enrichment) . andl52 Batbh : G/(3.03. w/o enrichment) assemblies.
EThe: 52 : low enrichment . Batch G/ assemblie's contain 8 burnable- poison d .s &. ids
.'per l assembly. -l Figure 3-l' shows ;the fuel management. pattern to be; employ,ed :
~
Lin Cycle;5.; Figure 3-2Lshows the? locations. of the poison pins within 'the' L .. . . ~ _. ~ lattice- of ? the ? Batch iG/ assemblies' and . the fuel' rod ' locations fin the unshimmed O Batch
- G' . - assemblies. . This; pattern .will accommodate. Cycle' 4-termination burnups from 10,600 MWD /T to 11,600~ MWD /T.
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' n The-Cycle.5~ core-loading pattern is 908 . rotationally symmetric. .
That.is, g W if.' {one' quadrant of the core were rotated -900 into its neighboring
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~ quadrant, ;each assembly would; be aligned with -a- similar assembly. . This similaritp fincludes batch type, number of- fuel rods,. initial enrichment
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and s burnup. -It doesi not- include: guide tube sleeves, derronstration rod locations;and empty'CEA tubes for an inconel irradiation experiment Figure 3-3 shows;the beginning.of Cycle 5 assembly burnup distribution for a Cycle; 4 itermination - burnup. of - 11,100 MWD /T. Theinitial.enrichmentiof the fuel assemblies is also-shown in Figure 3-3. '* in t ' + ) > f w
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'l I i -E/EPRI Test' Assembly
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' =i In the. center position of the core, one Batch D assembly will be carried Lto a fourth cycle of irradiation. -Among the standard Batch D
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, fuel' rods 4(3.03 w/o U-235), .15 C-E/EPRI test pins will be inserted during
.. the refueling outage. It is_ planned that these'15 rods will include 8 fuel rods with 4 cycles; df. irradiation, 5 fuel rods with 3 cycles of irradiation and 2 internajly ' pressurized non-fueled rods. These rods will be inspected
. 3-to confirm satisfactory performance prior _to insertion into the carrier assembly. (EONdle 0047) for_ Cycle 5. The pins have been thoroughly examined
)b and-dimensionally characterized at each of the first three refueling outages as part of the'C-E/EPRI Fuel-Performance Evaluation Program. Results of these examinations have been reported in References 3 ano,4. These pins will be plauea_in non-limiting positions within.D047.
,i , 3.2- Inconel ' Irradiation Experiment ?
In' order to further the basic material property data on irradiated E
.Inconel-625 CEA cladding, 37 empty CEA tubes _will be placed within the center guide tubes of selected high flux regions of the Calvert Cliffs-I core at the start of _ Cycle '5. 'T'h6 overall design of the test tubes will be similar to the standard C-E design for the neutron source assembly described
, in Section 3.3.7.5 of the Calvert Cliffs FSAR, Reference 5. Basically, an upper
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N-l'1 end fixture and a lower end cap are welded to an_ empty CEA cladding tube and f, ' j _ appropriate flow' holes introduced to eliminate stressi.s from differential system pressure on the test cladding. One of the three tubes utilizes ' tubing appropriate to 14x14 CEA systems and two utilize Inconel cladding of-the size. for;16x16 systems. The upper end fitting has " ears" which sit on p
- top of the guide post and :the upper guide structure-compresses a spring j-W* loaded _s'upport~to keep the specimens in place. The intent of the program is_ to 'providelsufficient material: for mechanical tests on the fully irradiated w }wA ... b.ladding % 1022'nyt) at an appropriate hot' cell.
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_m e 3.3J SCOUTiDemonstration-Assembly-
'The~ SCOUT demonstration assembly;was described in both
Reference:
6 and the Calvert Cliffs Unit 1 Cycle'4' Refueling License
-Amendment, Reference ~1. _ Cycle 5 will.be the second cycle of irradiation for.this Batch F. test-assembly. Prior to returning to the core for its second cycle, the assembly will receive-a detailed visual examination to-confirm satisfactory fuel' performance.
3.4 PROTOTYPE Demonstration' Assemblies The PROTOTYPE-demonstration assemblies are an extension of the SCOUT demonstration assembly. - Similar fuel designs are included, while the number of fuel rods is. larger to provide a statistical basis for fuel performance determinations. PROTOTYPE is comprised of 4 assemblies, each containing standard'.Calvert Cliffs-I, Batch G, fuel ods and PROTOTYPE test fuel rods. The 4 assemblies will be irradiated in symmetric positions
<\ within the core. A more detailed test description can be found in Appendix 0. -7
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'i TABLE 3-1
-CALVERT CLIFFS UNIT-I CYCLE 5 CORE. LOADING '
L , Batch Total .Tofal Average Initial- Number Number Initial 'Burnup Poison Poison of of
- Assembly.; ' Number of Enrichme~nt EOC 4 = Rods Per' Loadino Poison Fuel
- Designation: Assemblies .wt% lf-235 11,100- LAssembly wt% B4C - Rods-- Rods tD~ l 3.03 31,600 0 0 0 174(2)
E- 48 3.03 20,200 0 0' . 0 '8448' <
~ E/ - L. 4 2.73 21,500 0 0 0 704 F 48 3.03 9,300 0 0 0- " '
8448
'F/
.24 2.73- 13,0^ 0 0 0 4224
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G 40- 3.65 0 0 0 0. 7040-G/ 52 3.03 0 8(I) 3.03 416~ 8736; , TOTALS 21 7 ' 416- 37774' Notes
- (1) Shim B10 concentration equals .02685 gms B10/ inch (2) This assembly also contains two non-fuel pins f
4 L_.- _ . . _ _ -- __m.__._ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ . - ~ _ 2
a m 1 G G G G G/: F FI G- GI F/ G/ E GI G F F/ GI E F F G GI F/ 'F E F E F i G FI G/ E GI E GI El GI GI E F- E F E F G F E F E G/ E GI FI G F/ - GI F F El F F/- D
- LOCATION OF DEMONSTRATION ASSEMBLY-(SCOUT)
~
" LOCATION 0F PR0T0 TYPE ASSEMBLY.
-"* LOCATION ~0F FUEL PINS FROM DISCHARGED BT03 TEST ASSEMBLY 1
- BALTIMORE' o GAS & ELECTRIC CO.' CALVERT CLIFFS UNIT I CYCLE .5 "S"" .
- Nuc r Plant : CORE: MAP? 3-1
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l UNSHIMMED -ASSEMBLY . l
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4 8 POIS0N ROD ASSEMBLY $ l J X X ! X - X i X X i c- X X [ FUELR0D LOCATION POISON R0D LOCATION BALTIMORE- FiSure
- GAS & ELECTRIC CO. -CALVERT CLIFFS UNIT I CYCLE 5 -
-:Calvert Clirfs' : ASSEMBLY FUEL AND OTHER R0D LOCATIONS.
' Nuclear Power Plant-3_2
+ '
t 4 INITIAL' ENRICHMENT W/0 U-235 3.65 3.65-
- . B005 BURNUP1MWDIT)E0C4 = 11,100 MWDIT -0 0 3.65 3.65 3.03 3.03 2.73 0 0 0 11,900 13,300 3.65 3.03 2.73 3.03 3.03- 3.03 0 0 13.500 0~ 20,400 0 3.65 3.03 2.73 3.03 3.03 3.03 3.03 0 7,700 12,500 0- 19,300 8,700 10,800 3.65 3.03 2.73 3.03 3.03 3.03 3.03 3.03 l 0 0 12,500 7,700 19,000 -7,400 22,400 9,400 3.65 2.73 3.03 3.03 3.03 3.03 3.03 2.73 :
- 0 13,600 0 19,300 0 21,100 0 21,400 3.03 3.03 3.03 3.03 3.03 3.03 3.03 3.03 0 0 19,300 7,400 21,200 9,500 18,700 10,100 3.65 0
l 3.03 3.03 3.03 3.03 3.03 3.03- 3.03 2.73 3.65 11,900 20,500 .8,700 22,500 0 18,700 0 13,100 0 , ! 2.73. 3.03 3.03 3.03 2.73 3.03 2.73 3.03
.13,400 0 10,100 9,700 21,400 10,800 12,100 31,600 l
l CALVERT' CLIFFS ONIT-I CYCLE 5 Figu're GA ELE T IC co"- - calvert cliffs . ASSEMBLY AVERAGE BURNUP.-AND INITIAL Nuclear. Power Plant; ENRICHMENT DISTRIBUTION' 3-3
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'FUEE SYSTEM DESIGNH 4.1 : MECHANICAL DESIGif . .
The~mechanicalidesign for the'standar'd Batch G reload fuel'is identical
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a to that.of the standard Batch F fuel used in-Calvert Cliffs 1 and described . E in the reference cycle ' submittal, Reference 1. ' Details 1of the standard Batch D and E fuel design parameters can-be found
~
in References .7 and 8, respectively.
'C-E-has performed analytical predictions of cladding creep-collapse time
. for all- Calvert-Cliffs Unit 1- fuel batches that will be irradiated in Cycle 5 and has concluded that1the collapse resistance of all standard fuel . rodss is ~ sufficient'to preclude collapse during their des'ign lifetime.
This lifetime will not be exceeded' by the Cycle'5 duration (Table 4-1). These analyses _ utilized the CEPAN computer code (Reference 9) and included 'as input conservative -values of internal pressure,1 cladding _ dimensions, cladding temperature and neutron flux. Table 4-1 Minimum E0C5 Batch Collapse Time Exposure ' D >33,500 EFPH 32,450 EFPH E >27,500 26,300 F >27,500 19,200 G >27,500 10,450 The clad collapse information presented above is applicable for the test rods in the SCOUT and PROTOTYPE bundles (see Sections < 3.3 and 3.4, respectively) . There are also a limited number of Batch B test rods that will be inserted into fuel assembly 0047 (see 'Section 3.1). All but one of these rods have previously been measured for cladding diameter change and ovality. These data Save been reviewed, and have demonstrated that there are no regions of high local.ovalities which would indicate the presence of.significant axial gaps in the pellet column. Since fuel densification effects would have been
' complete at the time of measurement,'it is concluded that there is no. ~
possibility of cladding collapse in these rods during Cycle 5. The remaining Batch B-test rod has'been. analyzed for collapse using the CEPAN computer code:
~
and best estimate values of internal pressure, cladding dimensions, cladding-temperature, and neutron flux. The results of this analysis show that .the
~
predicted collapse time exceeds this rod's cumulative operating time thru the end of Cycle 5 (38,850'EFPH).
.The metallurigical requirements of the fuel cladding and the fuel assembly
- structural members for the Batch G fuel are identical to those.of the Batch D, E and F fuel from Cycle 4. Thus, the chemical or_ metallurgical perfonnance of'the Batch G fuel will remain unchanged from the' performance
- of the. Cycle _4 fuel.
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^l 4.2- ' HARDWARE-MODIFICATIONS TO MITIGATE GUIDE TUBE WEAR All standard: fuel assemblies which will be placed in CEA locations in Cycle 5,
~
o with the exception of the Batch D assembly Lthat is going'into the core _ center. location, will have stainless steel sleeves installed-in the guide tubes' to prevent guide tube wear. A. detailed ' discussion of-'the design of the sleeves ' and their effect on reactor ' operation is contained' in Reference 10. ;
- A total of thirty-two-other fuel assemblies which will be placed in CEA locations in Cycle 5.contain modifications,-rather than stainless steel-
-sleeves, to mitigate guide tube wear. Twenty of these assemblies
-(sixteen Batch F and four Batch G) contain the.same modifications as-the ~i sixteen Calvert Cliffs II Batch D' demonstration assemblies (described in Reference 11). .The twelve other assemblies (all Batch G) contain modifi-cations (described in Reference '12) which are very similar to the modifi- '
cations of the Calvert Cliffs.II Batch'D demonstration assemblies. A detailed discussion of these modifications is contained.in Appendix E, 4.3' THERMAL DESIGN Using the FATES fuel evaluation model (Reference 13), the thermal performance of the.various fuel assemblies .(fuel Batches D, E,- F and G) has been evaluated with-respect to prior burnup, the proposed burnup during Cycle 5, their respective fuel characteristics, and expected flux -level during Cycle 5. In addition, three groups of additional pins will be present in Cycle 5 as described in Chapter 3 (PROTOTYPE-pins, SCOUT _ pins and pins from the C-E/EPRI ~ test assembly which are being inserted into assembly D047). These pins'were also examined for thermal performance. The fresh fuel, Batch G, has been~ determined to be the limiting fuel batch with respect to stored energy. 1 (i[
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y f5.0 NUCLEAR DESIGN-
~
5.l~ PHYSICS' CHARACTERISTICS'
'5 . 1 . 11 . Fuel Management-The . Cycle 5 fuel management employs a mixed central region' as described =in Section 3, Figure 3-l'. The fresh Batch' G is comprised of two ' sets of assemblies, each having a' unique enrichment in order to minimize radial power . peaking. There are 40 assemblies with an enrichment of 3.65 wt% U-235 and 52 assemblies with an enrichment of 3.03 wt% U-235 and 8 poison shims per assembly. With this ' loading, the Cycle 5 burnup ' capacity for full power operation is expected to be between 13,100 MWD /T and 13,800 . MWD /T, depending on the final-Cycle-4 termination point. The Cycle 5 core characteristics have been examined for - Cycle. 4 terminations between 10,600 and 11,600 MWD /T and limiting values established for the safety analyses. The loading ' pattern (see Section 3) is applicable to any. Cycle 4 termination point between-the stated extremes.
Physics characteristics including ' reactivity coefficients for Cycle 5 are listed in Table 5-1 along with the corresponding values from the reference . cycle. Please note that the values of parameters actually employed. In safety analyses are different from those displayeo in Table '.5-1 and are typically chosen to conservatively bound predicted values with accommodation for appropriate uncertainties and allowances. Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for Cycle 5 with- a comparison to reference
= cycle data.- Trable. 5-2 -generally _ characterizes the changes in ;
. reactivity . that' occur during a trip from full . power with a corresponding. change :In core parameters to the zero power state. It
.is not~1ntended:to-represent any particular limiting A00.or accident, lalthough' theiquantity shown 'as " Required Shutdown -Margin" represents the numerical ~ value .of. the warth which is applied to the hot zero y power steam - lineL break: accident. For the analysis of' any specific a'
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-accident or A00, conservative.or-" limiting" values are used. The CEA
~
groupfidentificat1ons remain the same. as in the reference cycle. ~The power -dependent = insertion :-limit (PDIL) curve is the same as for the L reference cycle. JTable 5-3 shows. the freactivity ~ worths of ivarious CEA' groups calculated at full. power conditions for Cycle 5 L and the reference cycle.
' 5. l ~. 2 Power Distribution Figur'es ~ 5-l ' through 5-3 illustrate the all rods out (ARG) planar radial ~ power distributions. at 80CS, MOCS and EOC5 that -are characteristic - of the high-.burnup end of the Cycle 4 shutdown window. These planar radial power peaks are characteristic of the major portion of the active core length between about 20 and 80
-percent of the fuel height. The higher burnup end of Cycle 4 shutdown window tends to increase the power peaking in this axial central region of the core for- Cycle 5.
The planar radial power distributions for the above region with CEA Group 5 fully inserted at begin'ning and end of Cycle 5 are shown in Figures 5-4 and 5-5, respectively, for the .high burnup end of the Cycle.4 shutdown window. The maximum planar radial ' pin peak of 1.48 occurs at' beginning of cycle and decreases over the cycle. It is characteristic of both' AR0 and Bank 5 inserted conditions that the Cycle 5 peaks are highest near BOC. TheJ radial power distributions described in .this section are calculated data without uncertainties or other allowances. However, single. rod power' peaking values-do include the increased peaking that is characteristic of fuel- rods adjoining the water holes in the fuel assembly lattice. For both DNB and kw/ft safety and setooint analyses -in either . rodded or' unrodded configurations, the power peaking- values 'actually used are higher - than those expected to occur at any time during- Cycle 5. - These conservative- values, which are fused in Section .7 of this ~ document, establish the. allowable limits for power peaking to be observed during~ operation. u -
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i The2 range 5ofi allowableiaxialI peaking - - is Ud sfined L by theElimitingi conditions 9for? operationicovering Jaxial ~ shape index' (ASI):: ;Within. Lthese; ASI: limits,Sthe:necessary DNOR andlkw/ft margins are' maintained - , 5 for ? Ea: - wide range 1 Jofy lpossible , axial. shapes. The' maximum ! tthree-dimensional: lori total E peaking . Jfactor - anticipated iin Cycle . 5' _ during';, normal" base load,lallirods 1 out Loperation at D fu11. power - is ,
.l.81,v not including uncertainty' allowances and augmentation. factors. ,'
' 5.1.3 Safety'Relatied Data
-5.1.3.1IEjectedCEA The' maximum . reactivity- worths. and planar radial ~ power peaks associat'ed with .an ejected CEA event are shown. in. Table 5-4 for Cycle l
- 5 and1. the - reference . cycle. .These ~ values encompass the worst conditions anticipated. during Cycle. 5 for any . expected -Cycle 4
; termination ; point. JThe values' shown for Cycle ~ 5 are - the safety -
snalyses values' which are conssrvative with . respect to the actual
= calculated values. .
5.1.3.2 Dropped CEA'
. The limiting fparameters ?of dropped lCEA reactivity worth and maximum
-increase'in radial peaking factor are =shown in Table 5-5 for Cycle 5 and the:.referenceL cycle..: ~ These : data have : been calculated with the.~
pointwise -Dopplerf feedback technique : described. in Peference . 7. . This
~
- treatment Jis. consistentiwith the. safety analysis since the time; to i
minimum -DNBRLis.:on the - order of a l few minutes, allowing ample time
.for fuel temperatureiredistribution. foll' wing o the CEA drop.. The:
-values (shown1.forn Cycle ' 5 are the L safety _ analysis values, which are conservativeD with respect'to the. actual' calculated. values.
5;1;3.3 Augmentation Factors e , MAugmentationTfactorsLhave?been. calculated for the' Cycle- 5 core using fthe calculational'modellde' scribed in Reference 13. :The input informa-- '
% JtionEre,quiredOforithej calculation 1of augmentation factors that (is --
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- densification) characteristics, the radial pin power distributions and .,
theEsingle gap peaking' factors. = Augmentation factors for- the _ Cycle 15
. core have :been Lconservatively. calculated by combining for input -the largesti single- gapf peaking factors' with the ;most conservative
-(flattest). Tradial pin J power distribution. The_1 calculations yield non-collapsedcclad iaugmentation ifactors;_ showing 'a' maximum value. of
. l.055 'ati the top of the core. The' augmentation _ factors. for. Cycle _5 are ' compared:to the: reference cycle -values l calculated with the same - ,
model in Table 5-6. 1 5.2 ANALYTICAL INouT:TO INaCORE MEASUREMENTS, In-core detector measurement ' constants to be used in- evaluating- the-
~
reload cycle power -distributions will be calculated in the manner
~ described in Reference 14, which. is the 'same method used - for the reference cycle.
5.3- NUCLEAR DESIGN METH000u0GY
.MostL analyses t have been performed in the. same manner and with the same methodologies used for 'the reference cycle analyses. The only analytical: tooliwhich has been changed is the use of space-time scram insertions (FIESTA, Appendix 8).
^
'5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties which are applied to Cycle 5 are: .
~ ~
_Fq =t 7.0 percent ~ where Fq =-F_ X Fz , local power density-Fr.= 6.0 percent'.' These ' values. are ! to - be , used .for' monitoring power distribution
. parameters.during, operation.:
n , L. ~. - . __ _ -_ . _
n -
.y . -
; p- .
~
TABLE 5-l' CALVERT CLIFFS UNIT I CYCLE-5'
-NOMINAL PHYSICS CHARACTERISTICS
. UNITS REFERENCE CYCLE CYCLE 5 Dissolved Baron.
-Dissolved Baron content for Criticality, CEAs.
. Withdrawn Hot Full Pow =r, PPM- 1030 1010 Equilibrium: Xenon,80C Baron Worth Hot Full Paa r BOC. PPM /% Ap 95 101 Hot Full Power EOC -PPM /%Ap 83 83 Reactivity Coefficients
~
-(CEAs-Withdrawn)
Moderator. Temperature-
-Coefficients, Hot Full Power, Equilibrium Xenon lBeginning of Cycle 10-4ao /oF -0.4 -0.1 End of' Cycle . .10-4 40 /oF -2.1 -1.9 Doopler Coefficient Hot Zero Power BOC 10-5ap foF -1.50 -1.55 Hot Full Power BOC 10-5 Ap foF -1.20 -1.21
Hot Full PcLer EOC~
i 10-5 ap foF . -1. 37 . -1.40
. Total-Delayed Neutron-Fraction,Beff 80C .00616 .00628 EOC~ .00522 .00521 i Neutron Generation Time,f *- '
- 80C: 10-6sec 24.8 24.4
- E0C i- 10-6sec- 29.1 '29.7 L.:
l-- Nvj s '
TABLE 5-2
'CALVERT CLIFFS UNIT'I CYCLE 5-4 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES, %
BOC EOC Reference Cycle Cycle 5- ' Reference Cycle Cycle 5 Worth Available Worth of all CEAs 8.9 9.0 9.4 9.9 inserted Stuck CEA allowance 1.8 2.0 1.7 2.2 Worth of all CEAs 7.1 7.0 7.7 7.7 less highest worth
- CEA stuck out Worth Required (Allowances)
Power defect, HFP to HZP 1.6 1.7 2.4 2.2 (Doppler,T ,. worth loss,rediN$1bution) Moderator voids 0.0 0.0 0.1 0.1-CEA bite', boron deadband 0.5 0.5 0.5 0.5
- and maneuvering band
~
Required ShutdownM 'argin. 3.4 4.3 3.4 4.3 Total reactivity 5.5 6.5 6.4 7.1
- Available' Worth Less Allowances Margin available '1.6 0.5 1.3 0.6 F
1 -. - ---
scq. e s
- TABLE 5 'CALVERT CLIFFS UNIT-I CYCLE 5'
- REACTIVITY WORTH OF-CEA' REGULATING
~ CROUPSATHOTFULL. POWER,%Ag.
Beoinnino of Cycle End of Cycle Regulating Reference- Reference CEAs Cycle --Cycle 5 Cycle- Cycle 5 Group.5 0.64- 0.49 0.68 0.57
- Group'4- 0.29 0.32 0.39 0.39
. Group 3. 0.77- 0.97.- 0.88 0.93 Note Values shown assume: sequential'grcup insertion.
1
)
% i
y ; t, s
' TABLE 5-4 -
'CALVERT CLIFFS UNIT I CYCLE 5 =
.CEA EJECTION DATA-LIMITING VALUES Reference Cycle : Cycle ~5
. Safety' Analysis Value : Safety Analysis Value Power Peak' Full Power with Bank 5- 3.36 '3.6 inserted;: worst CEA ejected Zero Power with. 9.83 9.4 Banks-5+4+3+2
.-inserted; worst CEA ejected Maximum' Ejected
.CEA Worth (%ap)'
Full power with 0.32 0.31' Bank-5 inserted;- worst CEA, ejected Zero power with 0.60 0.63 Banks 5+4+3+2 inserted; worst
'CEA ejected Notes.
1)- Uncertali: ties and allowances are included in the above data.
'2) The Cycle 5 safety analysis values'are conservative with respect to the: actual Cycle 5 calculated values.
)
- . a .. i
- p. ; n -
, g, my .. -.=. ,---. ,
1
~ - ,
..- .e z r s 7 jn s- ,
> _ -- ' ,c -
- e s
- s?
.t ~ TABLE 5-5~- '
y
- CALVERT CLIFFS UNIT I! CYCLE 5.
. FULL. LENGTH.CEA DROP DATAj b
Limitino Values
- Reference Cycle ~.
. CycleL5L L
- Safety' Analysis Value' ._ Safety Analysis Value
- Minimum Worth (%Ap) .07 .04 -
x Maximum Percent' Increase' 1.16 1.16
-in Radial Peaking Factor Note.
- 1) The Cycle 5.. safety,and]ysis values are. conservative with respect to the
_ actual. Cycle 5; calculated. values. - e 1 4 t G y , 4 s r - r ggj > \
, ,s f
.. .m .
y _~ --._ - -
~
. TABLE 5-6
- CALVERT CLIFFS UNIT I CYCLE-5
= AUGENTATION FACTORS, AND GA* SIZES.
- Reference Cycle Cycle 5 Noncollapsed Noncolla'psed Core' Core- Clad- Clad-
-Height- .
Height . . Augmen.tation~ Gap-Size Augmentation Gap' Size (Percent) (Inches) Factor (Inches) Factor (Inches) 98.5 '134.7 1.049 2.94 1.055 1.74 86.8 ~118.6' 1.045 2.59 1.050 1.54 77.9 106.5 1.042 2.33 1.046 1.38 66.2 90.5 1.037. 1.98 1.041 1.18
~54.4 74.4 1.031 1.64 1.035 0.97 45.6 62.3 1.027 1.38 1.030 0.82 33.8 46.2 1.021 1.04 1.023 0.62 22.1- 30.2' l.015 0.69 1.017 0.41 13.2 18.1 1.010 0.43 1.011 0.26 1.5 2.0 1.001 0.086 1.002 0.05 Notes
- 1) Values are based'on approved model described in Reference 13.
- 2) The values-in this table for Cycle 5 were used in the safety analysis.
+ C. :;; _L
. v
_o l J
.. 'I 0.75 1.01
.l
-- 0. 74 1.00 0.98 1.02 0.99 o
.0.85- 1.07- 1.01 1.19 :0.95 1.20- -
0.84 1.07 1.05 1.22 0.99 1.14 1.13 e .
~
0.74 1.07 1.05 1.19 0.98 1.13 9.90 1.07 1.00- 1.01 1.22 0.% 1. 13 0.88 1.08 0.80 0.98 1.19- 1.00 1.-13 0.89 1.01 0.88 . 0.97 F
~0;75 1.02 .0.95 1.15 0.91 1.09 0.89 1.04 . O.85
. 1;00g
,: X 0.99 1.21- 1.16 1.07 .0.80 0.% 0.87 0.65 NOTEi X MAXIMUM 1-PIN PEAK = 1.46 t
^ Figure GAS EET C Co* . .
CALVERT C{.IFFS UNIT I CYCLE 5 ~ .
^
calvert Ciirt, ' ASSEMBLY RELATIVE POWER : DENSITY AT>BOC,. 5-1 Nuclear Power Plant - EQUILIBRIUM XENON
- , .. z - . ..-
u. 4 4 0.72 '0.91 O.74 0.98 1.03- 0.97 0.93
.0.82 1.10 1.00 1.24 0.93 1.21
, 0.82 1.00 1.01 1.27 0.98 1.09 1.07 X 0.74 1.10 1.01 1.13 0.97- 1.10 0.89 1.04 ) l O.98 1.00. 1.27 0.% 1.19 0.90 1.17 0.84 1 1.03 1.24 0.98 L10 0.91 1.05 0.94 1.03 0.71 ! l 0.97' 1.09 0.89 O.93- 1.17 0.94 1.19 0.94. 0.90 0.93 : 1.'22 : -1.08 1:04'. O.84 1.02 0.96 0.75 NOTE: .X= MAXIMUM 1-PIN PEAK =L44 i GAS CALVERT' CLIFFS UNIT I CYCLE 5 a sure ELE T IC CO ' coivert clirts. ASSEMBLY RELATIVE POWER < DENSITY AT 6 GWD/T' '
. Nuclear- Power Plant. EQUILIBRIUM XENON 5-2
. ,= . - - : .. - . .:
- . = . . +,- -
' ~
l l I l l 0.74- 0.90
-l
-0.77 0.99 1.08 0.97~ 0.92
'O.83 1.13 1.00 1.27 0.93 1.22
-X 0.82- 0.97 0.99 1.26- 0.96 1.04 1.02 0.77- 1.13 0.99 1.07 0.95 1.06 0.88' 1.01 0.99 1.00 1.26 0.94 1.21 0.91 1.20 0.86' 1.08 -1.27 0.96 1.06 0.91 1.04 0.95 1.03 0.74 0.97 0.93 1.04 0.88 -1.21 0.95 1.24 0.97 I 0.89
- l. 0.92 1.22 1.03 1.01. 0. 86 -- 1.03 0.98 0.80.
L
~
. NOTE: X MAXIMUM 1-PIN PEAK =1.42
[ l-l BALTIMORE' 1CALVER f: CLIFFS UNIT I CYCLE 5 ..
. Figure -l GAS & ELECTRIC CO.-
LASSEMBLY RELATIVE' POWER DENSITY AT E0C' .I
- Nuc r w sh EQUILIBRIUM XENON 5-3 I V ' -
f-
~
0.75 .1.00
-0.74 1.02 1.00 1.00 0.94 ,
0.71 1.03 1.02 1.21 0.91 l0.93l/
////
0.71
/
' 0.76
//' O.99 1.25 1.03 1.16 1.11-r y/
/ X L 0.74 1.03 0.99 1.18 1.03 1.20 0. % 1. 12
- -1.01- 1.02 1.25- 1.03 1.22 0.96 1.17 0.86 5
n 1.00 1.21 1.03 1.20 0.96 1.09- 0.94 1.02 0.75 0.99 0.91 1.16' O.96 1.17- 0.94 1.06 0.82 4 O. 93 d 1.13 1.12 0.85 1.00 0.83 /
/' // I// /
, NOTE: X MAXIMUM 1-PIN PEAK =1.48 i ' /'j// / CEA BANK 5 . j/z LOCATIONS L LCALVERT CLIFFS UNIT I CYCLE 5 Figure .
' GAS E .E T C CO- -
! tcalvert cliff, : ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5' LNuclear Power! Plant - INSERTED, HFP, BOC' '5-4 l
y - - - a f ^ 0.74 0.90.
.1 0.78 '1.02 1.10- 0.95 0.86' 0.72- 1.11 .1.04 1.28 0.88-0.71
/
' ///
0.95 1.29 1.00 1.04 0.98-A.707
//// X T
0.78 1.11 0.95 1.08 1.01 1.13 0.93 1.05 o 1.02 1.04- 1.29 1.00 1.28 0.98 1.27 0.92 1.10 1.28 1.00 1.13 0.98 1.11 1.00 1.07 0.74 y 0. 95 0.88 1.04 0.93 1.27 1.00 1.20 0.89 1
' 0.89' 0.86-
/ //
f .89 0 0.99 1.05 0.91 1.06 0.91
'////
0.45 c-
/b y/
/ /
E NOTE: X MAXIMUM 1-PIN PEAK =1.45
. CEA BANK 5 LOCATIONS a
i I CALVERT CLIFFS UNIT I CYCLE 5 Figure h ' GAS ELE T IC CO . , 1corvert cliffs- ASSEMBLY RELATIVE POWER DENSITY WITH BANK 5
- Nuclear Power Plant- -INSERTED, .HFP, E0C- 5-l,
y . g _ y e . <
~ -- ~
6.0.iTHERMAL HYDRAULIC DESIGN
.m 6.1 'DNBR Analysis Steady state: DNBR analyses:of Cycle' 5 at-the rated power level of
-2700.MWT-have been performed using the TORC computer code described in Reference 15, the CE-1 critical heat flux ' correlation described in Reference 16, and the simplified modeling methods described in-Reference 17J
~
A variant of TORC l called CETOP, optimized for-simplified modeling
~
applications, 'was used in this. cycle to develop the " design . thermal margin - model" described generically in Reference -17. Details of CETOP are discussed in References .18 and 19. . In ger.eral, this, code differs from earlier versions of TORC only in that enthalpy transport coefficients-are used to improve mod _eling of. coolant ' conditions in the-vicinity of the tot subchannel,o and in that more_ rapid equation-solving routines are used. Direct' comparisons show that CETOP models tend to be slightly
.more conservative _than TORC design models in computing minimum DNBR
.for limiting cases. (Note that application of the methods of Reference -17 assures that design models set up with either TORC or CETOP are always conservative relative to detailed TORC analyses).
CETOP is used-only-because it reduces computer costs significa'ntly; no margin gain is expected or taken credit -for. Table.6-1.contains a list of portinent thermal-hydraulic design parameters used for both safety analyses and for generating reactor protective system setpoint 'information. Also note that the calculational factors (engineering heat flux factor, engineering factor on hot channel heat input and rod pitch, bowing and clad diameter factor) listed in Table 6-1 have been combined statistically with other uncertainty factors at the - 95/95 confidence / probability. level (Ref. 20) to define a new design limit on CE-1 minimum DNBR when iterating ~on power as~ discussed in Ref. 20. ' The statistical penalty factors were not applied in all transients; some. transient analyses were done using deterministic methods.
.For this reason, two columns of T-H data are listed 'for Cycle 5.
. Investigations have been made to ascertain the effect of the CEA guide tube wear problem and the sleeving repair on DNBR margins as
' established by this type of analysis. The findings were reported to the NRC in Reference-10 which concluded that the wear problem and the sleeving repair do not adversely affect DNBR margin ~.
6.2 Effects of. Fuel Rod Bowing _on DNBR Margin Effects of fuel rod bowing on DNBR margin have been incorporated in the safety and'setpoint analyses in the same manner as discussed in: Reference 21. Thir reference,contains penalties on minimum DNBR
~ due' to fuel Trod bowing as a function of burnup using NRC guidelines- contained"in ~ Reference 22.
h
n. m.
,q JTNBL:6-1
, Calvert ' Cli ffs' Unit 1 L Thermal-Hydraulic ' Parameters at Full Power Reference - - Cycle 5
- General Characteristicst ,
' Unit: ' Cycle 4*' Deterministic
- Statistical **-
Tctal Heat Output (cere only). ~ 2700-- '2700' 2700-K!g 10 - ' Btu /hr -92151 9215 9215.
~ Fraction of-Heat Generated .in - .975 :.975- .975 Fuel. Rod Primary System Pressure llominal psia- 2250- 2250- 2250-Minimum in' steady state- psia 2200 2200
- Maximum in . steady state psia 2300 2300
~ Inlet Temperature *F 550 550 548 Total Reactor _ Coolant Flow- gpm 370,000 370,000 381,600 (steady. state) 106 lb/hr 139.0 139.1 143.8
' Coolant . Flovi Through Core 106 .lb/hr 135.3- 133.9 138.5 I!ydraulic Diameter- .
.ft 0.044 0.044 0.044 '
(nominal: channel) 106 lb/hr-ft 2 2.53 2.61.
~
Average Mass Velicity 2.53 Pressure' Drop. Across Core' psi 10.6 10.4 11.1
. (minimum steady state flow irreveNible op over entire fuel assembly)
Total Pressure Drop Across-Vessel psi- 32.6 32.4 .34.4 . (b sed on nominal ' dimensions I and minimum steady state ' flow). 2-
~
Core Average Heat Flux ~(accounts BTU /hr-ft ~181,200 186,435*** 186,435*** for above fraction of heat generated Lin fuel' rod- and axial-densificatio'n factor): 2 48,192***
- Total lleat: Transfer Area -(accounts.- ft 49,600 48,192***
- for' axial densification- factor)
Film Coefficient at ~ Aver $ge Conditions Btu /hr-ft2: 5815 5765 5930 F Average Film Temperature Difference *F 31 32 '31 3 Average. Linear Heat- Rate of- kw/ft. 6.05 6.45*** 6. 45***
--Undensifled Fuel' Rod _ (accounts , for above' Traction .of. heat generated iin ' fuel: rod)-
- Average: Core Enthalpy . Rise- Btu /lb .68 68.8 66.5 Maximum Clad Surface- Temperature *F' 657-- 657 657; T
h. 1 I \ a ?~. ' _ _A u
- p. .- -
y - -
- x. .
~
k-g y m Hg Table _6-l'(cont.)- 1;d Calculationalt Factors Reference P Cycle 4 Cycle 5 m,
' $ Engineering Heat Flux Factor - 1.03L 1. 03 '++
L l . 02 + , ++ [ Engineering Factor'on Hot Channel Heat Input
- J1.03 -
Inlet Plenum Nonu'niform Distribution ~ 1.05 '-Not- Applicable Rod Pitch,' Bowing'and Clad Diameter- 11.065. 1.065++ Fuel .Densification Factor-(axial)- u 1 . 01 ' 1. 01. - v
- 41. ..
~ ' ~ ~ ~' ~ ~ ' ~ ~ ~ " '
NOTES <
~~ '
r
- Design'^ inlet temperature and nominal' prirnary system pressure were uses to calculate g- .these parameters.
**Due to the statistical combination of uncertainties described in Refs. 20, 23, and--24, the nominal: inlet. temperature and nominal primary system pressure were used to ' calculate some of 'these parameters.
~
*** Based on 1100~ shims. m ,i
+ Based on "as-built" information' .
++ These factors have been combined statistically with other uncertainty factors at:95/951 confidence / probability level' (Ref. 20) to ' define a new design limit
. cn CE-l' minimum DNBR when-iterating on power as ' discussed in Reference 20.
4 . r b'
> 4 kI 4 i m 5
~,
A-
,J --
7.0 TRANSIENT ANALYSIS-The ' purpose.of this'sectionLis to present the results of the Baltimore Gas & Electric Calvert.. Cliffs. Unit 1, Cycle 5 Non-LOCA safety analysis at 2700'MWt..
- The Design Bases' Events (DBEs) considered in.the safety analyses are
' listed in Table 7-1. These events can be categorized in the following groups:
- 1. Anticipated Operational'0ccurrences for which the. intervention of the Reactor Protective System (RPS) is necessary to prevent exceeding Acceptable Limits.
- 2. Anticipated Operational Occurrences for which the intervention of the RPS trips and/or initial steady state thermal margin, maintained by Limiting Conditions for Operation (LCO), are necessary to prevent exceeding Acceptable Limits.
- 3. Postulated Accidents.
For.'all DBEs so indicated in Table 7-1, an explicit analysis was performed
~
to determine the consequences of these events. Reanalysis of the indicated events was required for one or more of the following reasons: 1.- changes in kay t' parameters due-to eighteen month fuel cycles which. adversely ....ce transient behavior and which were not bounded. by the' reference .jcle,
- 2. changes in methodology. described in References la, lb and ic.
- 3. -. post-Till requirements to manually trip Reactor Coolant Pumps (RCP's) following Safety Injection-Actuation Signal on Low Pressurizer Pressure and Automatic. Initiation of Auxiliary Feedwater Flow on low steam generator l level indication.
- Core parameters input to the safety analyses for evaluating approaches to DflB and centerline temperature to melt fuel design limits are nresented in Table 7-2.
t
>,9 4
,-.- ~ , .
n' TABLE'7-1 CALVERT CLIFFS UNIT--l: CYCLE 5 DESIGN BASIS EVENTS CONSIDERED'IN NON-LOCA SAFETY ANALYSIS Analysis Status
- 7.1 Anticipated' Operational Occurrences for which intervention of.the RPS-is necessary to prevent exceeding acceptable limits:
7.1.1i Sequential .CEA Group Withdrawal ' Reanalyzed 7.1.2 Boron. Dilution Reanalyzed. 7.1.3 Startup of an Inactive Reactor Coolant Pump Not Reanalyzed 7.1.4 Excess Load Reanalyzed 7.l.5 Loss of Load Reanalyzed 7,i.6 Loss of Feedwater Flow .Not Reanalyzed 7.1.7 Excess Heat Removal due to Feedwater Malfunction Not Reanalyzed 7.1.8. Reactor Coolant System Depressurization Not Reanalyzed 7.2 - Anticipated Operational Occurrences for which RPS. trips and/or sufficient initial- steady state thermal margin, maintained by,the LCOs, are necessary to prevent exceeding
.the acceptable limits:
7.2.1 Loss of Coolant Flow Reanalyzed 7.2.2 Loss of AC Power Not Reanalyzed 7.2.3 Full Length CEA Drop- Reanalyzed 7.2.4. Transients Resulting from the Malfunction of One Reanalyzed Steam Generator 2 7.3- Postulated Accidents: 7,1.1 CEA : Ejection. Reanalyzed 7.3.2 Steam'Line Rupture Reanalyzed 7.3.3- Steam Generator Tube Rupture. s Reanalyzed 7.3.4 Seized Rotor ' Reanalyzed
. Requires Low- Flow: trip; event.is discussed in Section 7.2.
2 Requires' trip.on high differsntial steam generator pressure, this trip function
-isidescribed.in: Appendix A.
n , ., it
y V TABLE 7-2 CALVERT CLIFFS UNIT ~1 CYCLE 5 CORE PARAMETERS' INPUT TO SAFETY ANALYSES - FOR'DkB AND CTM:-(CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle.5 Physics Parameters Units Cy(cle CycleValues 4)- Values Radial Peaking Factors-For DNB Margin. Analyses (F ) Unrodded Region 1.'53 - Bank 5-Inserted 1.66 1.65) 1.82
- For Planar Radial Component (Fxy)-
.of 3-D Peak-(CTM Limit-Analyses)
Unrodded' Region 1.61 1.65 Bank 5-Inserted 1.74 1.82),, Maximum Augmentation Factor 1.069 1.055. Moderator Temperature Coefficient 10~4ao/*F -2.5 + +.5 -2.5 + +.5*** Shutdown Margin (Value' assumed %4p -3.4 -4.3 in Limiting EOC Zero Power SLB) ' Tilt Allowance % 3.0 3.0 Safety Parameters. Power Level MWt 2754 2700*,** Maximum Steady State Core Inlet *F 550 548* Tempsrature Minimum Steady State RCS Pressure psia 2200 2225* 6 Reactor Coolant Flow 10 1b/hr 135.3 138.5* Negative Axial. Shape.Index LC0 I .14 .15*,** extreme assumed at Full Power P Maximum CEA. Insertion at Full Power % of Insertion of 25 25 Bank 5 Maximum Initial Linear Heat .KW/ft 16.0 ' 16.0 Rate for-transient other than '
-LOCA-Steady State Linear HeatLRate to Fuel' C2nterline llelt Assumed .in tha Safety KW/ft 21.0 21.0 Analyses CEA;DropyTime from Removal of. sec. 3.1 3.1 Pow;r to Holding Coils to 90%-
Insertion-For 'ONBR calculations, effects of uncertainties-on these parameters were accounted for statistically.
, ** ' For CTM calculations, effects - of uncertainties on these parameters are accounted .for statistically , Numerical = values of these uncertainties and the procedures used in the' statistical: combination.of uncertainties. as- they. pertain
-to :0NB'and CTM limits .are . detailed. in:-Reference' 9.
E(**9The Jeffective initial flTC for!' the LSLB. event is -2.2X10-ke/*F. {:_ _
. 7:
7.1 . ANTICIPATED'0PERATIONAL' OCCURRENCES FOR WHICH Tile RPS ASSURES N0 VIOLATION OF LIMITS _ The ' events;in this categor:y ware' analyzed for Calvert Cliffs Unit 1 Cycle 5 to' determine that Acceptable Limits on DNBR, fuel- Centerline temnerature . to Melt ~(CTM), Reactor. Coolant System (RCS) upset pressure..and 10CFR100 site boundary dose rate guidelines will not be exceeded. Each of the event. writeups-.in the section identifies which criterion the event in question addresses. Protection against exceeding these lidts will continue ,to _ . be assured by the Reactor Protective ~ System (RPS) Limiting Safety System - Settings (LSSS)setpoints. The setpoints will be modified '(as;necessary) to . -
- include changes' necessitated by the results of the analyses ofsthese
.event s.. The methodology. us'ed to generate the Limiting Safety System Settings (LSSS) for the.TM/LP and ASI RPS trips .is ' discussed in CEN-124(B)-P, (Reference la).
For those' events in this section where DNBR or CTM values were calculated and quoted, the calculations were performed using the nominal values of key NSSS parameters-listed in Table 7-2. Uncertainties were accounted forlin determining the values of DNBR or CTH by applying appropriate values of aggregrate uncertainties identified in CEN-124(B)-P to the limiting rod
- power. For'those events analyzed to determine that-the RCS upset' pressure limit or 10CFR100 dose limits are not exceeded, the methods used are the same as previously. reported in the FSAR or subsequent reload lic'ensing submittals. -Effects of NSSS parameter uncertainties = on these limits are not assessed statistically. Instead, applicable uncertainties are assumed to occur simultaneously in the most adverse direction. When values
- of- the NSSS parameters used in evaluation of the RCS pressure and dose limits differ from those given in Table 7-2, they will be specifically noted.
The results of the analyses are provided in the following sections. , o 4
/
d g r .A
- i. . $ I ,
.L L -. ._
^
L . -D 3 . -
's v 4 F
~
,, , e:
'711.1 :CEA WITHDRAWAL EVENT The CEA Withdrawal event was reanalyzed for Cycle '5;to. determine that
!the DNBR:and CTM design limits are not exceeded.
. As stated ,in CENPD-199-P :(Reference 2), the CEA Withdrawal event initiated at. rated.themal power 51s one of the DBEs analyzed to detennine a bias factor used in establishing the TM/LP setpoints. This bias factor,-along-
.with -conservative temperature, pressure, and power trip input' signals assures j that the TM/LP trip prevents the-DNBR from dropping below the-SAFDL' limit-
-(DNBR=1.23 based:on' CE-1 correlation, see-Reference ib) for a CEA Withdrawal event. .Hence, this event was' analyzed for Cycle-5 to generate the bias term
' input to the TM/LP trip.
The'CEA Withdrawal transient may require
- the DNBR and fuel centerline melt (KW/ft)SAFDLs. protection against exceeding Depending both on -the initial conditions and the reactivity insertion rate associated with the CEA ,
withdrawal, either the Variable High Power. Level or Thennal Margin / Low Pressure (TM/LP) trip reacts to prevent exceeding the DNBR SAFDL. An approach to the KW/ft limit is terminated by either the Variable High Power Level trip or the Axial Flux Offset ~ trip. The zero power case was analyzed to demonstrate that SAFDLs are not exceeded. ' For the zero- power case, a reactor trip, initiated-by the Variable High Power trip at 40". of rated thermal power is assumed. - The key parameters for the cases analyzed are reactivity insertion rate due o to rod motion, . moderator temperature feedback effects and initial axial power-distribution. The Resistance Temperature Detector (RTD) response : time is also important in determining the pressure bias factor. '
.The range of reactivity insertion rates considered in the analysis is given in Table 7.1.1-1, along with the values of other key parameters used in the analysis.of this event.
The' maximum _ reactivity insertion rate assumed for Cycle 5'is 1.5X10'4Ao/sec. This ! reactivity withdrawal rate was calculated by combining the maximum CEA differential worth of 3.2X10-4ao/ inch and the maximum CEA Withdrawal speed of_30~ inches per minute. Scram' reactivity versus insertion used in.the analysis of both zero and full powe'r cases corresponds to a bottom peaked shape characterized by a ASI of +.5.. This. power distribution maximizes the time required to
, . terminate the decrease in DNBR following- a. trip.
The CEA Withdrawal transient- initiated at rated thermal power results in the - maximum pressure' bias factor;ofs70.0 psia. This bias factor accounts for R DNB margin' degradation' from the time a reactor. trip is initiated until ninimum DNBR is-reached. This pressure bias factor is used.in generating TM/LP- i
- trip setpoints to prevent the DNB SAFDL from being exceeded. l The zero power; case initiated at the limiting conditions for operation i L ;results in a minimum DNBR of 1.30. Also, the analysis shows that the. .l
' fuel: centerline temperatures are well'below those corresponding to'the-l . - fuel centerline melt SAFDL.
?TheSequenceof~.eventsffor-thezeropower'caseispresentedlinTable7.1.l-2,.
! Figures; 7.1- 1-lj to 7~.1.1-4' oresents the transient behavior _.of. core power,
.,P
~__:_
)~ __
L_
p core averag] heat flux, the RCS ~ pressure.and the RCS temperatures. The analysis of a CEA Withdrawal event presented herein, shows.that the Of48 and fuel centerline melt SAFDLs will not be exceeded during a CEA
- Withdrawal transient.
+
k k I-a c.- 8 J.
. ['i . _ _ . . ,
a
y - 1 , TABLE 7.1.1 ~ KEY PARAMETERS ASSUMED IN THE CEA WITHDRAWAL-ANALYSIS Reference Unit.1 Parameter. Units Cycl e* ' Cycle 5' Initial. Core Power Level -(HZP, HFP) % of 2700'MWt '0, 102 '0, 100+ Core Inlet Coolant Temperature 'F: 532,-550 532, 548+
'(HZP,HFP):'
Reactor Coolant. System ?ressure p:ia 2200 2215+ . Moderator Temperature Coefficient 10~4ap/*F +.5 +.5 Doppler Coefficient Multiplier. .85 . 85
'CEA Worth at Trip - FP 10-26o -5.14 -4.3 CEA Worth at Trip .. ZP. 10~2ao -3.4 _4. 0 '
Reactivity Insertion Rate X10-4Ap/sec 0 to 1.3 0 to 1.6 Holding Coi.1 Delay Time sec 0.5 0.5 CEA Time to 90 Percent Insertion sec 3.1 3.1 (Including Holding Coil Delay)
~
Resistiince Temperature Detector sec 8.0 8.0 Respons t ' Ti.me(T) Rod Group Withdrawal Speed in/ min 30.0 30.0 Maximum CEA' Differential Worth X10~4ac/ inch 2.6 3.2
~
* < Cycle 4 - last. detailed analysis presented (Reference 3).
1+; ForDNBNcalculationsaffectedbyuncertainties,onlythese'parameterswere combined statistically.
.c m
-3 ..
' ~
16 - wha _ _ __.__ . -}
TABLE 7.1'.1
- SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM ZER0' POWER Time (sec) Event Setpoint or Value 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion 39.0 High Power Trip Signal Generated 40% of 2700 MWt 39.4 Trip Breakers Open --
39.9 CEAs Begin to Drop Into Core -- 40.4 Core Power Reaches Maximum 139% of 2700'MWt 41.5 Core Heat Flux Reaches Maximum 65% of 2700 MWt 41.5 Minimum DNBR Occurs 1.30 43.6 Pressurizer Pressure Reaches flaxinun 2380 psia D
,, s
140 . . . i 1 ZERO POWER 120 - - h100 8 Ri. gm - - u 5a. . of 60 - - Wo c_ E o a g - - 20 - - 0 ' J' ' 0 20 40 60 80 100 TIME, SECONDS y
^
GAS CEA WITHDRAWAL EVENT Figure ELE T IC CO. cnive,i Cliih CORE POWER vs TIME 7.1.1-1 Fluclear Power Plant--
i l l 120 , , , , 100 - ZERO POWER - E lE 8 Mg - _ , 8 1 a
$60 - -
x' 5 u_
< 40 - -
W E o u 20 - 0 . 0 20 40 60 80 100 TIME, SECONDS l .
^ '
GAS ELE T IC CO. CEA WITHDRAWAL EVENT sgure O't-<t ai'R CORE HEAT FLUX vs TIME ~ 7.1.1-2
- Nuclear Power Plarit
3% i i i i ZER0 POWER 2450 - -
< 2400 -
0; c Isi 5? 2350 - - O e C - M o
" 2300- - -
2250 - - J 22 % i ' ' ' 0 20 40 60 80 100 TIME, SECONDS
..BALTIM ORE CEA WITHDRAWAL EVENT s9are GAS & ELECTRIC CO..
- cnwn,t citer, RCS PRESSURE vs TIME 7.1.1-3
' tinclear Power Plant L_ -. __ l
580 ,_ , , i ZERO POWER 570
%0 - -
m T E / OUT i2 i I y 550 - - E A Ww lt T AVG I 1 o
- 540 -
I\ , I \ fr's.
.. I vj i '
.530 - -
T IN i 520. ' ' ' ' 0 20 40 60 80 100 TIME, SECONDS l
^
GAS EET C CO. .CEA WITHDRAWAL _ Figure cnivert cliir$
- - Nuclear Power Plant .RCS TEMPERATURES vs TIME 7.1.1 ,
9
~
~7.[.2.BORONDILUTION-EVENT' cThe Boron. Dilution event was reanalyzed.for; Cycle 5 to determine whether <
sufficient time -is available for an-operator to identify the cause and to terminate an approach to criticality for allLsubcritical; modes of ' . operation. - It is also analyzedEto establish corresponding shutdown margin requirements.:for modes ~ 3 through'5 as'they are defined by.the Technical Speci fications~. . 1
- An inadvertent boron. dilution adds. positive. reactivity,. produces power-and: temperature: increases, and during operation at power (for mode 1-and 2).can cause an approach to both the DNBR and CTM limits. .Since the TM/LP , ,
i trip system monitors the transient: behavior of core power level and' core inlet temperature at power ithe TM/LP. trip will intervene, if ncesssary, to_ prevent--the DNBR limit from being exceeded.- For more rapid power ~
. excursions' the Variable High' Power Level; trip initiates -a. reactor trip. ~ The approach to the:CTM limit is terminated .by either the Axial F Jx Offset trip, Variable High Power Level: trip,.or the DNBR related trip discussed above. The trip which is' actuated depends on the rate of reactivity insertion resulting from the dilution event. For~ a boron dilution initiated from p hot.zero power, critical, the power transient resulting from the slow-
. reactivity insertion rate is terminated by' the Variable High Power Level , trip prior.to_ approaching the limits. Table 7.1.2-1~ compares the values of the key transient parameters assumed in each mode.of operation for Cycle 5 and the reference cycle (i.e., Cycle 3).
~
The conservative input data' chosen consists of high critical boron concentrations and lowLinverse baron worths. These choices produce the most adverse effects by reducing the calculated time to criticality. The time to . criticality was determined by using the following expression: 9 C Initial at crit
= T BD X in C
crit where atdrit
=
Time interval to dilute to critical T = Time. constant BD C crit
' Critical boron concentration (PPM)
C '= Initial baron : concentration Initial (PPM) iTable 7 ..- 1 2 2 compares the results- of the analysis Jfor Cycle _5 wi ch.those lfor Cycle'3. The: key-results are the minimum times _-required to lose prescribed _ shutdown margin inleach operational: mode._ As -seen fnxn Table 7.1.2-2,
; sufficient timeLexists for the operator to initiate-appropriate action L to ' mitigate the: ccnsequences of .this Lavent.
__O - --
$P t
I TABLE 7.1.2-1 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Reference.
- Parameter Cycle
- Cycle 5 Critical Boron Concentration, PPM (All-Rods Out, Zero Xenon)
PowerOperation(Mode 1)_ 1300 1800 Startup_(flode 2) 1400 1900 Hot Standby (Mode 3) 1400 1900 Hot Shutdown (Mode 4) 1400 1900
- Cold Shutdown (Mode 5) 1400 1900 Refueling (Mode 6) 1300 1800 Inverse Baron Worth, PPM /%ap Power Operation 70 70 Startup 65 65
' Hot Standby 55 55 Hot Shutdown 55 55
. Cold Shutdown 55 55 Refueling 55 55 Minimum Shutdown Margin Assumed, %Ao Power Operation -- --
Startt.p -3.4 -4.0-
.. Hot Standby. -3.4 -4.0 Hot Shutdown -3.4 -4.0 Cold Shutdown 1.0 -3.0
.-Refueling -7.64 -9.09
~
_ Cycle 3.--last: detailed ' analysis presented (Reference :4).
+.
-TABLE 7.1.2-2'
-RESULTS OF THE BORON DILUTION EVENT
-Criterion For 111nimum
- Time to Lose. Time to Lose Prescribed Shutdown Prescribed Shutdown Mode Marcin (Min) Margin (Min)
Cycle 1 Cycle 5 Startup 87.75 69.8 15 Hot Standby 64.6 59.6 15 Hot Shutdown 64.6 59.6 15 Cold Shutdown 21.0 19.7 15 f Refueling 44.2 38.6 30 - k I,_a
7.1.3_' STARTilP OF AN INACTIVE REACTOR COOLANT PUMP EVENT The Startup 'of 'an Inactive Reactor Coolant Pump event was .not analyzed for Cycle 5 because the Technical .. Specifications do not. permit operation at power.(modes 1 and-2) with less than 4 Reactor Coolant pumps operating. l l-( !. L
17.1.4 EXCESS' LOAD EVENT ,
.The Excess Load Event was reanalyzed to' determine:that the DNBR and CTM design limit are not exceeded during Cycle 5.
The analyses included' the_ effects of manually tripping the .RCP's. on SIAS due-to low pressurizer pressure and the automatic initiation of auxiliary feedwater flow on low steam generator level trip signal. , E The High power level land Thermal _ Margin / Low Pressure (TM/LP) trips provide _ primary protection to prevent exceeding the DNBR limit during this event.
- - Additional protection is provided by other trip signals- including.high rate of change of power, low ~ steam generator. water. level, and low steam generator pressure. In this analysis, credit is.taken only for the. action of the High Power trip in the determination of the minimum transient DNBR.. ,
The approach to the'CTM limit is terminated by either the Axial Flux 0ffset , trip, Variable High. Power Level trip or the DNB related trip discussed above.- The most limiting load increase events at full power and at hot standby ' conditions, for approach to the' DNBR limit of 1.23 (CE-1), are due to 4 the complete opening of the steam dump and bypass valves. For conservatism in the analyses, auxiliary _feedwater flow rate corresponding L to 21% of full power main.feedwater flow was assumed (i.e., 10.5% of full - i power main feedwater flow per generator). Also, the addition of the auxiliary feedwater. to each steam generator was conservatively assumed to occur , 180 seconds after reactor trip. The addition of the auxiliary feedwater flow to both steam generators results in anadditional cooldown of the RCS i and a potential for a_ return-to-power (R-T-P) or criticality arising from reactivity feedback mechanisms. 4 The Excess Load event at full power was initiated at the~ conditions given in Table 7.1.4-1. A Moderator Temperature Coefficient of -2.5X10-4ao/F was assumed in this analysis. This MTC, in conjunction with the decreasing , coolant inlet temperature, enhances the rate of increase of heat flux at the time of reactor trip. A Fuel Temperature Coefficient (FTC) corresponding to beginning of cycle conditions with an uncertainty of 15% was used in the . analysis since this FTC causes therleast amount of negative reactivity
~
change for mitigating the transient increase' in core heat flux. The minimum CEA worth assumed to be available for shutdown at the time of reactor trip: for full power operation is 4.3%Ap. The analysis' conservatively-assumed that the worth of boron injected from the safety injection tank is -1.00%Ap per 105 PPM. The pressurizer pressure control system was assumed to be inoperable 'because. this minimizes. the RCS pressure during the -event and therefore reduces-the calculated DNBR. All other control. systems were , -assumed to be in manual mode of operation and have no impact on the results'ofithis event, a TheE full Power Excess Load event results in a High Power- trip at 7.2 seconds.
~The minimum DNBR calculated for the event at the conditions.specified in Table -7.1;4-l is. l.48 ' compared to the design limit of 1.23. Theiaximum' 1
; local linear heat . generation rate' for the event is 18.1 1Gl/ft. o
'* - r h t )
- ." . . ~ . .- , . . , . , , - .
' F8r the Excess Loa'd : event' initiated from HFP conditions, SIAS .is' generated
- at,34.3 seconds 1at which time the RCP's are manually. tripped by the
, operator.c The coastdown of the ' pumps decreases the rate.of: decay heat removal and.therefore keeps theiRCS coolant temperatures and' pressure'at higher values.
Auxiliary feedwater flow is deliveredito.both steam generators at 187.2 .
. seconds. The feedwater flow causes additional cooldown ~of the RCS. The '
Edecreasing temperatures. in combination with a3 negative MTC . inserts positive reactivity which enables the core to approach criti_cality. -The negative reactivity inserted due to .the CEAs and . Boron: injected -via;the High Pressure Safety Injection '(HPSI) pumps.however is sufficient to maintain the core subcritical at all times.- - Table :7.1.'4-2 presents the. sequence o~f events for an Excess Load event initiated at HFP conditions. Figures 7.1.4-1 to 7.'l.4-5 show the
.NSSS response for power,-heat flux, RCS temperatures, RCS pressure, and steam' generator pressure'during this event.
The Zero Power' Excess Load event was initiated at the conditions given in Table 7.1.4-3.- The-MTC and FTC. values assumed in the analysis are the same as for the full power case for the reasons previously given. r The minimum CEA shutdown worth available is conservatively assumed to be -4.02p. ' The results of the analysis show that a variable high power trip occurs at 35.9 seconds. The minimum DNBR calculated during the event is 2.92 i and the peak linear heat generation rate is 14.1 IGI/ft.
, As with the HFP Excess. Load event, an SIAS signal on low pressurizer pressure. is generated-at 76.6 seconds for the zero power excess load event. - At 215.9 seconds aux ~iliary feedwater flow is delivered to both steam generators. The. additional . positive reactivity due to the cooldown of the RCS is mitigated by the negative- reactivity inserted due to CEA's and the boron injected viu the HPSI pumps. The core remains s'uberitical at all times during.an: Excess Load event initiated from HZP conditions.
The sequence of events for the zero power cas_e'is presented in Table 1 - 7.1.4-4. Figures 7.1.4-6 to -7.1.4-10'show the NSSS response for core power, core heat flux, RCS temperature,'RCS pressure a'nd steam generator p pressure. For the full and zero -power-Excess Load events initiated by a full: opening E of theisteam dump and bypass-valves'the DNBR and CTM limits are not exceeded.: In addition the core remains subcritical even after automatic
~ initiation;of the auxiliary feedwater flow and following manual trip of the RCP's on SIAS.due~to low pressurizer pressure. .The reactivity transient during a HFP and1HZP Excess Load event'is less limiting than the corresponding-Steam!Line Ruoture> events-(See Section 7.3.2).
p , t. f>
~ ,
. I 1 ~ '
2
- . --_ . _ . -- . , e...--. . - - , .
n . ,- - - [ . TABLE'7 l 3-1 KEY PARAMETERS ASSUMED FOR FULL POWER EXCESS LOAD EVENT' ANALYSIS , Parameter Units- Cycle 5 Initial Core Power Level MWt 27CM Core. Inlet Temperature
*F 548+~
Reactor Coolant System Pressure -psia '222f 6 Core M' ass Flow Rate X10 1bm/hr 138.5+
- Moderator Temperature Coefficient. X10-4Ap/ F --2.5 CEA Worth Available at Trip %Aol -4. 3 .-
Doppler Multiplier .85 Inverse Boron Worth PPM /%Ao 105 Auxiliary Feedwater Flow Rate Ibm /sec 175.0/S' G. High Power Level Trip Setpoint- % of Full Power lid ! Low S. G. Water Level Trip Setpoint ft. 30.9 I L i Reference Cycle is FSAR. 1 Full Power Excess Load results were not presented ( in.FSAR, therefore no comparison is made.
~ + For DNBR calculations affected by uncertainties, only these parameters were
- combined statistica11.v.
e L e L:
~
nz . b- ..
TABLE 7.1.4-2 SEQUENCE 0F EVENTS FOR -THE EXCESS LOAD EVENT AT FULL POWER TO CALCULATE MINIMUM DNBR LTime(sec) Event- Setpoint or Value 0.0 . Complete Opening of Steam Dump and -- Bypass Valves at Full Power 7.2 High' Power Trip Signal Generated 110% of full power 7.6 Trip Breakers Open - 8.1 CEA's Begin to Drop Into Core -- 8.6 Maximum Power; 113.2% of full powa Maximum Local Linear Heat
. Rate Occurs, KW/ft 18.1 9.0 - Minimum DNBR Occurs 1.48 10.6 Low Steam Generator Level Trip Setpoint Reached 30.9 ft 34.1 Pressurizer Enpties --
34.3 Safety Injection Actuation Signal Initiated; 1578 psia Manual Trip of RCP's 52.5 Main Steam Isolation Signal 548 psia 60.1 Rampdown of Main Feedwater Flow Completed- 5% of full power main feedwater flow 96.5 Pressurizer Begins to. Refill -- 132.5 Isolation of Main Feedwater Flow to Bnth -- Steam Generators 187.2 Auxiliary Feedwater Flow Delivered to Both 175.0 lbm/r,e: to Steam Generators each steam generator 600.0 Operator Terminates Auxiliary Feedwater -- Flow to Both Steam Generators E 4 w-m *' G
p - - s - TABLE-7 li4-3
. KEY PARAMETERS ASSUMED FOR HOT STANDBY EXCESS LDAD EVENT ANALYSIS .
Reference
- Parameter Units Cycle Cycle 5'
' Initial Core Power'. Level ' MWt- 1 1+ Core . Inlet . Temperature 'F 532 532+
Reactor Coolant System Pressure. psia 2250 2225+ 4 6 Core Mass F1ow' Rate X10 1bm/hr 137.1 141.35+
. Moderator Temperature X10-4ao/ F -2.5 -2.5 Coefficient CEA Worth Available' at' Trip %Ap -2.4 -4.0 Doppler Multiplier
~
.85 .85 Inverse' Baron Worth PPM /%ap 87 100 variable High Power Trip - % of full **
40 Setpoint power Low.S. G. Water Level Trip ft. 30.01 30.9 _Setpoint - Auxiliary Feedwater Flow lbm/sec - 175.0/S. G. + Rate Referer.ce Cycle is FSAR No credit was assumed for Variable High Power Trip; Reactor trip occurred on Low S.~G. level.
+ . For.DNBR calculations affected by uncertainties, only these parameters were
- combined statistically ~ .
, t s
t
'I
TABLE '7.1.4 -4
~
-SEQUENCE 0F EVENTS FOR EXCESS LOAD EVENT AT HOT: STANDBY CONDITIONS TO CALCULATE MINIMUM DNBR Time (sec) Event Setpoint or Value 0.0 ' Steam Dump and Bypass Valves Open to --
_ Maximum Flow Capacity 35.9 Variable High Power Trip Signal Generated 40% of_ full powes 36.3 _ Trip Breakers Open -- 36.9 Core Power Reaches flaximum 40.4% of full 37.6 Minimum DNBR (CE-1) 2.92 72.3 Pressurizer Empties -- 76.6 _ Safety Injection Actuation Signal Generated; 1578 psia Manual Trip of RCS Coolant Pumps -- 82.6 Main Steaa Isolation Signal Generated. -548 psia 88.7 Low Steam Generator _ Water. Level Trip Setpoint 30.9 ft Reached 106.8 Pressurizer Begins to. Refill
'162.6 Isolation of. Main Feedwater Flow to Both --
' Steam Generators 215.9 Auxiliary Feedwater' Flow Delivered 'to 175.0 lbm/sec Both Steam Generators to each steam generator p-
7. 1 TABLE 7.1-4-4 (CONTINUED)
. Time (sec). Event. Setpoint or Value 600.0 Operator Terminates Auxiliary Feedwater ---
Flow'to Both Steam' Generators e D
'f b
,J d
b '. [_,- 3
o l 120 , , , , , FULL POWER 00 [ _ E
- s 80 --
8 N M5
- 60 -- -
of 2 E 40 - - 8 20 - - 0 0 100 200 300 400 500 600 TIME, SECONDS
' EXCESS LOAD INCIDENT Figure GAS $^E E T IC CO.-
cnivert ains . CORE POWER vs TIME 7.1.4-1 n .<.i.,,, e....., et,,ne .
120 i i i i i FULL POWER [ 80 - _ z 8 m 60 - s n
>2 40 -
} _
LE E g 20 - _ W U g -
' ' i , ,
0 0 100 200 300 400 500 600 TIME, SECONDS i
' BALTIMORE GAS & ELECTRIC CO. EXCESS LOAD INCIDENT rigo,,
C HEAT FLUX vs TIME 7.1.4-2 n>cI 7Nw '.!"e'r,..it l L-
n l t l 600 , 1 FULL POWER T OUT
' T 500 . - ---
- ( ._ s
[ AVG -
. , . ' j ~~~
e - T
..'-..L -
- m IN o
Esf 2 . Q300 e E
- s E! TAVG = AVERAGE CORE COOLANT TEMPERATURE 200 - -
TOUT = CORE OUTLET TEMPERATURE TIN = CORE INLET TEMPERATURE 100 - - 0 0 100 200 300 400 500 600 TIME, SECONDS
^
GA5 E E T IC CO. EXCESS LOAD INCIDENT Figure calverrcirik TEMPERATURE vs TIME 7.1.4-3 Mucit ar Power Plant -
r
'eq 2400 , , , , , 1f h FULL POWER'
< 2000 G
- a. .
5?1600 - -
!O E
s '
$1200 -
m 4 5 ' y800 - e
??
o 6 e 400 - 0 ' ' ' ' ' 0 100 200 300 400 500 600 TIME, SECONDS - BALTIMORE GAS & ELECTRIC CO. EXCESS LOAD INCIDENT- Figure coivea Citirs , tluclear Power Plant : REACTOR COOLANT SYSTEM PRESSURE vs TIME 7.1.4-4 l
~'-
i
\
e c T 1000 i , i i i FULL POWER _. D; 800 - a.. isi . 0 - 0 x 600 - - CL
- M j2 m 400 -
s W - u s W m 200 - - e' ' ' ' ' '
'0 0 100 200 300 400 500 600 -
IIME, SECONDS BALTIMORE ^ GAS & ELECTRIC ~CO' EXCESS LOAD INCIDENT- Figure ca m e ci;n e iMAIN STEAM PRESSURE vs TIME 7.1.4 Nucibor. Pow:3r: Plant .
100 , , , , ,
~
HOT STANDBY __ 80 - - Ei! s 8 N 60 - - 8 ci h c. 40 -
/
E f 8 20 - 0 J ' ' ' ' ' ' 0 100 200 300 400 500 600 TIME, SECONDS
?
r r BALTIMORE GAS & ELECTRIC CO. EXCESS LOAD INCIDENT sgure c,i,crt cian CORE POWER vs TIME 7.1.4-6
- Nucle:a Power Plant ,
100 i i i
. i '
HOT STANDBY ~ _ _ ~ 80 - -
- i-E -60 -
8 er a 40 - 25 LL.
$ 20 -
s si k 03 W. 0 100 200 300 400 500 600 o TIME, SECONDS
- BALTUAORE GAS & ELECTRIC CO. EXCESS LOAD INCIDENT Figure:
coivert clius HEAT FLUX vs TIME-
'Nuchor Ppwer Plant 7.1.4-7 n,
700 -- HOT STANDBY . 600 - - T OUT a- .- w 500
- _- T AVG -
m i2
~ w% % s =
g-w ==.~
$ 400 -
W T IN T 300 - AVG = AVERAGE CORE COOLANT TEMPERATURE _ TOUT = CORE OUTLET EMPERATURE TIN = CORE INLET TEMPERATURE 200 ' ' ' ' ' 0 100 200 300 400 500 600 TIME, SECONDS , i I l . GAS ELE T IC CO,= EXCESS LOAD INCIDENT Figure. Cnier,cuik Nuclear Power Plant: TEMPERATURE vs TIME 7.1.4-8
-3
~,
2400 , , , , , HOT STANDBY-
< 2000 - -
Di - CL Ya
$ 1600 g
cu h ' y1500 - - m
!;;E --
5 O 8 800 - - m f2
- a 6
m 400 - 0 i i i i i 0- 100 200 300 400 500 600 TIME, SECONDS
^
GAS E E T IC CO. ' EXCESS LOAD INCIDENT Figure Cnivne Cilih REACTOR COOLANT SYSTEM PRESSURE vs TIME- 7.1.4-9 Pl6cleor. Power Pkint;
1000 , , i i i HOT STANDBY
$800 c.
E' 5? O 600 E 8 2 5 400 - z id
- E
$200 -
0 ' i i i i i 0 100 200 300 400 500 600 TIME, SECONDS i r
^
GAS E .E T CCO. EXCESS LOAD INCIDENT Figure o,ivro crus MAIN STEAM PRESSURE vs TIME 7.1.4-10 Nuc leo Power Phint
wwsm mt m' Qt__ F
;. g wm -
m *; a
'@n , g-*
_E
- 5. m +3; 4; . . ~
y g7 s , g q,u' ~ , >>
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9
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$ 7 ,N* -
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g v;7_- i 3a, - , w . ._- e. .
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7.135kOS3T0F10hDEVENTb 4
~
u s ..
=4 " 5
- m. , ' a ., , . .
.. -me -
'f 2iThe LossTof Loadn event vm$sw[ reanalyzed l fopj Cydle15' to' ' determine thati the - - -
.-; transient {DNBRLdoes not: exceed the;new[ design limit 4and-that:the.RCSL o
e M T
, 1pressurefupsetilimitiof{2750 psih .is not exceeded. !
g t . y _. .. - , .- m p g#./- Theiassumptions7used(to maxim;ize RCS. pressure during thettransient are: o. O a)fThe; event is/assumedNoie' suit'. fromItheisudden closure:of the' tdrbine . N MA- ^ >stopfvalves withoutia. simultaneous" reactor trip. 7This1assumptionicauses: 7the greates.ttreductioniin-theirate"of heat lremoval from the reactor'
- n ,
coolantfsystem and-thus resultsiin'the-most rapid; increase inLprimary +
, fpressureiandethe' closest approachLto the RCS!pressur'e upset limit; s
i 3 b)2Th"esteamTdUmplandibypassTsystem,the.p~ressurizerspray; system,and-R '
?the power; operated pres'surizerir.elief valves are~ assumed not be- '
Loperable.8 Thisitoo maxim'izes"the-primary pressure" reached'during_ ?. .the L_ transient. : ^Y bJ 'The Loss ofl Load event wasiinitiatedIat the conditions shown in Table 7;1.5-1.z Thefcombi_n'ation of' parameters shown intTable 7.1.5-1 maximizes theicalculated peaklRCS pressure. - As can~ be inferred from the table, _ c :theikey. parameters Jfor7this? event' are the Linitial primary and secondary M ' pressures (and the?moderatorJand fuelt temperature coefficients of reactivity. [ 1;The initialicore 'averagelaxial power.. distribution fori this analysis was E
. assumed to ebe a. bottom peaked- shape. ; This~ distribution is assumcd because. "
M it ninimizss thel negative? reactivity inserted 'during the. initial. portion a" .of/theiscram followingta:reactop trip and maximizes;the time required.to
" mitigate the -pressure and. heat fluxLincreases.; The Moderator _ Temperature
.Cdefficientf(RITC);of +LSX10-41o/ F'wasiassuded in this analysis. Thi s'.
i ^ MTC?in: conjunction.mith?theLincreasing_ coolant _ temperatures,-maximizes T -
- the; rateTof; changaiof heat < flux and,the pressure at the tima of reactor
- trip. "A Fuel 1 Temperature Coefficient (FTC) corresponding to beginning: 4
? Lof(cycle conditionsLwas'used;in the; analysis.- This FTC causes the least
~
,s amountzof negative'Treactivity-feedback to mitigate the? transient increases
, in:both the core l heat fluxiandithe' pressure. The uncertainty on the ~
1.. ~ .FTCiusedlin-the?analyseslis?shown in' Table 7.1.5-1. The lower limit on C initialLRCS5 pressure;is used -to. maximize the rateL of change of- pressure, ' a nd (thu sJ pea k ipre s s ure t fol l owi ng ( tri p; '
~
(The)LossofLoadevent,binitiatedfromthecond_itiensgiveninTable7.1.5-i, fresultsfintathighypressurizer.pressurettriptsignalat8.3 seconds..At
~
a J"' - s11.55secon.dsf thelprimarffpressure reaches uits maxicum palue of 2550.0 y t ', psiaMThe ~increa~se. in 'secoddary pressursL is limited by fthe opening of '
- thel main? steam;safetyivalves,ihich:open Jat 3.7 seconds. .The' secondary pr6ssUre! reaches 3its. maximum =value ;of.1050.0 psia at -11.4 seconds after
, _ jinitiationio'f the 'evente - '
1The event?was?als'orreanalyze'd with the' initial conditions' listed in Table- t u ' Wa742- toidetermineithat;the' acceptable!DNBR limit is not ' exceeded. 'The A 3 minimum transientiDNBRfcalculated forsthefevent-:isil.38:as icompared to the
% ,.i ,
fdesign3imitiofil 23 T
, y nun ,
y .
+ -
t
%q lh W * ~~
g[" , - ; + ' yy ' my , , .' , ;(2 [ W] W ^ .,
'1*"'~~
; g .
~
~
Y N $ zh h '1 y ,- Q ; [ ' l M- l-m k , _g
' 0' p;
~l 4
g
,9;
- p 9p;q ' m.;p 4;,L ;p;. ,
- m %w . +
g*w -,fV Lyn; hic k MW
, gu . , u
-qq m ; ' st-u' d m-mm> ya ; c w 39,
+
p y 7 ,'. M9 . MA + , u 4 s ' - " "
p- -
= 8v _
g;
, r.
H,, .,
~$
- Table 7.115-2 prasents the sequence 'of events. for this event. _ Figures
- 7.1.5-1; to 7.1.5-4 show'.the transient behavior of power, heat: flux, .RCS coolant ; temperatures, ~ and RCS - pressure.
15 m
-The results of this ' analysis demonstrates that during a Loss of Load event the . peak RCS pressure and the minimum DilBR do not exceed their respective design limits..
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. .T.ABLE ' 7.1.5 1'
!KEYjPARAME ERS: ASS 0 !!ED'IN. Tile:: LOSS 0F LOAD' ANALYSIS r , a- iT0_ MAXIMIZE! CALCULATED RCS. PEAK PRESSURE
^
, , Reference *- . .
. Parameter _
Units- Cycle -Cycle 5
- Initial Core Power L'evell MWt 2754 ~2754:
- Initialf Core' *Inlet Coolant - *F- 552 550~
Temperature 0 Core Coolarit Flow 'X10 lbm/hr. 133.5. 133.9 Initial Reactor Co'olant ~ ~ psi a '- 2250 2200
; System Pressure
. Initial: Steam Generator psia 875.0 864.0 Pressure. _
Moderator Temperature - X10~4Ap/ F +.5 +.5' Coefficient
. Doppler Coefficient .85 .85
- Multiplier -
- CEA Horth-a't Trip '%Ap- -5.14 -4,7-Time to 90$1 Insertion of- sec 2.5 3.1 Scram Rods-Reactor. Regulating System :0perating Mode Manual Manual Steam Dump.and Bypass-System
. ~0perating liode Automatic Inoperative-Cycle.2 ~1ast detailed' analyses presen'ed t (Reference 5).
e 9 s ,;9 as
.' ~ v
.m -
+ . . ;.-% .
$ TABLE-7.1.5-2 i
SEQUENCE OF EVENTS FOR - THE LOSS OFLLOAD EVENT
- TO MAXIMIZE CALCULATED'RCS PEAK PRESSURE =
Time (sec)' Event Setooint or Value-
?0.'0' - LLoss of Secondary. Load __
3.7 - - Steam Generator Safety Valves-Open. 1000 psia 8.3 High Pressurizer Pressure Trip. 2422 psia - Signal Generated 9.7' CEAs Begin to~ Drop'Into Core -- 9.8' Pressurizer-Safety Valves Open 2500 psia 11.4 Maximum Steam Generator Pressura 1050 psia 11.5 Maximum RCS Pressure 2550 psia 13.4 Pressurizer Safety Valves are Fully 2500 psia Closed O a w a
- h. ' 4 r
120 , i i i 1@ - - h 80 8 m
- 60 I cd i
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- 8-20 - -
0 ' ' ' ' 0 20 40 60 80 100 f TIME, SECONDS BALT!MORE LOSS OF LOAD rigur. GAS & ELECTRIC CO. o,iv,.,i cliis CORE POWER vs TIME- 7.1.5 Nuc. lei <ie Power Plant -
m -- 120 , , , ,
.- . / -
100
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@ 80 - -
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o 0 ' ' ' ' 0- 20 40 60 80 100 TIME, SECONDS l
' f.
~ 1 1
BA LTIMORE . GAS & ELECTRIC CO.~ LOSS OF LOAD Fi 9 "
~
- Cniven Cinh CORE AVERAGE HEAT FLUX vs TIME- 7.1.5-2 Fldeleni. Power Pl<rnt
-l
r, , - 2700 , , , , 5 m C 2500 - - u 2 m m a am _ - 2 W w m 5 2100 - - 5 o o o e R 1900- - - o 6 m 1700 ! 0- 20 40 60 80 100 TIME, SECONDS l l o I 1 i GA5 ELE T IC CO;_ L0SS OF LOAD Figure cnivert cinis-
; Nuclear Power ; Plant . : REACTOR COOLANT SYSTEM PRESSURE vs TIME- 7.1.5-3
630 , i i ' O vi M 610 - T - OUTLET i? E h590 - T AVERAGE
~
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/\
Y f y s s m N M 570
,.N -
5 'N N 8 o
/ +
$ 550 s T o
INLET- ' h-= :- -- E t5 x
-530 I I 0 0 40 60 80 100 EME, SECONDS BALTIMORE GAS E CT ' ~
LOSS OF LOAD R" S c,,Ci tlor.lein Power Plon! REACTOR COOLANT SYSTEM TEMPERATURES vs TIME 7.1.5-4
.~ .
.. 2 ,
./i e .
J
-.7.1.6 JLOSS-0F:FEEDWATER FLOW EVENT
, 1 The Los's;of lFeedwater5 Flow event l;is.' analyze' d ito($erermine- that -
' the-.DNBR ~ limit :and the RCS. pressure upset limit: of.2750 psia:are not : -
exceeded. LIn ' addition, the': event'is: analyzed to: determine ^ that at least
~
'. 10 minutes' exist to blow. dry-the.. steam generators following this' event. .
~
-The Refere'nce cycle (i.e.,1 Cycle' 2, . Reference 5 )T analysis 1shows _that the' DNBR limit'and ; ;the Peak 'RCS pressures ~ obtained.during this event- are -
.less :1.imiting than the:results' of the-Loss of Load, event.
The -Lossiof l.oad ' event' was reanalyzed :for Cycle:- 5 (See ' Section -7.1.5) -
.and the results show- that the DNBR and Peak RCS. Pressures,do not exceed-their respective. design;; limits. Thus,..it can.be' concluded that the' Loss L of Feedwater event, .iff reanalyzed for Cycle 5, Willa not: exceed.the DNBR-and,RCS pressureilimit, u4 M
Hence, it.canbe concluded that the Loss of Feedwater event will not exceed
- the '0NBR limit'and .RCS pressure upset limit and that' the steam generators willinot blow dry during this: event..
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-7.1.7L EXCESS HEAT-P.EMOVAL' DOE TO FEEDWATER MALFUNCTION
- The Loss of High' Pressure Feedwater heaters is the most adverse feedwater malfunction = event in 7tenns of cooling action on the 'RCS. This event-has the same effect on. the primary system as a small- increase in turbine
' demand which-is not matched by an increase in core power. - As a1 result,-
;the'DNBR degradation associated with this event is less severe than for;the Excess' Load event.(See Section 7.1.4) where a larger increase in turbine _ demand'is analyzed. Consequently, this event was not analyzed.
'l
.V _
7 .
.~7.1.8. RCS'DEPRESSURIZATION EVENT.
.i
, l
.The-RCS Depressurization event is' analyzed to determine the pressure bias -factor-input to the,TM/LP tr$p. 'None.of the key transient parameters
- to detemine the pressure bias factor for th'is event are outside the-range:of-the reference. cycle analyses-(Cycle 4, Reference 3). -Hence, the:results~andiconclus' ion reached.in the reference cycle analyses are appl.icable for Cycle 5..
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- J0VERPOWER? MARGIN /AND/0RTRPSiTRIPS:FOR PROTECTION:-AGAINSTiV10LATION f
-x 30F1 LIMITS:
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? The s eventsjinYthis (ca tegoryserela'nalyzed ; for Calvert Cl i ffs ; Uni t (1 ~ - ,
? Cycle" 5 to determine;the!initialimarginstthat must b'eLmaintained by, -
v ithe1 Tech SpecILCO limitsisuchethatl acceptable DNBR,LCTM.and upset: -
~
. tpressureklimitsWill;not3be4 exceeded idu ri ng /any f o f thes e L events .= 1 -
. sJThelini.tialimarginiequiredito: pre' vent? the appropriate limits' from-being ~
%exceededhfor any of Jthe'se. events- wasidetermined using the in_itial:
~
'1
~
!conditionsspe:ifiddfinTable:7-2.= - For each" event ~c'ondition's'.wereichosen to Lassure:that: sufficient;initialfoVerpower margin isiavailable:at the-
!initiatio'n'off the. mostilimiting A00 :i.n' thisteategory. '1he method of:
~ H generating" Limiting Conditions; forj 0peration:-(LCO);isidiscusssd . in . ~
Reference 3cl
.Asinoted:iniSection 7.1,zihitialEconditions'usedTin the evaluation of- -
.'upsetipres'sure" limit?and- dose rate may differ ,from' those given :in
- Table 7-2F since for:these : limits 'the ' effects' of NSSS parameter - ~
- Duncertatnties tare'not combined statistically. ' .
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U 3.i -' (7.2.15? LOSS 10F-COOLANT FLOW EVENTK iTheLloss off Coolantiflow. event lwas. reanalyzed for. Cyclei5 to determine - 7the~ minimum -initial margin :that must-be maintained 'by ~the Limiting _ 6 Conditions Lfor Operationsi(LCOs) such that:in conjunction withithe RPS
- (low 1 flow trip). .the DNBR limit will ? not be exceeded.
"- The methods'used to analyze this event /are -consistent with those, discussed _in Reference 'lc.except CETOP/CE-1 was used instead of-TORC /CE-1
-to~ calculate DNBR The.4-pump loss of? Coolant Flow produces a rapid' approach to the DNBR limit; duef to the rapid decrease.inithe core coolant flow.: Protection -
against exceeding-the DNBR; limit-for this transient is provided by the-initial steady state thermal margin which is~ maintained:by adhering to the -Technical: Specifications ~LCOs on DNB and by"the response off 1 the;RPS;which>provides an .automaticf reactor trip on low reactor ' coolant' flow as measured:by the steam generatorJdifferential pressure transmitters. The : transient is characterized by the flow coastdown curve given in cigure 7.2.1-1. . Table 7.2.1 lists the key- transient parameters used in the present.analy' sis. Table 7.2;1-2 presents'the NSSS and RPS. responses during a four-pump _ loss of flow initiated at a 0.0 shape index. The low flow trip setpoint is reached at 1.00 ' seconds -and the scram rods start dropping-into the core at 2.0 seconds. A~ minimum CE-l'DNBR of 1.23 is reached at 3.00-seconds. Figures 17.2.1-2 to L7.2.1-5 -present' the' core power, heat flux, RCS pressure, and core coolant temperatures as a function of time.
- The event initiated from the Tech Spec LCOs in conjunction with the Low Flow -Trip, +dll ensure that~ DNBR will not be exceeded. -
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SM.EY / PARAMETERS?NSSijMED Ib THE? LOSS-0F!COOLAtCFLOWDANALISLSE *
~ '
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. 7
-< y .:-
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- . U lUnitA '
rReference1 Cycle * " cCyclel5 *-
, , 7 - _
fin'iitallCoreiP'ower Level iMWtc J2/54 w. w < , L27009 JInitialiCoreilnletiCoolanti >
'F.L ;550~ 548+.'
- Temp::raturel > ~
.3 6
3InitialTCoreilahFlow1Ratsi - 1 1bm/hr ... 135.24 .138l5I m -- . R_eactor[CoolantSystemiPress.ure: , 1 psia: # 2200. ?2225L
; Mod le ra t'o r : Tempe ra tu re /Coe f fi hi en t(10 ap/F
~
- +.5. +.5 3 n.n . . . , . ..
~ .. .
TDoppler Coefficient Multiplier . -
- 1.00- 1.00**1
~
JLFT. Response TimeL :sec, .5 .5- >
}CEA Holdin'g Coil 1 Delay sec. 0.5
~
T0.5 LCEA; Time (to:90% Insertion sec. 3.1 ' J(Inclu' ding Holding- CoillDelay) 3.1. i sCE.(Worth ?at:: Trip?(al'l? rods (out) . 2 103ap- -5. 72. :-5.60 ~ LUrirodded Rhdial[Pealiing .Fhetor. 1.58 71. 7+ 1(FJ);~ '
;4-Pump:RCS' Flow Coa'stdown' Figure . 7. 3-1 Figure. ~ " '
_.u 7.2.1;l.
- a .
r_ s ? ,' . l*i cCycle 4? J 11ast . detailed ~ analysis present'edf(Reference 3 ). . WSincettihis s.is 'a' second order effect andethe nos't liniting~ donglerlmultinlier ivariesJduring[theltransient,fi.nominalivalue
! is used. +
j ForTDNBR. calculation's;'ffec' a tid b/_uncertainti.es fonly(tnese parameters were' combined : O d.fstatistically.y~, ". 3;; y / y +
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- TABLE 7.2.1-2 SEQUENCE 0F. EVENTS FOR-LOSS'0F FLOW Time (sec) Event Setpoint or Value 0.0L Loss = of Power to all-Four Reactor -----
. Coolant Pumps
~
l '.'00 . - Low Flow Trip Signal Generated .93% of 4-Pump Flow 1.50 Trip Breakers Open ---- 2.00 Shutdown CEAs Begin to Drop Into Core ---- 3.00 Minimum CE-1 DNBR 1.23
~S.70 Maximum RCS Pressure, psia. 2308.0 Y
b b i
I'l0 4-PUMP'C0ASTDONN
~
O'8
~
0 '6 - 5 Co ~ -
@ 0.4 5
d u - 8 0.2 I I I 1 0 (0 8'.0 - 12"0 16'.0. 20.Q
~
TfME,SEC0NDS BALTIMORE GAS & ELECTRIC CO. LOSS OF COOLANT FLOW EVENT calvert cliff,- 7.2.1-1 Nuclear Power Plant C0hEFLOWFRACTI'ONVSTIME
l
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W a:: - o u- 40
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I i l . 0 0 2 4 6 8 10 TIME, SECONDS A EET CCO. LOSS OF COOLANT FLOW EVENT c=ivert CORE POWER vs TIME 7.2.1-2
- Nuclect Pow; cliffs er Plant )
a -
l i l 120 , , , , iii ~ 2 km .100- -- - 8 b g 80 a. x' 3 u_ - H 60 6 x uJ x-O
- 40 --
20 - i 0- I I I
.I l 0 2 4 6 8 10 l
TIME, SECONDS BAl.TIMORE,...,.e, .
' GAS & ELECTRIC CO.. l.055 0F COOLANT FLOW. EVENT ,
~
caleert cliff 5 CORE HEAT FLUX vs TIME
. Nuclear Power Plant l
P 2500 1 l l 1 2450 - - 5 2400 E u1
~5
$ 2350 -
E c_
.w a
g-2300 - 2250 -
.2200 I I ' '
0 2 4- 6 8 10 i TIME, SECONDS GAS E CCO. l.0SS OF COOLANT FLOW EVENT 7.2.1-4 1 LCalvert cliffs RCS PRESSURE vs TIME-- ., DNuclear P5wer Plant 4
620 i. i i i
'T OUTLET 600 1
E-T vi 580 - AVG - E$ / .- l2 . - Ri
$ 560 W T m IN
- o e 540 - - 520 - - 500- 1 I I I 0 2 '4 6 10 m
. BALTIMORE - -GAS & ELECTRIC CO LOSS OF'C00LANT FLOW EVENT Ccivert clirts RCS TEMPERATURES vs TIME-7'2'1-5 -
l Nuclect Power . Plant.- q _
~
- 7.2.2 ' LOSS OF ALL NON-EMERGENCV A-C POWER EVENT The loss of all _ Non-Emergency._ A-C Power event is analyzed to determine
- that:the~ DNBR limit.and the 10CFR100. site dose-limit will not be exceeded.
For the first few seconds of the transient, the' Loss of all Non-Emergency
'A-C Power behaves-like a Loss of Flow event. .Therefore,-the transient
-DNBR.' calculated' forL the' Loss of Flow event (See-Section 7.2.~1) 'are applicable for this' event also.
However,-none of the key transient parameters in calculating site bocadary-dose limits are more adverse _than the reference cycle analysis (i.e., Cycle
'2,- Reference 5). 1Hence,.the results-and conclusions reached in the '
reference cycle analysis are applicable for_ Cycle 5. l 6 I l !? l~ w M' g f
p - -
~
. 7.2. 3 FULL' LENGTH CEA DROP- EVENT The Full' Length CEA' Drop: event was' reanalyzed f.orLCycle'5.to' determine-the initial thermal margins that must be maintained by the Limiting
- Conditions _for- Operation ;(LCOs); such that' the DNBR' and fuel centerline melt-design limit vill not.be exceeded.
The' methods used ~to analyze-this event are consistent with those-discussed in Reference. Ic ;except CETOP/CE-1 was-instead of TORC /CE-1 to- calculate DNBR. . .
- Table L7.2.3-1 lists' the key input parameters used:for . Cycle 5 and compares them to the reference cycle values. Conservative assumptions used in the analysis. include:
- 1. .The most negative moderator and fuel temperature coeffi~cients of
. reactivity ~(including uncertainties), because these coefficients s produce'the minimum RCS coolant' temperature' decrease upon return to 100% power level and lead to the minimum DNBR.
- 2. Charging pumps and. proportional heater systems are assumed to be
-inoperable during the transient. This maximizes the pressure drop during the. event.
- 3. All other systems are assumed to be in manual mode of operation and' have no impact.on this event.
i-The event is initiated by dropping a full length CEA over a period of-1.0 second. The maximum increases in (integrated and planar) radial peaking factors-in either rodded or unrodded planes were used in all axial regions of the core once tne power returns to the initial level. Values of 16% were assumed for these peak-increases at full-power. The axial power shape in the hot channel is assumed'to remain unchanged and hence the increase
..in the 3-D peak is proportional to'the maximum increase in radial peaking
. factor of 16%. : Since there is no trip assumed, the peaks will stabilize at these asymptotic values after a few minutes since the secondary side continues to demand 100% power. ~
Table.7.2.3-2 presents the sequence of events for the Full Length CEA Drop event. initiated .at the' conditions described in Table 7.2.3-1. The transient behavior of key NSSS parameters are presented in Figures 7.2.3-1 to'7.2.3-4. The transient initiated at the most negative shape index LC0 ( .15) eri at thel maximum power level allowed by the.LCO, results in a minimum
-CE-liDNBR!of 1.23.. A maximum allowable initial linear heat generation
- rate:of'.17.9l KW/ft could ~ exist as an initial. condition:without exceeding 21.'O K!!/ft during this transient. This' amount of margin is assured by setting the' Linear Heat Rate related LCO's based on the,more limiting
- allowableLlinear. heat rate for LOCA. ~
Consequently,11t'is concluded that the Full Length CEA Drop event Linitiated from the Tech Spec'LCOs will not exceed the DNBR and centerline . .to melt' design' limits. _
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TABLE 7.2.3-1 KEY PARAMET$RS ASSUMED IN'THE FULLLLENGTH CEA DROP ANALYSIS Parameter Units Reference Cycle
- Cycle-5
' Initial'. Core Power- Level- MWt 2754 2700+
Core. Inlet Temperature ~. *F 550 548+ Reactor. Coolant 1 System Pressure psia 2200 2225+ 0
- Core ' Mass Flow Rat'e X10 lbm/hr 134'.2 138.5+
Moderator Temperature Coefficient X10-4ao/F -2.5 -2.5 Dopple'r Coefficient Multiplier -- 1.15 1.15 Maximum CEA Insertion at-Allowed : Insertion of 25 25 Power Bank 5 Dropped CEA Worth %Ao unrodded .07 .04 PDIL .04 Most Negative Axial Shape Index .06 .15 Allowed at Full Power (LCO)- Integrated and. Planar Radial linrodded Region 1.16 1.16
- Peaking Distortion Factor Bank Inserted 1.16 1.16 (Full' Power)~ Region Cycle 3 - last' detailed analysis: presented-(Reference 4).
+ For DNBR calculations affected by uncertainties, only these parameters were
- combined statistically.
1 m m
'f 5
th - o. < J. 2.
r;-
- TABLE 7.2.3 '
SEQUENCE OF' EVENTS FOR CEA DROP Time (sec) Event Setooint Value 010 CEA Begins to Drop ---- 1.0 CEA Fully Dropped -0.04%Ao 1.1 Core' Power Reaches Minimum 92.2% 4.2 Core Heat Flux Reaches Minimum 98.1% 250. ' Minimum DNBR Reached. 1.23 300.. Core' Inlet Temperature Reaches Minimum 546.5 F. 300. RCS Pressure Reaches Minimum 2204.3 psia
)
m A
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13 0 - - 100 -
-90 -
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- 60 - -
of ' 50 - l c. E 40 - O o 30 - 20 - - 10 - -
.0 i. I i i i ' ' ' '
0 20 40 60 80 100 120 140 160 180 200- , TIME, SECONDS i
-l BALTIMORE FULL LENGTH CEA DROP- Figure GAS &~ ELECTRIC CO.
Calvert Cirrrs CORE POWER vs TIME - 1 Nuclecr Power Plant 7.2.3-1 l
J 120 , , , , , , , , , 110 - - J 100 g 90 s 80 - - 8 N 70 - - 8
- 60 -
J
>i 5 50 - -
Em - - u 8 30 - -- 20 - - 10 - 1 0 ' ' ' ' ' ' ' ' ' 0- 20 40 60 80 100 120 140 160 180 200 TIME, SECONDS
. ~ BALTIMORE -
GAS & ELECTRIC CO. .. FULL LENGTH CEA DROP Figure C=ivert Cliffs CORE-AVERAGE HEAT FLUX vs TIME
-Nuclecr. Do wer Plant 7'2'3-2
-2250 , , , , , !
l 5 2240 E . E'a p 2230 - - E o .
- E W
p 2220 - - w 5 O g 2210 - 5 E3 6 m 2200 - 2190 ' ' ' ' ' 0 50 100 150 200 250 300 TIME, SECONDS GAS FULL LENGTH CEA DROP Figure E E T IC CO. coIvert cliffs REACTOR COOLANT SYSTEM PRESSURE vs TIME ~7'2,3-3 ~ -Nuclear Power Plant
600 i ' '
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$ 570 -
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~ Ill 540 ' ' ' ' '
0 50 100 150 200 250 200 TIME, SECONDS
^
GAS ET CCO. - FULL LENGTH CEA DROP Figure calvert cliffs REACTOR. COOLANT SYSTEM TEMPERATURES vs TIME 7,2,3-4
. Nuclear Power Plant -
L.
7 . k. i 7.2.4 A00'S RESULTING FROM~THE MALFUNCTION OF ONE STEAM GENERATOR The transients resulting from the. malfunction:of:one steam generator
~
. were: analyzed'for Cycle.5 to determine the Linitial margins that must be maintained-by the LCO's such that in conjunction' with the RPS (Asymmetric Steam Ganerator Protective trip) . the ONBR and fuel centerline melt design limits are 'not exceeded.
The methods Nsed~to-analyze these events'are consistent with those reported
~in..Section 7.2.3 of Reference 2, except that CETOP/CE-1 was used instead of.COSM0/U-3 to calculate the DNBR.. In addition, the Asymmetric Steam Generator. Protective: Trip (ASGPT) replaces low steam generator level trip as- the primary trip to' mitigate this' event. A description of this addition to the RPS is-described in Appendix-A. .
~
.The four events which affect a_ single generator are identified below:
-1.; Loss of Load to One Steam Generator
- 2. Excess Load to One Steam Generator
- 3. Loss of Fee'dwater to One Steam Generator '
~4. Excess Feedwater to One Steam Generator Of-the four events' described above, it-has been determined that the Loss of Load to.One Steam Generator-(LL/lSG) transient is the limiting asynynetric event. Hence,-only the-results of this transient are reported.
~
The event is . initiated -by the inadvertent closure of a-single main steam isolation valve. ~Upon the loss of load _to the single steam generator,
.its. pressure and temperature increase to the opening' pressure of the secondary safety valves. The intact steam generator " picks up" the lost load, which causes ,its temperature.and pressure to decrease. The cold leg asymmetry causes an inlet' temperature tilt which results in an azimuthal power tilt, . increased PLHGR and 'a degraded DNBR.
- The LL/1SG was-initiated at the conditions.given in Table.7.2.4-1.
. A reactor trip is-generated by:the. Asymmetric. Steam Generator Protection Trip at 2.6 seconds based on high differential pressure between the steam generators.
Table 7.2.4-2 presents the sequence of events for the Loss of Load to
-One Steam Generator. The transient behavior of key NSSS parameters are
~
presented i.n Figures' 7.2.'4-1 to 7.2.~4-5. E A maximum ~ allowable initial linear heat generation rate of 19.0 KW/ft could exist ~as an initial condition without exceeding 21.0 KW/ft during
~ this transient. This amount of margin is; assured by setting the
' Linear Heat RateLLCO based .on the more limiting allowable linear heat
[ -rate for'LOCA.--
- The event initiated:fr' om the extremes 'of the LC0 in conjunction with the-2ASGPJ trip will not_. lead. to DNBR or centerline fuel ~ temperatures which j
y l 4 exceed the DNBR and. centerline to; melt ' design limits. L The minimum 1transie~nt 'DNBR calculated for-the LL/lSG event is 1.43 as cr . pared to thelmi.nimum s acceptable;DNBR of:1._23. w
+ '
- e
,mn 1 -
-e
y- _ a: , '
- t. :
' TABLE 17. 2.4 -1 KEY PARAMETERS ASSUMED IN THE ANALYSIS.0F LOSS OF LOAD TO ONE STEAM GENERATOR ~
~
Reference Parameter Units' Cycle
- Cycle 5~
- Initial' Core ~ Power EMWt, 2700 2700+
Initial Core -Inlet. F 552 548+ Temperature
-Initial Reactor. Coolant psia 2200 .222f System Pressure
-4 Moderator Temperature 10 ho/*F -2.5 -2.5 Coefficient Doppler Coefficient - --
0.85 0.85 Multiplier
.This event was not analyzed in the FSAR, but was evaluated in CENPD-199-P (Reference 2). Thus Reference 2 parameters are compared with Cycle 5 parameters.
+ For DNBR calculat ons affected by uncertainties, only these~ parameters were combined statistically.-
l
)
.j
-.e ._- ex a m.
')
-TABLE 7.2.4-2
~ SEQUENCE OF EVENTS FOR LOSS OF LOAD T0,0NE STEAM GENERATOR
-Time (sec) Event Setpoint or-Value 0.0 Spurious closure of a single cain . steam ----
isolation valve 0.0 Steam flow from unaffected steam generator- ---- increases to maintain turbine power - 2.6 ASGPT*- setpoint reached (differential pressure) 175 psid 3.2 Oump and Bypass valves are open ---- 3.5 Trip breakers open ---- 4.0 CEAs begin to insert ---- 4.0 Safety valves open on isolated steam generator 1000 psia
~5.5 Minimum DNBR occurs 1,43 10.1 Maximum steam generator pressure 1050 psia ASGPT - Asymmetric Steam Generator ' Protection Trip
+
N-.- - . _ _ .
.r
- -120 , , , , ,
100 - - sii -
- s 80 -
8 M
# 60 - -
ei W 2 Eid 40 - - 8 . 20 - - 0 0 25 50 75 100 125 150' TIME, SECONDS 1 l , BALTIMORE GAS & ELECTRIC CO. LOSS OF LOAD /1 STEAM GENERATOR EVENT Figure Caivert Ciirrs CORE POWER vs TIME 7,2A-1 L Nuclear Power Plant
l l I 120 i i i i i 100 m - E
- E g M - -
Ri 8-x' 60 - - 3 n.u e g 40 - - o O 20 - 0 ' ' ' ' i 0 25 50 75 100 125 150 TIME, SECONDS BALTIMCRE LOSS OF LOADl1 STEAM GENERATOR EVENT Figure 1 GAS & ELECTRIC CO' I Calvert C!iffs CORE AVERAGE HEAT FLUX vs TIM" Nuclear Power Plant 7l2l4-2
'[
1050l, , , i i i ' r 1000 - - G ! ISOLATED STEAM GENERATOR c. vr 950 - - E is 0 E 900 - -
~
m: D UNIS0 LATED STEAM GENERATOR 8 850 - -
- E 6 :
M 1 800 - - 750 ' ' ' ' 0 25 50 75 100 125 150 TIME, SECONDS GASYEIE T C CC).- LOSS OF LOADl1 STEAM GENERATOR EVENT Figure Ceivert Cliffs STEAM GENERATOR PRESSURE vs TIME 7,2A-3 LNuclear. Power Plant- - j
~
DM i . i i i 2200 - - 2100 - - G CL El 5 2000 - - O E m 1900 - - L 1800 - - 1700 ' ' ' ' ' 0 25- 50 75 100 125 150 TIME, SECONDS
. . BAl.[IMORE LOSS OF LOAD /1 STEAM GENERATOR EVENT rigure GAS & ELECTRIC CO'-
Calvert airr, REACTOR COOLANT SYSTEM PRESSURE vs TIME Nuclecr Power Plan, 7.2A-4
. .. j j
-l i
615 , -, , , . 600 -- -
'0UT .
u. 585 - - d E e 5 570 -
$ AV W
w l5! . - 555 - 540 - - 525 ' ' ' ' 0 25 50 75 100 125 150 TIME, SECONDS
' BALTIMORE L0SS OF LOAD /1 STEAM GENERATOR EVENT rigure-
-cII,. t c!;$,
REACTOR COOLANT SYSTEM TEMPERATURE vs TIME 7,2A-5 Nuclear Power Plant.-
?
7.3 POSTULATED ACCIDENTS- . The events ~in this cat.egorf were analyzed for Calvert C1tffs Unit 1, Cycle 5 to ensure. acceptable consequences. For these transients some amount of
- fuel-failure is acceptable prov; J the predicted site boundary dose meets 10CFR100 guidelines.
The following sections present the results of the analyses. i'
^
- c 7.3.1 CEA EJECTION EVENT
- The CEA Ejection event was reanalyzed for Cycle 5 to determine the fraction of fuel pins. that exceed the criteria for clad damage.
The analytical method employed in the reanalysis of this-event is the NRC approved Combustion Engineering CEA Ejection method which is described in CENPD-190-A, (Reference 6). The key parameters used in this event are listed inLT able 7.3.1-1. With these key parameters, selected.to add conservatism, the procedure outlined in Figure 2.1 of Reference 6 is then used to determine the hverage and centerline enthalpies-in the hottest spot of the rod. The calculated enathlpy values are compared to threshold enthalpy values to determine the amount of fuel exceeding these thresholds. These threshold enti1alpy values are (References 7, 8, and 9). ~
- Clad Damage Threshold
Total Average Enthalpy = 200 cal /gm Incipient Centerline Melting Threshold: Total Centerline Enthalpy = 250 cal /gm Fully Molten Centerline Threshold: Total Centerline Enthalpy = 310 cal /gm To bound the most adverse conditions during the cycle, the most limiting of either the Beginning of Cycle (B0C) or End of Cycle (EOC) parameter values.were used in the analysis.' A BOC Doppler defect was used since it produces the least amount-of negative reactivity feedback to mitigate the transient. . A B0C moderator. temperat..c coefficient of 'C.5X10-4Ao/*F was used because a positive MTC results in rasitive reactivity feedback - and thus increases coolcat temperatures. At E0C ' delayed neutron fraction was used in the analysis to produce the hignes; power rise during the t event. The zero power CEA ejection event was analyzed assuming the core is initially operating at'l MWt.- At zero power, a Variable Overpower trip is conservatively ' assumed to initiate at 40% (30% + 10% uncertainty) of 2700 MWt and terminates the event. The full and zero power cases were analyzed, assuming the value of 0.05 seconds for the total ejection time, which is consistent with the FSAR and previous reload submittals. The pcwer transient produced by a CEA ejection' initiated at the maximum L allowed _ power is shown in Figure 7.3.1-1. l power case-are shown in Figure 7.3.1-2. Similar results for the zero i l l 4 L.m.,.
p = .: The results' of~the two CEA ejection cases analyzed (Table.7.3.1-2) show
.-that the maximum'; total-energy deposited-during the event is-less than
.the criterion for clad' damage-(i.e.,-1 200 cal /gm). Also, an acceptably small fraction of the- fuel reaches the incipient centerline melt threshold.
Consequently, no fuel pin failures occur. O f 4 1 l t
l% , Gih L - m -e , TABLEi7.3 1-1 - KEY. PARAMETEP.S -ASSUMED IN -THE CEA EJECTION ANALYSES . g .- ;. ~ 1- . . Unit 1 Parameter ' Units Reference Cycle * ' Cycle 5 Full Power-Core: Power. Level - MWt- 2754 2754
- Core Average Linear Heat .
KW/ft, 6.12 6.52 Generation Rate'at=2754 MWt. Moderator Temperature 104Ap/*F- +.5 +.5 Coefficient Ejected CEA Worth %Ao .32 .31 Delayed Neutron Fraction,:8 .0047 .0044 Post-Ejected Radial Power Peak' 3.36 3.60 Axial Power Peak- 1.39 1.35 CEA' Bank Worth at Trip: %Ap -3.88 -3.00 Tilt Allowance 1.03 1.03 Doppler Multiplier 0.85 0.85 0.85 Zero Power Core Po'wer. Level MWt 1.0- 1.0 Ejected CEA Worth %Ap .60 .63
. Post-Ejected: Radial' Power. Peak. 9.83 9.40 Axial Power Peak 1.60 1.75
- CEA Bank Worth at Trip %Ap -2.58 -1.50 Tilt Allowance 1.10 - 1.10 CEA' Drop' Time sec 3.1- 3.1
. Doppler liultiplier 0.85 0.85 s
t Cycle L4 last detailed analyses : presented (Reference 3). g j _ g c e p -
~
a.e E y -TABLE 7.3.1 CEA -EJECTION EVENT RESULTS Reference Cycle Unit 1 Full Power - Unit 1, ' Cycle 4 Cycle-5 Total Average Enthalpy of Hottest Fuel Pellet 198. 178.
-(cal /gm)
Total: Centerline Enthalpy of Hottest Fuel' 268. 277. Pellet.(cal /gm) Fraction of Rods.that Suffer Clad Damage 0 0
- (avGrage Enthalpy-3 200 cal /gm) _
~ -Fraction of Fuel.Having a Least Incipient. .01 .04 Centerline Melting (Centerline Enthalpy > 250 cal /gm) ~ ' Fraction of-Fuel Having a Fully Molten Center- 0 0 line Condition (Centerline Enthalpy 1310 cal /gm)
Reference Cycle Unit 1 Zero' Power Unit 1, Cycle 4 Cycle 5 Total Average'Enthalpy of Hottest Fuel Pellet 177. 145 (cal /gm)
. Total Centerline Enthalpy of Hottest ~ Fuel 177. 221 Pellet (cal /gm)
Fraction of Rods that Suffer Clad Damage 0 0 (Average Enthalpy 1 200 cal /gm) Fraction of Fuel Having a Least Ir.cipient 0 0 Centerline Melting (Centerline Enthalpy > 250
-cal /gm) -
Fraction of Fuel Having a Fully Molten Centerline 0 0 Condition (Centerline Enthalpy 3 310 cal /gm) l I l
- 2.
I i .I l FULL POWER 3.0 l E E g- - M
- z. 2.0 - -
o it o u N 3 2 w y 1.0 - _ 0 0 :1. 0 2.0 3.0 4.0 5.O TIME, SECONDS
- BALTIMORE GA.S & ELECTRIC CO.- CEA EJECTION RATE Figure Coivert ci;rf5 CORE POWER vs TIME-Nuclecr: Power Plant - 7.3.1 I i- 1 I i
ZERO POWER 10.0 - g _ _ g
~
~
m _ _ o z S a 1. 0 __ __ E - - 5 3:: o - _ u .. _ a:: o u .
.0.1 .
0.03
'0 -1 2 3 4 :
TIME, SECONDS BALTIMORE GAS & ELECTRIC CO. CEA EJECTION EVENT Figure
- Nuc e r w Plant ~ CORE POWER vs TIME 7.3.1-2
" 1 7.3.2' STEAM LINE RUPTURE-EVENT-The: Steam.Line Rupture (SLB) event was analyzed for Cycle 5 to determine
~
'that the critical. heat. flux is not exceeded during this event.
'The analysis included the effect of automatic initiation of auxiliary
- feedwater flow in three (3) minutes from the -initiation of the event and the manual trip _ on the Reactor Coolant Pumps on Safety Injection Actuation-Signal due to low pressurizer pressure ,
The analysis assumed that the event.is initiated by a circumferential rupture of a 34-inch (inside diameter) steam line at the steam generator 1 main steam line nozzle. This break size is the most limiting, since it causes the greatest rate of temperature reduction in the reactor core region. . The SLB event was analyzed _with the assumption of a three minute delay between the time of reactor trip and the time when Auxiliary Feedwater (AFW) flow is delivered to the-affected steam generator. This is conservative with respect to the. expected time of AFW initiation since the generation of the AFW signal actually occurs at the time of the low steam generator water level trip signal. The analysis assumes, therefore, that AFW flow is delivered to the steam generator sooner than the flow is - actually available resulting in a conservative prediction of the resulting cooldown. A conservatively high value of the AFW flow was calculated assuming that all auxiliary feedwater pumps are operable. An AFW flow value of 21% of full power feedwater flow was used in-the analysis. This value accounts for pump run-out due to reduced back pressure. In addition, the analysis conservatively assumed that all the AFW flow is fed only to the damaged steam generator. The analyses assumed 'that the main feedwater flow is ramped down to 5% of full power feedwater flow in 20 seconds and that the main feedwater isolation valves-are comple.tely closed in 80 seconds after a low steam 1 generator pressure or a main steam isolation signal. These assumptions are consistent with Technical Specification limits (See Table 3/4.3.2). The manual trip of.the RCP's results in no flow mixing at the core inlet plenum. Thus, cold edge temperatures were used to calculate the
~
moderator reactivity insertion during the .cooldown of the RCS following an SLB. .Hence, the mathematical model described in Appendix C was used to simulate the SLB event in combination with manual trip of RCP's and automatic initiation of auxiliary feedwater flow. The' two SLB-cases considered in conjunction with automatic initiation of auxiliary feedwater _ flow and manual trip of RCP's are:
- l. 2 Loop- Full Load'(2754 MWt)
- 2. 2 Loop- -
'No Load (1 MWt)-
Theil-loop-full load and 1 loop-no-load cases were not analyzed since Technical Specifications prohibit operation in these modes. l 2
, e
7 e Two-Loop-2754 MWt The Two Loop-2754 MWt case was initiated at the conditions listed in Table 7.3.2-1. The Moderator Temperature Coefficient (MTC) of reactivity assumed in the analysis corresponds to end of life, since this MTC results .in the greatest positive reactivity change during the RCS cooldown-caused by the Steam Line Rupture. Since the reactivity change associated with moderator feedback varies significantly over the moderator density covered in the analysis,'a curve of reactivity insertion versus density
.rather than a single value of MTC, is assumed in the analysis. The moderator cooldown curve assumed is given in Figure 7.3.2-1. The moderator cooldown curve given in Figure 7.3.2-] was conservatively calculated assuming that on reactor scram, the highest worth Control Element Assembly is stuck in the fully withdrawn position.
The. reactivity defect associated with fuel temperature decreases is also - based on end of life Doppler defect. The Doppler defect based on an end of life Fuel Temperature Coefficient.(FTC), in conjunction with the decreasing fuel temperatures, causes the greatest positive reactivity insertion during the Steam Line Rupture event. The uncertainty on the FTC assumed in the analysis is given in Table 7.3.2-1. The 6 fraction assurred is the maximum absolute value including uncertainties for end of life conditions. -This too is conservative'since it maximizes subcritical multiplication and thus, enhances the potential for Return-To-Power (R-T-P). The minimum CEA worth assumed to be available for shutdown at the time of reactor trip at the maximum allowed power level is 7.15%ao. This available scra.a worth was calculated fo.r the . stuck rod which produced the moderator cooldown curve in Figure 7.3.2-1. The analysis conservatively assumed that on Safety Injection Actuation Signal that one High Pressure Safety Injection pump and one Low Pressure Safety Injection pump fail to start. The analysis also assumed a conservatively low value of boron reactivity worth of -1.0%ao per 105 PPM. The conservative assumptions on feedwater flow were discussed previously. The feedwater flow and enthalpy as a function of time are presented in Figures 7.3.2-2 'and 7.3.2-3 respectively. Table 7.3.2-2 presents the sequence of events for the full power case initiated at the conditions given in Table 7.3.2-1. The reactivity insertion as a function of time is presented in Fim re 7.3..;-4. The response of the hSSS during this event is.given in Figures 7.3.2-5 through 7.3.2-9. The results of the analyses shows that SIAS is actuated at 15.6 seconds, at which time the Reactor Coolant Pumps are manually tripped by the operator. _ The manual trip of RCP's slows down the rate of primary heat removal and thus delays the time when the affected steam generator blows
; dry. The affected steam generator blows dry at 94.5 seconds and terminates the cooldown of the RCS. The peak reactivity attained prior to delivery of auxiliary feedwater ficw is -0.10%ao at 125.5 seconds. A peak R-T-P of 8.2%, consisting of 7.2% decay heat and 1% fission power is produced q at 123.0 seconds. The continued production of decay heat from the fuel 4 after termination of blowdown, causes the reactor. coolant -temperatures to increase. This in turn reduces the magnitude of the positive moderator reactivity inserted and .thus the total reactivity becomes more negative, m _ _ ..
.. u 1
-The delivery of auxiliary- feedwater' flow t'o the affected steam generator at .183.3 seconds -initiates a further cooldown ~of the RCS which results in more positive reactivity. insertion.: The positive .eactivity _ insertion causes theicore to Ereturn to criticality. The peas. criticality attained
~is +0.036%Ao at 482.5 seconds. The reactivity transient is terminated by_the boron injected via the High Pressure Safety Injection Pumps. A peak R-T-P of .4.0%, consisting of 2;7%-decay heat:and 1.5% fission power is produced at 572.5 seconds.
The Steam Line Rupture _ event from HFP conditions with automatic initiation of auxiliary feedwater and manual trip of RCP's on SIAS shows that the core _does not produce'significant instantaneous fission power. Since-there is.no significant Return-To-Power, it can be concluded, as it was presented
-in the FSAR and toe subsequent license submittal, that critical heat fluxes will not be exceeded, a
Two Loop-No Load Two Loop-No Load case was initiated at the conditions given in Table 7.3.2-3. The moderator cooldown curve is given in Figure 7.3.2-10. The cooldown curve corresponds to an end of life MTC. An end of-life FTC was also.used for the reasons previously discussed in connection with the two loop-2754 MWt case. The minimum CEA ' shutdown worth available is conservatively assumed to be the minimum required technica: specification limit of 4.3%Ao. A maximum inverse boron worth of 100 PPM /%Ao was conservatively assumed for the safety injection during the no load case. The feedwater flow and the enthalpy used in the~ analysis are presented in Figures 7.3.2-11 and :7.3.2-12 respectively.
~
Table 7.3.2-4 presents the sequence of events for the Two Loop-No Load case initiated .from the conditions given in Table 7.3.2-3. The reactivity insertion as a function of time is presented in Figure 7.3.2-13. The NSSS responses during this event are given in Figures 7.3.2-14 to 7.3.2-18. The results of the analyses show that SIAS is actuated at 13.6 seconds, at which time the-RCP's are manually tripped by the operator. The affected steam generator blows dry at 180.0 seconds. Auxiliary feedwater flow is initiated at 183.7. seconds which continues the cooldown of the
-RCS. . Thus,-the: total core reactivity approaches criticality. The peak reactivity attained is~+0.075%4p at 422.0 seconds. The addition of boron via High Pressure Safety Injection mitigates the reactivity transient.. However, the core power level never exceeds the ' initial value of 1 Mut at any time during the event. Since there is no-R-T-P for the Two Loop-No. Load case, it can be concluded that the critical i heat _ flux will not be exceeded during this event.
c , i
y - 1.g m o- -
- x. a. " ? , = -
- : o
^
G
.,.- ,. t. , , - ' ' - -
?
f( ~" -
- TABLE 7.3.2 ..
~t JKEY PARAMETERS ASSUMED.IN-THE STEAM LINE RUPTURE'
- ANALYSIS 2-LOOP-275' MWT - '
q
~
f
,Referente-
. Parameter . : Units Cycle
- Cycle 5' Initial :Co're- PowerL MWt- 2754 '2754 Initial; Core' Inlet Temperature ~ F 550 ~550 .
'Initia1 LRCS LP'ressure psia 2250 2300:
i: Ini tial l Steam : Generator ' Pressure psia 870 853
- Low Steam Generator' Pressure Analysis psia- 478 '548
- Trip- Setpoint
~
, LSafety;InjectionActuationSignali psia 1578 1578 Minimum-CEA Worth-Availaole %Ap -5.7 -7.15 at. Trip'
~
' Doppler Multiplier 1.15 -1;15 Moderator Cooldown: Curve %Ap vs. density f**Ref!Ne$ce f3.2 Inverse' Boron Worth' PPM /%Ap 87 105
- Effective liTC X10-4Ap/ F -2.5 -2.2
- f8l fraction (including uncertainty) .0057 .0060-
* ~ Cycle 3 - last detailed' analyses presented (Reference 4).
'(Note:' The-last cc.aplete blowdown analysis for this Les4nt-was performed-for the.FSAR).
~. t e
6 4 e _ ' p L
'( . (
+
N _ _ _
c, .. . t TABLE 7.3.2 . SEQUENCE-0F EVENTS FOR: STEAM LINE RUPTURE E'/ENT WITH AUTOMATIC INITIATION OF AUXILIARY FEEDWATER AND MANUAL TRIP 0F REACTOR COOLANT PUMP 2-L00P-2754 MWT Time (sec) Event- Setpoint or Value 0.0 . Steam Line Rupture _0ccurs -- 2.4 Low Steam Generator: Pressure Trip - ~548.0 psia Signal Occurs 3.3 Main Steam Isolation Valves Begin 548.0 psia to Close
.s.
~3.3 Trip Breakers Open --
3.8 CEAs Begin to Drop Into Core -- - 9.3 Main Steam Isolation Valves -- Completely' Closed 14.8- Pressurizer Empties -- 4 15.6 Safety Injection Actuation Signal 1578.0 psia 15.6 Reactor Coolant Pumps Manually Tripped 1578.0 psia 23.3. Main Feedwater Rampdown 5% of Full Power Feedwater Flow 45.6 High Pressure Safety Injection Pumps Start -- 83.3- Main Feedwater Isolation -- 94.5 Affected Steam Generator Blows 0ry -- j 11 21.5 - Peak Return to Power 8.2% of 2700 MWt
'125.5- ; Peak Reactivity, Prior.to ' Auxiliary -0.10%ao Feedwater?F. low m _.-
~
7 -TABLE 7.3.2- 2 '(CONTINUED) Timeisec) Event Setooint or value
- 183.3 AuxiliaryLFeedwater Flow Initiated 350 lbm/sec to Ruptured Steam Generator 482.5 Peak Reactivity Post' Auxiliary +0.086%Ao
.Feedwater Flow 572.5 Peak' Return to' Power Post Auxiliary ~ 4.2% of 2700 MWt Feedwater Flow 600.0 Operator Isolates Ruptured Steam --
Generator and Terminates Auxiliary Feedwater Flow S l,
}-
4
~
TABLE'7.3.2-3 KEY PARAMETERS ASSUMED IN.THE STEAM LINE RUPTURE ANALYSIS 2-LOOP NO. LOAD Reference
' Parameter Units Cycle
- Cycle 5 Initial Core Power MWt 1.0 1.0 Initial Core' Inlet . Temperature F '532 532
. Initial RCS Pressure. psia 2250 2300 Initial Steam Generator' Pressure psia 890 899 Low Steam Generator. Pressure Analysis . psia 478 548 -
Trip Setpoint - Safety Injection Actuation Signal psia 1578 1578
. Minimum CEA Worth Available %Ap -3.4 -#.3 at Trip Doppler Multiplier 1.15 1.15 . Moderator Cooldown Curve %Ao vs. density SpeRe!Ne$ce*- 3.2$0 Inverse Boron Worth PPM /%Ao 87 '100 Effective MTC X10-44o/ F -2.5 -2.2 8 Fraction (including uncertainty) .0057 .0060
- Cycle 3 - last detailed analyses presented (Reference 4).
3 J D l
*~
TABLE 7.3.2-4 SEQUENCE.OF EVENTS FOR STEAM LINE RUPTURE EVENT WITH AUTOMATIC INITIATION OF AUXILIARY FEEDWATER AND MANUAL' TRIP OF REACTOR COOLANT PUMP 2-LOOP-N0 LOAD Time (sec) Event Setpoint or Value 0.0 Steam Line Rupture Occurs -- 2.8 Low Steam Ger.arator Pressure Trip 548.0 psia Signal Occurs 3.7 fiain Steam Isolation Valves Begin 548.0 psia to Close 3.7- Trip Breakers Open -- 4.2 CEAs Begin to Drop Into' Core -- 9.7 Main Steam Isolation Valves Completely -- Closed 11.6 Pressurizer Empties -- 13.6 Safety Injection Actuation Signal 1578.0 psia 13.6 ' Reactor Coolant Pumps-Manually Tripped 1578.0 psia 43.6 .High Pressure Safety Injection Pumps 1280.0 psia Start 83.7 Main Feedwater Isolation. --
-.180.0 Affected Steam Generator' Blows Dry- --
ij
~
183.7?' 1 Auxiliary Feedwater Flow Initiated to 350 lbm/sec: d Ruptured Steam: Generator i422.0: (. Peak Reactivity: ,
+.075f.ao
?600!O! 2 0perat'oriIsolates Ruptured Steam .. ~
6 , . iGerWGBPJMETCEIi&DGQa bhD%l9aPIt
I i I i
+6 _ 2 LOOP-FULL P0 lier _
+5 -
O ' d== +4 - -
? - '
E3 +3 6 m e 12 g +2 - E o
- E
+1 - -
0 - l l l -1 I ' ' I 40 45 50 55 60 65 DENSITY, LBM/ CUBIC FT
^
GAS E E T C co* STEAM LINE BREAK EVENT coivert clirr, - MODERATOR REACTIVITY FEEDBACK vs 7,3.2-1 Nuclear Power Plant MODERATOR DENSITY
1800 i. i i , , 1600 - 2 LCOP-FULL P0t!ER - 0 1400 .- _ S
" 1200 - - -
3 m
", 1000 - -
d 800 - _ . E5 Q 600 - _ R Muh % li!
"- 400 - _
7 200 - _ L 0- A J Ul@FFECTD SG
-200 i i i i i 0 100 200 300 400 500 600 TIME, SECONDS
'~~
' BALTIMORE GAS & ELECTRIC CO. STEAM LINE' BREAK EVENT FEEDWATER FLOW vs TIME 7.2.3-2 1Nuc r ow Plen, .
500 i i i i i 3 2 LOOP-FULL POWER 4
$ 400 - -
[ S
< 1 c
y 300 - w 5 4 3 8 200 100 -
~
0 0 100 200 300 400 500 600 TIME, SECONDS O BALTIMCiRE GAS & ELECTRIC CO. STEAM LINE BREAK EVENT-Calvert Clirr5 7.3.2-3 Nuclear Power Plant: FEEDWATER ENTHALPY vs TIME i( _
)
+8 i i i i 2 LOOP-FULL PCHER MODERATOR
+6 .
~
O
+4 -
+2 DOPPLER -
c. d se b TOTAL 3
. O x f n
5 BORON it o -2 - 6 c::
-4 .
~
SCRAM RODS
-8 ' ' ' ' '
0 100 200 300 400 500 600 TIME, SECONDS 1 BALTIMORE - GAS & ELECTRIC CO. STEAM LINE BREAK EVENT l kucI,Ir*NS[r"Aont REACTIVITY vs TIME 7'3'2 I
i 1 120 l I I I I 110 - 2 LOOP-FULL POWER _
~
100 . - - o 90 80 -- - BE r-
. 70 .. -
~
8 60 - h 50 fii ' - 8 c 40 e 30 S 20 - 10 -k - - 0 ' ' ' ' ' 0 100 200 300 400 500 600 TIME, SECONDS
~
BALTIMORE GAS'& El.ECTRIC CO. STEAM LINE BREAK EVENT calvert Ciirrs 7.3.2-5
.Nuclect Pcwer Plant CORE POWER vs TIME E-
110 , , , , , d 100 . 2 LOOP-FULL POWER _ 90 - - - o 80 -- e a: -- 70 - 8 S 60 - 8
", 50 -
s W 40 -
!E .
W 30 - si 20 - s:' w 10
'N
-~
0 0 100 200 300 400 500 600 TIME, SECONDS q BALTIMORE ' GAS & ELECTRIC CO.- calvert Cliffs STEAM LINE BREAK EVENT , Nuclear Power Plant CORE AVERAGE HEAT FLUX vs TIME 7'3'2 " ' i
1 650 , , 2 LOOP-FULL POWER o"- 600 - - vi i2 2 3 i-- 500 - - 3 m g - g T AVERAGE 5 o 8 400 - - g T OUTLET L3 3 m m 3M - T INLET I I I I I 0 100 200 300 300 400 500 i TIME, SECONDS i
^
GAS El.E T IC CO. STEAM LINE BREAK EVENT coivert cliffs 7.3.2-7
- Nuclear Power Plant REACTOR COOLANT SYSTEM TEMPERATURES vs TIME l
< 2500 i i i . i 0;
2 LOOP-FULL POWER uf 5 m ' O 2000 - E 3 m M ' E 1500 - 3 )- - 8 o 5 E 1000 - - u 500 i i i i I 0 100 200 300 400 500 600 TIME, SECONDS 1 GAS ELE IC CO. - STEAM LINE BREAK EVENT Calvert Cliffs REACTOR COOLANT SYSTEM PRESSURE vs TIME 7.3.2-8 , Nuclear Power Picnt. j
4
+1000 i i i i i ;
i r
+900 - 2 LOOP-FULL P0llER -
g' +800 - eg UNAFFECTED SG
+700 E
+600 - - -
8 E +500 - - b !8 +400 - - - 1 w
+300 - -
+200 - -
+100 -
~AFFECTED SG -
0 -
-100 0 100 200 300 400 500 600 TIME, SECONDS I
GAS EE IC CO. STEAM LINE BREAK EVENT calvert cliffs STEAM GENERATOR PRESSURES vs TIME 7.3.2-9 Ndef ear Pcwer Plant -
l 5.0 , ,. , 2 LOOP-N0 LOAD
, 4. 0 - -
<l <
s d g 3.0 - - A E u g 2.0 - 2 ai 8 E 1.0 - 0.0 / ' ' ' 45.0 50.0 55.0 60.0 65.0 MODERATOR DENSITY, LBM/FT3 BALTIMORE ' STEAM LINE RUPTURE EVENT Figure AsIII.tc;h c MODERATOR REACTIVITY vs MODERATOR DENSITY - 7.3.2-10 Nuclecr' Power Plant _=
f 2 LOOP-N0 LOAD 400 - 8 AFFECTED STEAM GENERATOR e 5 1300 - 3: 3 u x h200 - a te u_ 100 - UNAFFECTED _ . STEAM GENERATOR 0 ' ' ' ! 0 100 200 300 400 500 600 TIME, SECONDS i BALTIMORE - STEAM LINE RUPTURE EVENT Figure GAS & ELECTRIC.CO. Calvert Clins FEEDWATER FLOW vs TIME 7.3.2-11 _.Nuclecr Power Plant
100 i i i i ' 90 - 2 LOOP-fl0 LOAD _ 80 5 y70
~
m
>-' 60 I 50 -
5 e 40 - W h 30 tti
' 20 -
10 - 0 O 100 200 300 400 500 600 TIME, SECONDS BALTIMORE STEAM LINE RUPTURE EVENT Figure GAS & ELECTRIC CO' Coivert Clirts FEEDWATER ENTHALPY vs TIME 7.3.2-12 Nuclecr Power Plant l l
e 6 , , , , i LOOP-N0 LOAD 5 - MODERATOR - 4 3 - a 2 ' -
$ DOPPLER g' 1 -
f - M
<0 - r T0TAL z -_
-l -
g [ BORON C -2 -- E E -3 - SCRAM R0D WORTH f -
-5 -
-6 0- 100 200 300 400 500 600 TIME, SECONDS 1
i BALTIMORE
. GAS & ELECTRIC CO. STEAM LINE RUPTURE EVENT Figure
. coivert Clins REACTIVITY vs TIME Nuclear Power Plant 7.3.2-13 y + . , . , ,
100 , , i i , 90 . 2 LOOP-fl0 LOAD - 80 - - C
- E 70 8
S 60 - - 50 - 5 ' c-40
$30 a
20 - 10 - 0 ' ' ' ' ' 0 100 200 300 400 500 600 TIME, SECONDS l l l D BALTIMORE GAS & ELECTRIC CO. STEAM LINE RUPTURE EVENT Figure l Coivert Ciirrs CORE POWER vs TIME 7.3.2-14 Nuclear Power Plant 1
120 , , , , i 110 . 2 LOOP-N0 LOAD . 100 - - - 90 8g - - , Y d 70 s 6 x 60 - - E
& 50 -
5< g 40 - 8 - 30 . 20 - 10 - 0^0 100 200 300 400 500 600 TIME, SECONDS ! l I
~ BALTIMORE GAS & ELECTRIC CO. STEAM LINE RUPTURE EVENT Figure Coivert Clins CORE AVERAGE HEAT FL UX vs TIME .Nuclecr- Power Plant -
7.3.2-15
^
. 1 2500 , , , , i 2 LOOP-N0 LOAD 5 l m 1
' 2000 - -
u
!?!
O E 1500 - - a W:! m ! E 1000 - - 5 o 8 e - R o 500 - 6 m
-0 ' ' ' ' '
0 100 200 300 400 500 600 TIME, SECONDS BALTIMORE GAS & ELECTRIC CO. STEAM LINE RUPTURE EVENT- Figure
- Coivert Ciirr, Nuclear Power Plant REACTOR CO0LANT SYSTEM PRESSURE vs TIME 7.3.2-16
600 , , ., , . 2 LOOP-N0 LOAD
#- 500 -
d'. ' E S 400 -
\ S.s -
k:! VN . :E TN T OUT T 300 N AVE - m Ns g 5 N 17P s
- T--- h_ . _ _ _ _
o / 8 200
~
e L TIN S u
~
n" 100 - 0 ' ' ' ' ' 0 100 200 300 400 500 600 TIME, SECONDS BALTIMORE STEAM LINE RUPTURE EVENT sgure GAS & ELECTRIC CO' calvert clim REACT 0.R COOLANT SYSTEM TEMPERATURES vs TIME. 7.3.2-17 Nuclear Power Plant::
~
I 1000 , , , , ,
. 2 LOOP-N0 LOAD _
y n - _ : G - N
- 600 -
@ j-UNAFFECTFD SG Y l E : 1 e
S 400 si E 200 - h* '
- AFFECTED SG _ .
0
-100 0 100 200 300 400 500 600 TIME, SECONDS l
l L- BAl.TIMORE . STEAM LINE RUPTURE EVENT Figure 9AsIlhetc$;rS,' c STEAM GENERATOR. PRESSURES vs TIME 7.3.2-18
. Nuclear ?:wer Plant -
)
~
/
4 7.3.3; STEAM GENERATOR TUBE RUPTURE EVENT LThe' Steam. Generator Tube Rupture (SGTR) event was reanalyzed for Cycle 5 to verify that the- site b'oundary doses:will not exceed the- guidelines of t10CFR100.. -
. , a eThe analysis included the effects cf manually tripping the Reactor Coolant I
-Pumps on SIAS due to low pressurizer pressure.
The design basis SGTR is a d' ouble ended break of one steam generator U-tube.
' Table 7.3.3-1 lists the key transient related parameters used in this analysis. In the analysis, it is assumed that the initial RCS pressure
- is as-high as 2300 psia. This initial RCS pressure maximizes the amount of primary coolant: transported to the secondary staam system since the leak ~ rate is directly proportional to the difference- between the .
primary and secondary pressure. ;In addition, the higher pressure delays ~ the low pressurizer pressure . trip which prolongs the transient and therefore maximizes the total primary to secondary mass and activitics transported.
"or this event, the acceptable DNBR limit is not exceeded due to the action of the Thermal Margin / Low Pressure (TM/LP) trip which provides a reactor trip-to maintain the DNBR above 1.23. The tube rupture transient does not significantly affect the core power distribution. Therefore, the PLHGR SAFDL is not apprcached.
i The Thermal Margin / Low Pressure trip, with conservative coefficients which ' account for the limiting radial and axial peaks, maximum inlet temperature, 2 3 RCS pressure, core power, and conservative CEA scram characteristics, would , be the primary RPS trip intervening during the course of the transient. i However, to maximize the coolant transported from the primary to the i secondary and thus'the radioactive steam releases to the atmosphere, the " analysis was performed assuming the reactor does not trip until the minimum - setpoint' (floor) of the. Thermal Margin / Low Pressure trip is reached. This prol_ongs.the steam releases to the atmosphere and thus maximizes the site boundary doses. The Steam Cenerator Tube Rupture was analyzed with and without assuming a manual trip of reactor coolant pumos on Safety Infection Actuation
. Signal-(SIAS). The Steam Generator Tube Rupture (SGTR) with RCP trip on SIAS results in higher site boundary doses because: 1) RCP coastdown increases pressure difference between the primary and the secundary, which increases the'. leak rate,- and ' 2) - RCP_ coastdown decreases the rate of decay heat-
. removal which: increases the. steam flow through the atmospheric dump valves. '
The Sequence of Events for the SGTR event with manual trip of RCP on SIAS is presented in Table 7.3.3*2. Figures 7.3.3-1 through 7.3.3-5 present.
. the transient behavior of core. power, heat flux,.RCS pressure, RCS temperatures, and . Steam Generator pres ~sure. -
I-131 activity release is base'd on the primary to secondary leak and Lon the, steam, flow. required to reach . cold shutdown conditions. This . release is calculated as~the product of steam. flow, the time dependent
~
. steam activity _ and the decontamination factors: applicable to each '
Erelease pathway. ^ 1 eTho;0;to:2 hour I-131 :ite bcundary ~ dose, is calculated fron:
* .C'F 5 3 DOSE (' REM) [ Nb* ,BRT x 1 x m,
. . . _ .. ~ _ - .-- -
e- , - - - 9
.~:
.where[
.w .
I A-3h = I-131 Activity neleased LBR = breath,ing rate.. X/Q = dispersion coefficient j - CFI -131 . , . I-131' dose. conversion factor
-In determining the whole: body dose, the major assumption made.is that all nobel gases leaked through the ruptured tube will.be; released to the atmosphere. Therefore the whole body dose is proportional to the total primary to~ secondary leak and is calculated using the following
~
equation. whole body dose = + [.25L(E h E)3*L*g ARCS
- where: ,
E = average energy' release by gamma decay 7
~
'Eg = average energy release by beta decay
=
L total primary to secondary mass transport ARCS = n bel gas activity of primary. coolant X/Q = dispersion coefficient The results.of the analysis are that 82,462 lbs of primary coolant are
. transported to the steam generator secondary side. Based on this mass transport and values in Table 7.3.3-3, the site boundary. doses calculated are:
Thyroid (DEQ I-131) ~0.33 REM
- Whole Body- (DEQ XE-133): '0.08 REM-The reactor protective system (TM/LP) is adequate to protect the core from exceeding the 0:;3R limit. .Ths dose: resultir.g fecm the~ acti,/ity released as a consequence of a double-ended rupture of one steam generator tube, assuming- the maximun allowable Tech Spec activity for the primary concentration
-at a ' core power of 2754. MWt, are significantly below the~ guidelines of 10CFR100. .
iL-O
m , . g , i:.
- TABLE'7.3.3-1
- KEY.PARAtlETERS ASSUMED IN Tile STEAM ~ GENERATOR TUBE RUPTURE EVENT
~
- KEY TRANSIENT RELATED PARAMETERS _:
Reference-Parameter Units- Cycle *- Cycle 5~ Power MWt 2754- 2754 MTC X10-4Ap/ F -2.5 -2.5 Doppler Coefficient 1.15 1.15
- Multiplier ' -
Scram Worth %Ap -5.14 -4.3 Tg *F 552 550
~RCS Pressure psia 2250 2300 6
Initial Core flass Flow Rate X10 1b/hr 133.5 133.9 Initial Secondary Pressure psia 875 815
' Tube ID -Inches ~ .654 . .654'
~
Flow Constant- .956 1.17 ASI(forscram)-. +.16 +.41. u .
- Cycle 2~ 1ast detailed analysis presented (Reference' 5). I 4
9 9 r -
.a t'
> +-
Op-W- ,
~, l ' .
. - - _= . a
TABLE 7.3.3-2 SEQUENCE OF EVENTS FOR THE STEAM GENERATOR. TUBE RUPTURE EVENT WITH RCP C0ASTDOWN ON SIAS
. Time (sec) Event Setpofnt or Value 0.0 Tube Rupture Occurs --
660.8 Low Pressurizer Pressure Trip '728 psia Signal Generated 661.2 Dump Valves Open -- 661.5 CEAs Begin'to Drop Into Core -- 665.7 Pressurizer Empties -- 667.9 Safety Injection Actuation Signal 1578 psia Generated, RCP's Manually Tripped 669.2 Bypass Valves Open -- 621.0 Maximum Steam Generator Pressure 910 psia 724.6 Minimum RCS Pressure 1123 psia 1800. Operator Isolates Damaged Steam -- Generator and Begins Cooldown to 300 F 12365. P~ arator Initiates' Shutdown Cooling
.iAV=300*F) l b
TABLE 7.3.3-3 ASSUMPTIONS FOR THE RADIOLOGICAL EVALUATION FOR-THE STEAM GENERATOR TUBE RUPTURE Cycle 5 Parameter ~ Uni ts Value Reactor Coolant System Maximum pCi/gm 1.0 Allowable Concentration (DEQ I-131)3 Steam Generator Maximum Allowable pCi/gm .1 Concentration (DEQ I-131)1 . s Reactor Coolant System Maximum Allowable pCi/gm 100/if Concentration of Noble Gases (DEQ Xe-133)j Steam Generator Partition Factor --
.1 Air Ejector Partition Factor --
.6005 Atmospheric Dispersion Coefficient 2 sec/M 3
1.80X10'4
- Breathing Rate 3 M /sec 3.47X10~4 .
Dose Conversion Factor (1-131) REM /Ci 1.48X10 6' I Tech Spec limits 2 0-2 hour accident condition J
y y 120 , , , , , 100 - 1 E E 80 - - 8 m
.8 -
. 60 - -
x - IU e
.w 8
U 40 - - 20 - - Y ' ' ' 0 0 300 600 900 1200. 1500 1800 TIME, SECONDS
- BALTIMORE . STEAM GENERATOR TUBE FAILURE EVENT GAS'& ELECTRIC CO.
rigure Celvert Cirrrs WITH RCP C0ASTDOWN ON SIAS Nuclecr Power Plant - . CORE POWER vs TIME: 7.3.3-1
L' l 120 , , , , , 100 - - g i sg _ _ 8 Ri - E5
- 60 - -
x' - 3 u. W d5 z 40 - - E - 8 20 - - 0 ' ' ' ' ' . 0 300 600 900 1200 1500 1800 TiviE, SECONDS BALTIMCRE STEAM GENERATOR TUBE FAILURE EVENT Figure
' GAS & ELECTRIC CO.
ccivert ciurs WITH RCP COASTDOWN ON SIAS Nuclear Power ?lant CORE AVERAGE HEAT FLUX vs TIME- 7.3.3-2
2400 , , , , , 2000 - -
< 1600 M -
c. E' h1200 u c. 0 800 - - 400 - 0 ' - t ' ' 0 300 600 900 1200 1500 1800 TIME, SEC(;WDS B A LTIMORE STEAM GENERATOR TUBE FAILURE EVENT sgure GAS & ELECTRIC CO. WITH RCP C0ASTDOWN ON SIAS Nuc e on, REACTOR COOLANT SYSTEM PRESSURE vs TIME 7.3.3-3
700 , , , i i 650 - -
- g. -
d600 - - 5 \ N e \ m f T g g550 f n d'^
$ x 8500 - -
g T IN - x 450 - - 400 0 300 600 900 1200 1500 1800 TIME, SECONDS BALTIMORE STEAM GENERATOR TUBE FAILURE EVENT Figure GAS & ELECTRIC CO. WITH RCP COASTDOWN ON SIAS Nuc cw P!cnt REACTOR COOLANT SYSTEM TEMPERATURE vs TIME 7.3.3-4
1
.1 i
1000' , 1 (,
+
s 900 - - Oi a- ::. \ w' g 2 y ' ce: . . 4 - r 2 -
~,s m s M 3 w ,
.~-..
- e. T.
cL. e o 700 - - r w \ 5 s 600 - 3E a g u w + i 500 - J -
@ 'L '
y,
,h
-400 '- ' ' ' ' ' AJ 0 300 '600 900 A 1200- 1500. '"1800 .
TIME, SECONDp :l t h ~ ~ ly .~ l
\
. e '
l' ( \j ' 9 J BALTIMORE i :);*ure . GAS & ELECTRIC CO. ' STEAM GENERATSR 1 PRESSURE vs TIME 'n TUBE F l Caisert Cliff $
.WITH RCP C0ASTDOWN ON SIAS s Nuclear Power P!cnt STEAM GENERATOR l
-73i3_ Gi
., y y
..=
7.3.4 SEIZED ROTOR EVENT The Seized Rotor event was reanalyzed for Cycle 5 to demonstrate that the RCS upset pressure limit of 2750 psia will not be exceeded and only a smallifraction of fuel pins are predicted to fail during this event. The methods. used to analyze this event are consistent with the reference
~
cycle ~ analysis methods except CETOP/CE-1 with a DNB limit of 1.23 was used
-instead of TORC /CE-1 to calculate DNBR.
The single reactor coolant pump shaft seizure is postulated to occur as a consequence of.a mechanical failure. The single reactor coolant pump shaft seizure results -in a rapid reduction in the reactor coolant flow to the three-pump value. A reactor trip for the seized rotor event is initiated by a low coolant flow rate as determined by a reduction in the sum of the steam generator hot to cold leg pressure drops. This . signal is. compared with a setpoint which is a function of the initial number of operating reactor coolant pumps. For this event a trip will ,,s
. be-initiated when, or before, the flow rate drops to 93 percent of initial flow.
The initial conditions for the Seized Rotor event are listed in Table 7.3.4-1.
-These conditions are consistent with the initial conditions assumed for
- the LOF event (See Section 7.2.1). Other assumptions on key parameters are also listed in this table.
b In Table-7.3.4-2, the NSSS and RPS responses are shown for the seized rotor event' initiated from an axial shape index value of .22. The pressurizer pressure ' reached a maximum value of 23T3 psia at 3.50 seconds. Figures 7.3.4-1 through 7.3.4 4 show core power, core average heat flux, RCS pressure, and coolant temperatures during the transient.
~
A conservatively " flat" pin census distribution (a histogram of the number afo pins with radial peaks in intervals of 0.01 in radial peak normalized
- to the maximum' peak) is used to determine the number of pins that experience DNB.- The results show that the number.of fuel pins predicted to fail is equal to 0.85% in comparison to 0.5% for Cycle 4. This is a slight increase t'
- over Cycle.4, and remains a small fraction of.the total number of fuel pins.
1
,For the. case of the loss of coolant flow resulting from a seizure of a reactor coolant-pump shaft, a trip on low coolant flow is initiated to limit the predicted fuel failure to only a small fraction of the total number of pins. -Based on the . low probability of this event, the small number of predicted fuel pin failures is acceptable. In addition, the
- maximum RCS pressure experienced during the event will be well under the upset pressure limit of 2750 psia.
A ??>._ ) f: y r
. TABLE 7.3.4-1 L
~ KEY PARAMETERS ASSUMED IN SEIZED ROTOR' ANALYSIS Refdrence. Unit 1:
Parameter Units Cycle
- Cycle 5 Initial Core Power Level MWt 2754. 2700**
Core Inlet Coolant Temperature *F 550 548** 6 4-Pump Core Mass Flow Rate 10 1bm/hr 135.2 138.5** - 0 3-Pump Core Mass Flow Rate 10 lbm/hr 104.4 106.8**
-Reactor Coolant System Pressure psia 2200 2225**
Moderator Temperature Coefficient X10-4ap/*F +.5 +.5 Doppler Coefficient Multiplier --
.85 .85 CEA Worth at Trip "Ap
. -5.7 -5.6
~ Integrated Radial Peaking - 1.58 1.70**
Factor with Tilt;- F[
. Axial Shape.Index .15 .24**,+
Cycle last detailed analyses presented (Reference 3). CC . Uncertainties on these parameters were combined using the methods discussed
. in Section 7.2 and are. consistent with.LOF.
s+ ^ The ASI- used'.is conservative with respect to the most negative' ASI allowed by
--the DNB LCO..
r w _ _
TABLE 7.3.4-2 SEQUENCE-OF EVENTS FOR-SEIZED ROTOR Time (Sec) Event Setpoint or Value 0.0- ' Seizure of One Reactor Coolant -- Pump 0.0 Low Coolant Flow Signal 93% of Initial 4-Pump 2 Generated Flow - 0.50 Trip Breakers Open -- 1.00 CEAs Begin Dropping into Core -- 3.50 Maximum RCS Pressure, psia 2313 . r i 9
120 , , i i 100 - Ei! E 80 -
~
8 m I $ l * ! g 60 - 2 u o 40 - o -
-1 20 -
0 0 4 8 12 16 20 TIME, SECONDS gas $^E!E T IC CO. SEIZED ROTOR EVENT rigure Calver, cinh Muclear Power Plant CORE POWER vs TIME 7.3. 4-1
120 , , i i 100 - E ~
$ 80 - -
!E N
E5 8
- 60 - -
x 3 u_ l-- tl5 3 x 40 - - o O 20 - - 0 ' ' ' '
. 4 8 12 16 20 TIME, SECONDS 1
I L 1 I I
^
GAS ELE T (C CO. SEIZED ROTOR EVENT Figure cnive,r clah IIt#4 lrwit Pt tW (*t l'lt t t g l CORE AVERAGE HEAT FLUX vs TIME 7.3.4-2
2400 , , , , 2300 - - 5 2200 - - E u' 2 El2100 - u c_ 0 e 2000 - 1900 - 1800 0 4 8 12 16 20 TIME, SECONDS BALUMORE SEIZED ROTOR EVENT Figure Ascn O i cNiS. REACTOR SYSTEM COOLANT PRESSURE vs TIME 7.3.4-3 tjoc lear Power Plant
620 , , , i 600 vr T AV E 580 ,, s o w e /' ' N. x % E N s s T
~~. ' ' ~ ~
g 560 - IN ---- w a ._. . _ _ _ . . . .
- -'~
540 - - 520 ' ' ' ' 0 4 8 12 16 20 TIME, SECONDS I 1 i i BALTIMORE SEIZED ROTOR EVENT Figure GAS & ELECTRIC CO. j cnivert cmrt REACTOR COOLANT SYSTEM TEMPERATURES vs TIME 7.3.4-4
.. Nuclear. Power Pla.nt
~y a: ~ -. _.
^
[ .y
, iH
~
8.0! ntroduction and' Summary: + ' An 'ECCS ' performance $ analysih was performed for Calvert C11ffs-Unit I Cycle 5-
~
. to- demonstrate . compliance 'with 10-CFR~50~.46 which -presents the 'NRC Acceptance
, =. Criteria: for Emergency Core . Cooling . Systems for. Light-Water-Cooled reactors U) .
.The1 analysis' justifies an, allowable. peak linear heat generation rate-(PLHGR) fof 15 4.kw/ft.- ThisL PLHGR. represents an increase over the Cycle 4 limit .
.of 14.2 kw/f tiand.is' equal to the ext.; ting-limit for Unit II. The method of analysis and. detailed results which support-this. value are' presented herein.
.a - - - -
. 8.1 iMethod'of Analysis The a'nalysis forjunit:II Cycle =2 operation (6) , approved by the NRC, was usec
-as the reference cycle analysis for. the Unit. I Cycle 5 evaluation. The Uni'. II
- analysis was selected as the reference since the core in Unit I Cycle 5 is comprised only of high density. stable fuel as in Unit II Cycle 2.and the
. PLHGR~ for. Unit 'II is 15.5 kw/ft'. The'one residual low density
= fuelo assembly contained in Unit I Cycle 4N was removed.
The method of analysis used.the-NRC approved C-E evaluation model(2) . The
.model-was used to..re-evaluate the limiting large break LOCA performance.
The bicwdown and refil1-reflood hydraulic calculations employed in Unit II
. Cycle -2 were' performed generically for both Units I & II and apply to the Unit.I fifth fuel cycle. Therefore, only the STRIKIH-II(3) calculations-
- were necessary to account for the different fuel- pin conditions.
Burnup dependen(calculations were ' performed using the FATES (4) and STRIKIM-II(3) codes to determine:the limiting-' condition for the ECCS performance analysis.
- The PARCH (9) code was not utiilized in the Cycle 5 evaluation.
- The : late reflood heat trar,sfer'benefiti frem the use of the MRCH generated steam cooli.ng; heat: transfer coefficients would have reduced the peak clad temperature repo. rte'd' herein'.
- J N $
4 s' "3.- . ..'
y ,
=
r 4- - 4,
~
i L 852 L Resul ts ' , Table -1 presents the analysis results reported, for the 1.0 DES /PD* break. The
.1.0 DES /PD break is the. limiting break for-Unit I Cycle 5. The reference cycle
~
analysis for: Unit .II Cycle 2 definesithis as the' limiting break size for high density. fuel when,' clad ~ rupture occurs during the refill period as predir.ted in this evaluation ifor Unit I. The results of the evaluation confirm that-
-15.5 kw/ft is an acceptable value for the PLHGR in Cycle 5 The peak clad temperature:and maximum laeal and core-wide clad oxidation values as shown in Table 1 are below 10 CFR 50.46 acceptance limits.
a Table 2 Presents a list of the significant parameters displayed graphically
'for the limiting 1.'0 DES /PD break.
8.3 Eva'suation of Results The reason for the lower peak clad tempe ,ture'(PCT) for t ait I Cycle 5 (Table 1) as compared to the Reference C.vcle, Unit II Cycle '2, desp te a higher initial sto' red energy for Cycle 5,(Table 3) was due to the niore f. vorable overall fuel performance, a lower heat sink temperature and improved h at transfer conditions, e.g.,- a lower fuel' rod gas pressure and a lower hot bundle linear heat rate (Table 3), he'nce a lower; hot bundle average power. Since Unit 'I Cycle 5 had a lower hot bundle average power 1. nan in tne LReference-Cycle,'the transient enthalpy during the later portions of the ' blowdown period was ' lower. - Therefore, the' residual fuel stored energy sand clad temperature at the-start of the refill period were also lower. q t
'
- DES /_PD = Double-Ended Slot at Pump Discharge-4 k
a . O
, s =_ -
= -
~
, - Th7 fueliand clad heat-up duringLthe refill period therefore i proc:.eded. at
~
J
- a slow:riratfresultin9L botli;iallov;er clad temperatures' and lower. clad :
;oxi.dation:Shaniin the Reference [ Cycle. lThe.hotrodgasLpressure'(Table1 3)
.whichhwas initially lower than .in the Reference Cycle,- together with the
~
- lower.refil1 period fuel and: clad.tamperatures,'resulted in clad rupture-occuring.4.0' seconds later than in the: Reference Cycle.i As a consequerce
'of Lthis delay Lin clad rupture, the more favorable refill period heat removal .
from-the fuel and clad was prolonged. 'During the reflood period, after
~
- reflood rates have fallen 1below 1.0 inches;per_ second, the lower average
- hot > bundle power enhanced the rod-to-rod radiation cooling _ of the hot rod
.by provi. ding la lower heat sink temperature. The net -result was a slightly-D
' lower PCT-(by 49 F) and a= lower. peak local claa oxidation (.by-2.1%) as shown<in Table 1. - -- .. .- -
8.4L Conclusion As discussed above, conformance'to the ECCS criteria is summarized by the analysis results presentedLin Table 1. The results of the analysis identified the peak clad. temperature as 1942 F as; opposed to the acceptance limit of
-- 2200 F'. The peak local clad oxidation was 8.2% versus the acceptance limit oof -17% and the-peak core-wide clad oxidation was less -than .51 % versus the acceptance limit of .l.0%. Hence, Unit I Cycle 5 operation at a peak linear heat generatien. rate of 15'.5 kw/f t and at a power level' of 2754 Mwt (102% of
~
12700' Mw t
) will result in acceptable ECCS performance.
?8.5 Computer Code Version Identification The following version of Combustion Engineering ECCS Evaluation Model computer _ code was used in this analysis:
STRIKlif-II': Ve'rsion !!o. 77036 G e k as
~. l e'
w b' #
- .u
- i Table l ~
i Calvert Cliffs Unit l Cycle 5: Mmi. ting Break' Size (l'.0 DES /PD)
~
., . B1'owdown Peak Peak Clad-
- Time of Peak- Time of Clad Peak' Local Total Core-Wide. .
(Analysis- , Clad Temperature Temnerature Clad Temperature Rupture' Clad 0xidation Clad 0xidation: , , I * , . M
- Uni,t I,
- Cycle 5 _1725 F 1942'F 253. sec- 33.4 sec '8.2% G .51%~ ,
.c
-RbferenceCycle.' .
. , n-t (Unit 11,~ Cycle. 2) ' 1725 F 1991*F 248.'sec 29.4 sec 10.24% . < . 51% '
-.? .
' 't "'
s. l , r e
- 4 b
i p l. l. E
g5g - - =-- --- + q::, . u.
; . y
; by >
.-: x b _
w J-{ 9., . Table 2 - - e
.Calvert Cliffs'I Cycle 5-
' Analysis Plots b, . ..
. _' . Figure Variables- . Designa tion .
Peak ~ Clad: Temperature 1 A~ s Hot Spot Gap' Conductance 1B Peak LocallClad Odidation 1C Cla.d Temperature, Centerline Fuel Temperature,
- Average -Fuel Temperature and Coolant : Temperature for Hottest Node 1D
- Hot Spot Heat Transfer. Coefficient 1E Hot 1 Rod Internal Gas' Pressure. 1F
., g P
.,\ *
. i.
'C 9 f t
= -
t: _ p >: - w
' ^'
e ; - - -
~
[;tl . '. ' ' y s 4
- r; ,
i Table 3
~
Significant Parameters _' > Quanti ty Unit I Values Unit II Values Cycle 5- Cycle 2 Reactor Power Level /102", of Nomi_nal) 2754 2754 :Mw ' t Av: rage Linear Heat Rate (102% of Nominal) - 6~.45 5.52 kw/ft
- P;ak-- Linear Heat. Generation Rate-(PLHGR)
Hot Assembly, Hot Channel .15.5 10.5 kw/ft Peak' Linear-Heat Generation Rate .(FLHGR) Hot Assembly, Average Channel 13.43 13.57 kw/ft Gap-Conductance at PLHGR : 1704* 1731* BTU /hr-ft-5 Fuel Centerline Temperature at PLHGR 3626* 3604* F Fuel Average Temperature at PLHGR 2242* . 2219* *F 'H'ot Rod Gas Pressure 1144* 1198* psia Hot-Rod Burnup -753* 1522* MWD /HTb CFori high ' density fuel .:when gap ~ conductance is minimum I y 4 7
? V s p
~
.. lj h hL :- w ,
fik .
g; x ,2 m - 27-
== - w qn g nrc. m. n -
- =
=1x_ :CALVERTililFFS b.1iTCYCLE 5 m-- l',= 0 L x l.u, bulE. tuot9 ,_ . ... .. S.w!:i. -. _&_,,, ..s ;, .i . Pb.a ? .ul S .. r, J. .in
., .u.
_G
. PEAR CLAD-TEJ.PEFdi~bRET 2200'. t i. , , , ,
1!!GG.tS!TYFhELAT15.5i:il/FT ;
- 2000. -
. ~
s
, s
%^ .,
/ N
-1800. /
s
.s.
~
t .% t
-l- % . . u_ :. .e s o
c s ,y ..
* : ~
t
< 1600. 4,o i
~'
s
~
I '
'g- g .- % ,,,
- ca g .
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' t. r -
s %r %\
- M t , I
- LLJ .. , ~\
- a_- n( s 1.
.s' r-go 1L100. l- -
PEAK CLAD TEMPERATURE N0DE ca _ - - - RUPTURE N01]E
. e.k I ,
iL ) : i If hr. ~
$-1200. . m. .
1 in 31000. .- ,
- u l
o N 9
'q -me ww.,a
-l
? 600. . . !
0 . 100_ 20D. . 300
- F. ^t n ~-
. R. . .~ r, i
. ... .. 1
, t n ' 4, 4 _
L
'h
= _ . -_. . _.
t CALVERT CLIFFS U!11T"I CYCLE?5 ~ 1,0--x!D0UljLE EiidED SLOT 13REAK IN PUi1P DISCI)ARGE LEG
; HOT SPOT GAP C0r!DUCTANCE V. , .
p c t, - kn600.' ,. , c . f 1 r ' ;11!00. -
-HIGH DEiiSITY FUEL AT 15.5 KluFT ~
L 1200. - iu
- t-
[1000. x I s D-F- (D - s;3.800. -- (_y -
=c .
I-O:
,2 r.. .
g 600. i Q w-4 . r1 ! 'l10 0. -. .
. 'i
] .-;~...______
200, - e 10. L' - - - -
' ' ' ' J--
- r. 200 .iUO 4Cu Suu 600 ,.
10e . ' 100.: . I
.TIIiEr SEC0iiDS g
p' - jMp'ebubsh'i;t
- 3. M" E u gf JR +
~~ --
< -i
, JCALV$iri :C'UlFFS Ui!IT Hli CYCLE 5 1.0:x I:003LE Ei!.6ED: SLOT IIREAK IMP $iP DISCHARGE LEG.
~
PEAK' LOCAL: CLAD OX1DAT101
-16. - ,. , ,. _, , ,
- lilGH DE!iSITY FUEL AT 15,5 Kil/FT
.14. -
-12. - -
'10. - -- PEAK CLAD TEMPERATURE i!0DE -
a .
- - -- RUPTURE tiODE e
.O.- <C. E3 w
-8, __
a -
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~
'CALVERT CLIFFS-'J.11Til: CYCLE ~c._
!1.0 x LI:0L'3LE:Edi.TD ' SLOT Li!REAK li! POMP DISCHARGE LEG CLAD'TEMPEilATURE,: CEUTERLli'E FUEL TE'iPERATURE, AVERAGE-FUEL 1 TEMPERATURE A!!O C00lldT TEMPERATURE FOR !!0iTEST NOD i
HIGk !E'SITY Fl!EL AT 15,510UFT
.. 3500. -
i , 3000,- - l-e i
'2500.- - : " ",:
FUEL CEliTERLII!E V
-2000, 1 w%... N
[ ( ,
/
\ %
j 1500.
,/ , j- ,
AVERAGt FUEL
/ -
CLAD I, i 1000. ,IL- ,
~
3 uo0t.hni: -
- 500. -
N
- c. , - .-
- t. r i . . _ a.__...__.i.......__._'_- - - - -
0- J100L 200' 300 4'00 '500 600 , TIP.E, SEC0!DS F**o'*Jo'TNfA\ L
' =
dau e JL J N H _a
CALVERT CLIFFS Ur!!T.I CYCLE -5 1.0 x DOUBLE EiluED SLOT BREAK li! PUnP DISCl:ARGE LEG liOT SPOT !! EAT TRANSFER COEFFIClE:!T 160, i i i i . i lilo. - HIGH dei!SITY FUEL AT 15.5 KW/FT u_ 120. - i
~4 lZ e =
s 100. ;- .. g , i. ,.. l, 5!6!
?.d 80. '- -
u_ , ih - 8 l S3 h.
- 60. ~ -
EE 3 ' L10. ,l, - t . t ;
- 20. - -I
- l. g I- ' ' -l 1 -L 0.
l 0 100 cu0 500 : . uu. Sus tuu ,< U
$' j o TIIiE, SECO!DS
f TFIGURE -lF-b- ^ CALVERT CLIFFS L5!1T L CYCL.E: 5l - 11.0;x; DOUBLE lEi'DED SLOT ~ BREAK Ill-PUNP DISCHARGE LEG - F. 'l10T R00 IdTERWAL-GAS PRESSURE llIGH DE!iSITY FUEL AT 15.5 Kh'/FT 11400.- i i i i 1200' P = 1144,0 PSIA - INITIAL
/
RUPTURE AT 33.4 sac 1000. - - Fo ~ a_ 800. - -
- Lt s
R i: W -600. 400,- -- - 200, -- - O. ', LO: 20 40 6n 80 100 L. n ~ ' TlHE, SECOWDS-s ^ 5'; . k; <
V'
!9. ' TECHNICAL SPECIFICATIONS.
In this section all-. changes that' must be made to the '. Technical Specifications - are- provided in order. to make ,the Technical-Specifications.. valid for operation of Cycle 5. Each -page from the Technical Specifications which must be modified is
,, shown with.the modification included.
i l J m a b
Pags 4 TABLE 1. Calvert Cliffs,I Cycle 5 Technical Specification Changes PROPRIETARY INFORMATION
.:' COMBUSTION ENGINEERING, INC.
~ Change # ; Tech Spec # Action 1 Figure 2.1-1 page 2-2 -Replace Figure 2.1-1 with enclosed Figure 2.1-1 2 Table 2.2-1.page J-9 Change steam generator pressure-low setpoint from >500 psia to >570 psia 3 -Table 2.2-1 page 2-10 Add steam gene'rator pressure difference -
high setpoint 4 Table 2.2-1 page 2-10. Change steam generator pressure-low trip bypass from below 600 psia to below 685 psia 5 Figure 2.2-1 page 2-11 Reanalys'is for Cycle 5 has produced no changes in the kw/ft LSSS trip 6 Figure 2.2-2 page 2-12 Reanalys.is for Cycle 5 has produced no changes in TM/LP trip 7 Figure 2.2-3 page 2-13 Reanalysis. for Cycle 5.has produced no changes.in TM/LP trip 8 B.2 1.1 page B2-1 Remove' numerical specification of LHGR to centerline melt 9 B.2 1.1 page B2-1 Change high power level trip from 112%
-to 110%
10 B.2.1, B.2.2 Change minimum DNBR value from 1.19 to 1.23
* -: - pages B2-1, B2-3, B2-5, as indicated on noted pages B2-6 f'
-~ " ll- B.2.2.1 page B2-4 . Change maximum high power level-trip actuation from 112% of rated thermal power to-110%
12- B.2.2.1 page B2-5 Change steam generator pressure-low o
.setpoint from 500 psia to 570 psia- ;
413 B.2.2.1 page B2-7
~
Revis'e' description of. TM/LP trip and add asymmetric steam ganerator: transient.protec-tive trip function.' description i ri e >
s _ , s Paga 5 TABLE 1 (continued)
-Change # Tech Spec # Action 14' . 3.1.1.1 oage 3/4 1-1 Change Shutdown Margin Tavg >200 F.from
>3.4%Ak/k to >4.3%ak/k and change minimum boratTon concentration from 1720 ppm to 2300 ppm 15 3.1.1.2.cace 3/4 1-3 Change Shutdown Margin Tavg 1200 F from >l.0%;k/k to >3.0%a k/k and change minimum boration concentration from 1720 ppm to 2300 ppm '
16 3.1.1.4 page 3/4 1-5 Change MTC 'ess negative than 32 5x10~4ak/k/ F to less negative than -2.2x10 ak/k/cF i 17 3.1.2.2 page 3/4 1-9 Change Shutdown Margin equivalent from at least 1%Ak/k' at 200oF to at least 3%Ak/k 18 3.1.2.4 page 3/4 1-11 Change Shutdown Margin equivalent from at least 1%Ak/k at 2000F to at least 3%ak/k 19 3.1.2.6 page 3/4 1-13 Change Shutdown Margin equivalent-from at least 1%Ak/k at 2000F to-at least 3%Ak/k
?
20 3.1.2.7- page 3/41-14 Change refueling water tank minimum borated water volume from 9,978 gallons to 9,844 gallons' 21 3.1.2.7 page 3/4 1-14 Change refueling water tank mininum boron concentration of 1720 orm to 2300 com 22 Figure 3.1-1 Change minimum boric acid storage tank
. page 3/41- 15 volume function 23 3.1.2.8 page 3/4 1.16 _ Change rafueling water tank boron concentration from between 1720 and 2200 ppm to a minimum of 2300 ppm and Shutdown Margin equivalent from 1%Ak/k at 200 F to 3%Ak/k at 200oF
.24 - Figure 3.2-1 -
Change allowable peak linear heat rate-L page 3/4 2-3 from 14.2 kw/ft to 15.5 kw/ft 25 . Figure 3.2-2 Replace Figure 3.2-2 with enclosed
. page 3/4 2-4 Figure 3.2-2 h -
g - Page 6--
' TABLE 1 (continued)-
l Change # Tech Spec # Action 26 -Figure 4.2-1, Replace Figure 4.2-1 with en.losed page 3/4 2 Figure 4.2-1 27 3.2.2 page 3/4 2-6 Change calculated value gf Fxy from s_1.660 to 5.1.700 and Fxyl >1.660 to Fxy >l.700 28 Figure 3.2-3 Replace Figure 3.2-3 with enclosed page 3/4 2-8 Figure 3.2-3 29: 3.2.3 page 3/4 2-9 Change calculated value of FrT from
<1.571 to <1.700 and change FrT>l .571' t'n FrT>l .700 30 Figure 3.2-4 Replace Figure 3.2-4 with enclosed pace 3/4 2-11 Figure 3.2-4 31' Table 3.3-1, Add steam generator-pressure difference -
page 3/4 3-2 high description to table 32 Table 3.3-1, Change steam generator pressure-low t. rip page 3/4 3-4 bypass from below 600 psia to below 685 psia
-33 Table 3.3-2, Add steam generator pressure difference-page-3/4 3-6 high response time 34 Table'4.3-1, Add steam generator pressure difference-page 3/4 3-7 high surveillance 35 Table 3.3-3, Change Main Steam Line Isolation steam page-3/4 3-15 generator pressure-low trip bypass from below 600 psia.to below 685 psia 36 Table 3.3-4, Change Main Steam Line Isolation steam page 3/4 3-17 generator pressure-low setpoint from _>4,78 psia-to' > 570 psia 37- Table 3.3-5, Change Containment Purge Isolation Valve-
- page 3/4 3-20 Response time from <6 to <5 sec
'38 '3.5.1 page 3/4 5-1 Change safety, injection tank minimum boron concentration from between 1720 and 2200 ppm to a~ minimum of 2300 ppm.
39- 3.5.4 page 3/4 5-7 Change refueling +ater tank minimum boron concentration from between 1720'and 2200 ppm to a minimum'of 2300~ ppm y .
~ ,
m - 4 . m
-Paga 7 TABLE .1 ' (continued)
Change # Tech Spec # Action-40 3.9.1 page 3/4 9-1
~ Change refueling boron concentration of >l720 ppm to >2300 ppm and boration-at s40 gpm of 1770 ppm to boration ate 40gpmof2300 ppm and shutdown margin from 1%ak/k to 3%Ak/k 41 3.10.1 page 3/4 10-1 Change boration at >40 gpm of 1720 ppm to boration at ,>40 gpm of 2300 ppm 42 B 3/4.1.1.1 and B 3/4.1.1.2, '
Change minimum Shutdown Fiargin with Tav9 Page B 3/4 1-1 100 2 F from 1%Ak/k to 3%Ak/k and revise basis Change 3813 gallons of 7.25% boric acid solution to 6500' gallons and 47,204 gallons of borated water to 55,627 gallons. 43 B 3/4.1.2, pages Change Shutdown Margin of 1.0%Ak/k after B 3/4 1-2, B 3/4.1-3 xenon decay and cooldown t) 2000F to 3.0%Ak/k after xenon decay and cooldown to 200oF and the refueling water tank boron concentration from 1720 ppm to 2300 ppm 44 8 3/4.1.2, page Change 9,978 gallons of borated water to ' , B 3/4 1-3 9844 allons1and 439 acidgo737 gallons.gallonsof7.25% boric , 45 B 3/4.2.5, page Change minimunt DNBR of 1.19 to minimum B 3/4 2-2 DNBR of 1.23 ' 46 8 3/4.9.1, page Change minimum boron concentration B 3/4 9-l_ (1720 ppm) to (2300 ppm) 47 3.4.1.page 3/4 4-2 Include specific operation of reactor i coolant pumps for Mode 3 l l 1
^ , a w; , 7 _ x_ . w <
( l
w y? ' . . l:
;(
Y
..r -
Page~8 .
-TABL'E 2 Explanations for Cycle 5 ; Tech Spec Changes x-Change # Tech Spec:# Explanation Thermal Limit Lines have.been changed to 1: Fi9ure'2*1~1 reflect higher radial-peaking factors and.
implementation of margin recovery programs. 2 Table 2.2-1 The steam generator pressure-low setpoint is being increased to ' minimize the consequences of a Steam Line Break Event.- 3 Table 2.2-1 A trip for Asymmetric Steam Generator pressure has been added to minimize the consequence of the Loss of Load-to One Steam Generator Event. 4 Table 2.2-1 The steam generator pressure-low trip bypass has been increased to be consistent
-with the new trip value.
5 Figure 2.2-1 Reanalysis for Cycle 5 has produced no changes in the kw/ft LSSS trip 6 Figure 2.2 'Recnalysis for Cycle 5 has produced no changes.in TM/LP trip
.l p .j
[7 Figure 2.2 3 Reanalysis .for Cycle 5 has produced no changesinTM/LPtrip . J E " The numerical specification of centerline ineltL 8 B.2.1.1 limit'is being deleted ~to-standardize spec to other C-E plants.
-9: B.251.1- " statistical Combination of Uncertainties has removed.the 2% power uncertainty.
4 - from the~ transient analyses.
- 10 8.2.1,-B.2.2: w The minimum DilBR'has~been~ increas n of Uncertainties 411 - 'B.2.2.lc : -
- Statist!ibal' Combination' of Uncertainties
~
1, m
=has removed the 2% power. uncertainty.from if ~
m.the transient analyses. _- a;. t 'q
,1. s
.?. 3 _ c. cl '
-.2 _
r - T Page 9
^
~ TABLE 2-(continued)
Change # . Tech' Spec # Explanation 12- B.2.2.1 The basis of the-steam generator pressure-low trip setpoint has~been' changed to be consistent with Table 2.2-1.
, 13 B.2.2.1 The TM/LP basis has been streamlined for clarity and a description of.the asymmetric steam generator pressure trip has been added
-to the bases.
14 '3.1.1.1 The shutdown aargin has been increased to yield acceptable consequences from a Steam Line Break Event. The new baron concentration is consistent with the new .re-fueling water tank concentration for Cycle 5. 15 3.1.1.2 The shutdown margin las been increased to lengthen the operator action time required in a boron dilution evant. The new boron concentration is consistent with the'new refueling water tank concentration for' Cycle 5. 16 3.1.1.~4 Th' most negative MTC permitted for Cycle 5 has been made less negative to yie'1d acceptable consequences from a Steam Line Break event. 17- 3.1. 2. 2 - The required shutdown margin has been increased to be censistent with Tech Spec L 3 .1.1.2. 18 3.1.2.4 The required. shutdown margin has'been increased to' be consistent with Tech Spec 3.l.1.2. s
-19 3.1.2.6 The required shutdown margin has been ~
increased to be consistent with Tech Spec 3.1.1.2 20 : 3.1.2.7 Thevolumeofboratedw$terhasbeen decreased-due to the highehts61uble boron concentrations. / 21 3.1.2.7- The refueling water tank boron' concentration has been changed to be i
. consistent with Tech Spec-3.9.1. ' 1 4
522 Figure: 3.1-1 The volume of borated water has-been- >' v ' increased.to allow a higher' shutdown . '
-l boron 1 insertion - due.-to the;h'i.gh.er icore l g
y x .o averageenrichmentsofqut'ure}cyglesh
Page 10 TABLE 2(continued)
? Change #~ . Tech Spec # Explanation 23 3.1.2.8' The refueling water ~ tank boron concentration has been changed to be consistent with Tech Spec 3.9.1'and the required shutdown margin has been.
increased to be consistent with Tech Spec 3.1.1.2 - 24 Figure 3.2-1 The allowable peak linear heat rate is being increased from 14.2 kw/ft to 15.5 kw/ft. 25 Figure 3.2-2 The LHR. LCO is.being changed as a i. result of higher radial peaks and implementation of margin recovery programs. 26 Figure 4.2-1 Augmentation factors have been increased to envelope future, cycles 27 3.2.2 Radial peaking factors, both FxyT and FrT ,are being raised for Cycle 5. 28 Figure 3.2-3 Radial and FrT,are peaking factors,forboth being raised FxyT Cycle 5. 29 3.2.3 Radial peaking factors, both FxyT' FrT ,are
~
and being raised for Cycle 5. 30 Figure 3.2-4~ The DNB LC0 limits are changing due to higher radial-peaks and margin recovery programs.
- 31. Table 3.3-1 The asymmetric steam generator pressure trip has been added to the table.
32 Table 3.3 The steam generator pressure-low trip bypass has been increased to be consistent with the new trip value.
-33 ' Table 3.3-2 The asymmetric steam generator pressure trip has been~added to the table.
34 Table 4.3-1. The asymmetric steam generator pressure I trip has been added to the table.
- 35. -Table.3.3-3 The Main Steam Line Isolation steam generator pressure-low trip _ bypass '!
has.been increased to be consistent with the new trip-value. 36: Table:3.3-4 The Main Steam Line Isolation steam-
,m o 3 P generator pressure-low trip.setpoint ;
9 m _ W c u " has been increased ~to be consistent: .
;d[ z A ' ;* y / y
.e v .1 s * +
t' g p r
, i h s q. L
* ^
.j . rf ' !' Pace 11
'f r
\' '
TABLE 2 (continued) 1' ,
..i y[s
' h l' 11 Change # Tech Spec # '
w % Explana tJ,qn, , 37 Table 3.3-5 Containment' isolation value resqons'e ffe,e D is being reduced from'6 seconds to 5 1,econds) to satisfy NRC requirements. (fEC Branen Technical Position CSS 6 4) , s g 38 3 5.1 injection tank boron ' [ The safet[lonihas bcen increase concentrat assure'a uniform tsoron' concentration
') {
. ,in alh coolants that have access to the 5 -
i reactor vessel, t s ,g. s s ;w q M'e 39 3.5.4 The refueling water tank tioron concentrM ' 7 't tion has,been increased to t'e' consistent
. withTecd. Spec (3.9.1
'N
* *t. ,
40 3.9.1 g ' The' refueling boron cloce'ntrations. have
~"
teen ' increased dur;to the; higher core
,% average enrichdit 'cf future cycl se 'and
'the shutdown margi%has ifereased'to be consistent witL3.1.1.2.
n. {\t i < 41 3.10.1 The boratton co'ncentiations have been
, increased 'to be' consistent with .the new
\ 'soron concentration of the refusling
,' water tank.
5
- 3 t,3 42 B 3/4.1.1.1 and . ,The shutdown margins in the bases have B 3/4.1.1.2 '
1 i been increased to be consistent with
those in Tech ~ Specs 3.1.1.?s and'3.1.1.2 and
'*\
* $r;@isfq aoplicsbility of shutdoEn margin'
'e /for steamiline.
' break acci6ent.
<a t i
.3 ,,
The number of gallons of PPi1 boron has increastfe.to accommodate increased boron ' insertion requirements for future cycles. 2,, 't 43 B 3/4.1.2 The shutdow'n margin has been increased t in the bases tg be consistent with ,, Tech Spe: 3.1.1.2. The refueling ' water tank boron conceptretion in the bases has been increased to be consistent with Tech Spec .3.9.1 , 44 8 3/4.1.2 page. The volume of boIKted\ water in-BAST has p ' B 3/.4 1-3 been decreased due,to;the hiiher. soluble boron concentration'anj increased in RWT l due.to increased\ boron' insertion require-e #' s n
, 'ments.
.-7
-E
a v 7' ~ y it,b n.; y 1 Page 12-N TABLE 2 (continued) 36
., '4 'u C
y Change # . Tech Spec #- Explanation '.tY 4 45 B'3/4.2.5
, The minimum.DNBR has been increased 'to a,
.be. consistent with. Tech Spec B.2.1-
' 46 B 3/4.9.1 The refueling water concentration in r
l the bases has been increased to be consistent with Tech Spec 3.9.1
~47 3.4.1 One-loop no load conditions have not been
, analyzed for Cycle 5 t 4
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J G . j . TABLE 2.2-1 (Cont'd) 9 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS
; G 1
9
- FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES P
g . 4. Pressurizer Pressure - High < 2400 psia __ < 2400 psia u, * -
; , 5. Containment P,ressure - High 1 4 psig 1 4 psig i
b
^
- 6. Steam Generator Pressure - Low (2) > M ia
> a n 7. Steam Generator Water Level - Low > 10 inches below top . > 10 inches below top j _
of feed ring. of feed ring.
- 8. Axial Flux Offset (3) Trip setpoint adjusted to Trip setpoint adjusted to not exceed the limit lines not exceed the limit lines
' 9. ThermalMargin/LowPressre(1)
'? a. FourileactorCoolantPumps Trip setpoint adjusted to Trip setpoint adjusted to
- Operating not exceed the limit lines not exceed the limit lines
. of Figures 2.2-2 and 2.2-3. of Figures 2.2-2 and 2.2-3. l
'b. Three Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to Operating not exceed the limit lines not exceed the limit lines a
of Figures 2.2-4 and 2.2-5. of Figures 2.2-4 and 2.2-5. I
.I c. Two Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to j Operating - Same Loop not exceed the 1imit 1ines not exceed the 1imit 1ines ; y of Figures 2.2-6 and 2.2-7. of Figure 2.2-6 and 2.2-7. 1
- g. .
& d. Two Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to g Operating - Opposite Loops not exceed the limit lines not exceed the limit lines of Figures 2.2-8 and 2.2-9. of Figures 2.2-8 and 2.2-9. I O
s
- A '3 -
1
-}
s TABLE 2.2-1 (Continued) [Ty. .. / REACTOR PROTECTIVE INSTRUMENTATION TRIP-SETPOINT LIMITS
. ,R.
A - .c, n - FUNCTIONAL UNIT- TRIP SETPOINT ALLOWA8LE VALUES g.,j 1 9a LStsam ~ Generator: Pressure < J J35 psid
~
-< s Js 5 psid' n
' Difference - High ~ (1)
- E ' 10. Loss of Turbine -- Hydraulic -> 1100 psig -> 1100 psig
'Ze Fluid J Pressure <- Low -(3)
- 11. LRate'of. Change of Power - High (4) < 2.6 decades per minute'
_ < 2.6 decades per minute G
- TABLE NOTATION,
.(1) ~ Trip may'be bypassed below 10-4% of RATED THERMAL POWER; bypass shall be automatically removed when -
. THERMAL : POWER is : > 10-4% of RATED. THERMAL'POWE_R.
N taV5 7- [ ' y;(2) Trip may be manually bypass'ed below
)6 rpsia; bypass shall be automatically removed at' or above psia.^
- (3) Trip may be bypassed below 15% of RATED THERMAL POWER;; bypass shall be automatically removed when THERMAL POWER is > 15% of RATED THERMAL POWER.
(4) Trip may be bypassed below 10-4% and above 12% of RATED THERMAL POWER. L _ _
i* 2.1 SAFETY LIMITS '
-p-BASES I.1.1 2
REACTOR CORE. g i M " mit fuel cladding andinThepossiblerestrictions of this cladding perforaticn whichsafety no limit prevent overhe f I I he fuel is prevented by maintaining the "eady state peak Overheating of li l _ ate, t;elow the level at which centerline fuel moltinn will occur. Overheating of the fuel cladding is prevented by restricting fuel operaticn , to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. F Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of , l e departure from nucleate boiling (DNB) and the resultant sharp reduction
- in heat transfer coefficient. DNB is not a directly measurable parameter (
during operation and therefore THERMAL POWER and Reactor Coolant Temper- . ature and Pressure have been related to DNB throJgh the CE-1 correlation. ,
. pp The CE-1 DNB correlation has been developed to predict the DNB flux and
- b the location of DilB for axially uniform and non-unifom heat flux distri- I butions. The local DNE heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the '
local heat flux, is indicative of the margin to DNB. i' The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to +-W~~ l 23' l 2 This value corresponds to a 95 percent probability at a 95 percent con-fidence level that DNB will not occur and is chosen as an appropriate ( ' p-margin to DNB for all operating conditions. . I The ' curves of Figures 2.1-1, 2.1-2, 2.1-3 and 2.1-4 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and ' maximum cold leg temperature of various pump combinations for which the ' minimum DNBR is no less than 1- W for the family of axial shapes and corresponding radial peaks shown in Figure B2.1-1. The limits in Figures AMl 2.1-1, 2.1-2, 2.1-3 and 2.1-4 were calculated for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant inlet temperature is not a safety limiti however, operation above 5B0*F is not possible because of the actuation of the main steam line , safety valves which limit the maximum value of reactor 1 et temperature. r Reactor operation at THERMAL POWER levels higher than
- of RATED THERMAL POWER is . prohibited by the high power level trip setpo' t specified in L
^)g ll0 :
V CALVERT CLIFFS.- UNIT 1- B 2-1 Amendment No. 3 3 ? h.. j r l
.: 1 a'
Q - i
(~s. SAFETY LIMITS BASES Table 2.1 ,1. The area of safe operation is below and to the left of these lines. The conditions for the Thermal Margin Safety Limit curves in Figures 2.1-1, 2.1-2, 2.1-3'and 2.1-4 to be valid are shown on the figures. The reactor protective system in combination wita the Limiting Conditions for Operation, is designed to prevent any anticipated c.cmbina-tion of transient conditions for reactor coolant system temperature, pressure, and THERMAL F0WER level that would result in a DNBR of less thantvijandprecludetheexistenceofflowinstabilities. l23 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ('N- The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. - The reactor pressure vessel and. pressurizer are designed to Section III,1967 Edition, .of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.7, Class I,1969 Edition, which permits a maximum
. transient pressure of 110% (2750 psia) of component design pressure.
. . The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation. O B ?-3 Amendment No. 33, 3 9 CALVERT CLIFFS'- UNIT 1
h 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETP0lflTS The Reactor Trip Setpoints speci.fied in Table 2.2-1 are the values at which the Reactor Trips are-set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor ccolant system are prevented from exceeding their safety limits. Operation with a trip. set less conservative than its Trip Setpoint but within its spect-fied Allowable Value is acceptable on the basis that the difference between the trip setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses. Manual Reactor Trip The Manual Reactor Trip is a redundant channel .to the automatic protective instrumentation channels and provides manual reactor trip capability. , P_ower Level-High The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal ~ Margin / Low Pressure trip. The Power Level-High trip setpoint is operator adjustable and can be set no higher than 10% above the indicated THERMAL POWER level. Operator action is required to increase the trip setpoint as THERMAL F0WER is increased. The trip setpoint is automatically decreased as THERMAL pcwer l decreases. The trip setpoint has a maximum value of 107.0", of RATED THERMAL POWER and a minimum setpoint of 30% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual ste -state THEPNAL POWER level at which a trip would be actuated is f RF THERMAL F0WER, which is the value used in the safety analyses. } ; i f. Reactor Coolant Flow-Low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant flow. Provisions have been made in the reactor protective system to permit .
.a* ..e CALVERT CLIFFS - UNIT 1 B 2-4' Amendment No. 3 9
e . (~.. LIMITIttG SAFETY SYSTEM SETTINGS BASES operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. The low-flow trip'setpoints and Allowable
~
, Values for the various reactor coolant pump combinations have been derived in consideration of instrument errors a response tinLes M [ .23 equipment involved to maintain the .DNBR above er normal operation I and expected transients.. For reactor operation with only two or three reactor coolant pumps operating, the Reactor Coolant Flov-Low trip set-points, the Power Level-High trip setpoints, and the Thermal Margin / Low Pressure' trip setpoints are automatically changed when the pump condition selector switch is manually set to the desired two- o. three-pump position. Changing these trip setpoints during two and three pumo operation prevents the minimum value of DNBR from going below Y.19 during l normal operational transients and anticipated transients when only two or
, three reactor coolant pumps are operating.
$ Pressurizer Pressure-High The Pressurizer Pressure-High trip, bacFed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safoty valves and its concurrent operation with the power-operatec' relief valves avoids the undesirable. operation of the pressurizer code safety valves.
Containment Pressure-High The Containment Pressure-High trip provides assurance that a reactor trip -is initiated concurrently with a safety injection. The setpoint for this trip is' identical to the safety injection setpoint. Steam Generator Pressure-Low The Steam Generator Pressure-Lcw trip provides protection against an excessive rate of heat extraction from the' steam generators _and p1D subsequent cooldown'of the reactor coolant. The setting of 0 sia is sufficiently below the full-load operating point of 850 ps a so as not to interfere with normal operation, but still high enough to
'f provide the required protection in the event of excessively high steam ,
flow. This setting was used with an uncertainty factor,of 1 22 psi
.in'the accident analyses.
. CALVERT. CLIFFS - UNIT 1 B 2-5 Amendment No. 33
~~~
m .. 4 R p LIMITING SAFETY-SYSTEM SETTINGS . l BASES -
..l 1
Steam Generator Water Level The Steam Generator Water Level-Low trip provides core protec.tfon by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the. pressure of the ' reactor coolant system will not exceed its
' Safety Limit. The specified setpoint provides ' allowance that there will be sufficient' water inventory'in the steam generators at the time of trip' to provide a margin of more than 13 minutes befor~e auxiliary feedwater'is required.
' Axial Flux Offset l1L 3 The axial flux offset trip is provided to e ure that excessive axial peaking will not cause fuel damage. Th axial flux offset is gr%
determined from the axially split excore de ctors- The trip setpoints ensure that neither a ONBR of less .than 1 nor a peak linear heat. rate which corresponds to the temperature for fuel centerline melting will exist'as a consequence of axial' power ma1 distributions. These trip set-points were derived from an analysis of many axial power shapes with allowances for instrumentation inaccuracies and the uncertainty associated with the excore to incore axial flux offset relationship. . Ther' mal Marcin/ Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBRLis less thanh 1:2_3
.The trip is initiated whenever the reactor coolant system pressure signal. drops-below either 1750-psia or a computed value as described below..whichever is higher. ; The ccmputed value is a function of the
' higher of.aT power or neutron. power, . reactor inlet temperature, and the number of-reactor" coolant pumps operating. :The minimum value of . reactor 1
coolant flow rate. the maximum AZIMUTHAL POWER TILT and the maximum CEA ' deviation permitted for continuous operation are. assumed in the genera-tion irf this trip-function.' In addition, CEA group sequencing in accor-dance with' Specifications '3.1.3.5_ and 3.1.3.6 is assumed. Finally, the
" maximum insertion.of CEA' banks which can occur during any anticipated operational occurrence'priorito a Power Level-High trip is assumed.
(lb4L ,
' . 'I
.. CALVERT Ci.IFFSL- UNIT 1 -B 2-6 Amendment No.133,39 a -
LIti! TING SAFETY SYSTEM SETTIrlGS BASES M The Thermal Margin / Low Pressure trip setpoints 2rc dcr ;cd 'rc -the acc i
- P fa+y 14 4t* thenu? :pplictier Of ppr;priate allowances for equipment g
- fc ty 4 response maroin h pretime,
. i dcdmeasurement
,,hich 'ncluder - uncertainties 2- :lle"nnco f and processing Z vi RA u.D errorg"THER"A to ;;;per.nt; for potun:.iul n.c :=:ure-a-t cr e ; 2r 21!cw2rce ^# M .~. ts 40 + - os m h pneanti'l allowance of h psia to com+c pcsate r; tur;for-pcenure
...eo2 u n.mmcrurcrent L vow r u bserrer,
, 42.4 trip--
a further
-tyctcm prcccc:ing cr e , ime delay associated with providing effective tennination of the occurrence that exhibits the most rapid decrease in margin to the safety limit. ine 64 p>>o allcwance ir -2de up of a ?? pci; pie 5 cure-mcrurcment :llcwance end a 62 ps.o tim; dclay 2!!cw;r.cc, 7
___ 4 ~ Asymmetric Steam Generator Transient Protection Trip Function (ASGTPTF) N , The ASGTPTF utilizes steam generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure b4% % tw exceeds the trip setpoint. The ASGTPTF is designed to provide a reactor trijb- ic-for those Anticipated Operational Occurrences associated with secondary system malfunctions which result in asymmetric primary loop coolant temperatures. The :,ost limiting event'is the loss of load to one steam generator caused by a single M&in Steam Isolation Valve closure. { The equipment trip setpoint an$ allowable values are calculated to account for instrument uncertainties, and will ensure a trip at or before
\ s reaching the analysis setpoint.
Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. tio credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability cf the Reactor Protection System. i l Rate of Change of Power-High The Rate of Change of Power-High trip is provided to protect the core l during startup operations and its use serves as a backup to the administratively enforced startup rate limit. Its trip setpoint does not correspond to a l Safety Limit and no credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System. CALVERT CLIFFS - UNIT 1 B 2-7 Amendment No. 39
* ~ ~ ' ~
t@ 3/4.1 REACTIVITY C0tiTROL SYSTEMS D 9 3/4.1.1 B01ATIC'i C0fiTROL * * " ' " SHUTDOWN t%1GIli - T,y, > 200*F - LIMITIflG CONDITION FOR OPERATION I 4.37, # 3.1.1.1 The SHUTDOWN MARGIN shall be > M ak/k. _ l APPLICABILITY: MODES 1, 2*Y 3 and 4. 8
' ACTION:
4.3k With the SHUTCOWN MARGIN < h ak/k, immediately initiate and continue l boration at > 40 gpm the required SHUTDOWN MARGIN of istits [ ppm boric acid solution or equivalent restored.
! C23Oo SURVEILLANCE RE0VIREMENTS 4.1.1.1.1 h3D The SHUTD0hti MARGIN shall be determined to be >.46% ak/k: l
. a. Within one hour after detection of an inoperable CEA(s) anr* at least once per 12 hours thereafter while the CEA(s) is inograbic.
If the inoperable CEA is irm:cyable or untrippabie, the above required SHUTDCWN MARGIN shall be increased by an amount at Lleast equal to the withdrawn worth of the immovable or untrippable > CEA(s). j b. When in MODES 1 or 2 , at least once per 12 hours by verifying that CEA group withdrawal is wichin the Transient Insertion Limits of Specification 3.1.3.5.
- c. ' When in MODE# 2 ~a , within 4 hours prior to achieving reactor criticality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.
- d. Prior to initial operation'above 5'; RATED THER'4AL PCWER af ter each fuel loading, by consideration of the factors of e below, with the CEA grcups at .the Transient Insertion Limits of Specification 3.1.3.6.
l- ** See Special Test Exception 3.10.1. With Keff > 1.0.
- With Keff < 1.0.
D *)dhuence. fa Tichnied 9>ee:Fiedian. 3/3.1, o.s spc;fial in Sarvillme Atftirements +.t.I.I.I anures thi there i.s suffic;ed avslabic shddm naryin
& rc eeac.d.s . af k safery anstyse: thm he:si fje w.s
' to utch CALVERT +he shuthwn CLIFFS-UNIT 2 ma7 I 3/411 Amer Jment No. 9,18
b 'F _RfA_CTIVITY CONTROL SYSTEf1S SHUTD0Ull itARGirl - T < 200'F avo I LIMITit!G C0flDITI0ft FOR OPERATI0tl / 3.1.1.2 The SHUTDOWftf1ARGIll shall be ,> .0,ok/k. APPLICABILITY: MODE 5. 4 ACTI0ft: $? With the SHUTC0W'l MARGIN . .., Ak/k, immediately initiate and continue boration at > 40 opm of ~ ppm boric acid solution or equivalent until the required SHUTD0'.'N !%< GIN is restored. M SURVEILLAftCE REQUIREidENTS ,
... s (O 4.1.1.2 The SHUT 00Wil MARGIN shall be determined to be f 1.
ak/k:
- a. Within 0: aur after detection of an inoperable CEA(s) and at least once per 12 hours thereaf ter while the CEA(s) is inoperable. 1 If the inoperable CEA is im:r.ovable or untrippable, the above I required SHUTDOWN MARGIll shall be increased by an amount at i least equal to the withdrawn worth of the immc able or untrippable~ l CEA(s). ,
- b. At least once per 24 hours by consideration of the following l factors:
i
- 1. Reactor coolant system boron concentration,
- 2. CEA position,
- 3. Reactor coolant system average temperature,
- 4. Fuel burnup based on gross thermal energy generation,
- 5. Xenon concentration, and
- 6. Samarium concentration.
d s
.O CALVERT CLIFFS - UrlIT 1 3/4 1-3
~ ~ ~.- .. .. .
REACTIVITY C0flTROL SYSTEMS a MODERATOR TEMPERATURE COEFFICIEllT 5 LIMITIfiG C0flDITI0ft FOR OPERATI0ft 1 3 3.1.1.4 The moderator temperature coefficient (MTC) shall be:
- a. Less positive than 0.5 x 10-4 ok/k/*F whenever THERMAL POWER is < 70% of RATED THERMAL POWER,
- b. Less positi n than 0.2 x 10-4 ok/k/*F whenever THERMAL POWER is'> 70% of RATED THERMAL POWER, and
- c. ak/k/*F at RATED THERMAL Less negative than h (x 10-4 POWER.
APPLICABILITY: MODES 1 and 2*# ACTI0ft: I With the moderator temocrature coefficient outside any one of the above limits, be in at least HOT STAtlDBY within 6 hours. f ,. I
.j q
1 SURVEILLAtlCE REOUIREMEflTS i 3 6 ( 4.1.1.4.1 The MTC hall be determined to be within its limits by
's confirmatory measurements. MTC measured values shall be extrapolated y and/or compensated to permit direct comparison with the above limits.
1 1 l i 1 *With Keff ' I'0-
#See Special Test Exception 3.10.2.
l
~.
6- i O f CALVERT CLIFFS - UtlIT 1 3/4 1-5 i
i d AO REACTIVITY C0flTROL SYSTEMS i FLOW PATHS - OPERATING l LIMITIftG C0flDITIO:1 FOR OPERATI0ft 3.1.2.2 At least two of the following. three boron injection flow paths (j and one associated heat tracing circuit shall be OPERABLE: 1
; a. Two flow paths from the beric acid storage tanks via either a boric acid pump or a grar,ty feed connection, and a charging 1
pump to the Reactor Coolant System, and
- b. The flow path from the refueling water tank via a charging pump to the Reactor Coolant System. .
APPLICABILITY: MODES 1, 2, 3 and 4. CI ACTI0ft: 31' With only one of the above requ ed boron injection flow paths to the Reactor Coolant System CPERAc , restore at least two boron injection flow paths to the Reactor olant System to OPERABLE. status within 72 bO i hours or be in at least equivalent to at least STAfiDBY and borated to a SHUTDOWil MARGIN ak/k at 200*F within the next 6 hours; restore at least two flow paths t OPERABLE status within the next 7 days or be in COLD SHUTDOWil within the next 30, hours. i SURVEILLANCE REQUIREMENTS , d 3 4.1.2.2 At least two of the above required flow paths shall be demonstrated
.j OPERABLE:
- a. At least once per 7 days by verifying that the temperature of
) the heat traced portion of the flow path from the concentrated 1 boric acid tanks is above the temperature limit line shown on Figure 3.1-1.
! b. At least once per 31 days by verifying that each valve (manual, d
l power operated or automatic) in the flow path that is not locked, j sealed, or otherwise secured in position, is in its correct position.
- c. At least once per 18 months during shutdown by verifying that 3
each automatic valve in the flow path actuates to its correct
; position on a SIAS test signal.
do , CALVERT CLIFFS - UNIT 1 3/4 1-9
.,,,.-c----
r1
0 REACTIVITY CONTROL SYSTEMS CHARGING LUMPS - OPERATING 1 LIMITING CONDITION FOR OPERATION 3.1.2.;1 At least two charging pumps shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4.
; /
ACTIO1: - 3o Wi.h only one charging pump OPERABLE, restore at least two arging pumps to OPERABLE status within 72 hours or be in at least OT STANDBY and borated to a SHUT 00WN MARGIN equivalent to at least ' ak/k at 200*F within the next 6 hours; restore at least two charging pumps to OPERAELE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. i O g B 3 SURVEILLANCE REQUIREMENTS 0 h
.; 4.1.2.4 . No additional by Specification 4.0.5.Surveillance Requirements other than those required 3
a 3 ( s b a
! 0 -
tfj CALVERT CLIFFS - UNIT 1 3/4 1-11
- , , , - - , w-
._ ,r - - - -
p--- - - g, -M-- '",',"% =F' ~-' -"" F'"
- "-" a 7 =rg* ,'.- "T -."C,6I'M C+ "
1 REACTIVITY C0flTROL SYSTEMS BORIC ACID PUMPS - OPERATItiG v LIMITING CONDITIO!f FOR OPERATI0tl 1 3.1.2.6 At least the boric acid pump (s) in the baron injection flow path (s) required OPEPABLE pursuant to Specification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if
~l the flow path through the boric acid pump (s) in Specification 3.1.2.2a i is OPERABLE.
APPLICABILITY; MODES 1, 2, 3 and 4. . ACTION: With one boric acid pump reauired for the boron injection flow path's) pursuant to Specification 3.1.2.2a inoperable, restcre the boric acid
. pump to OPEPABLE status within 72 hours or be in at least HOT STAtiDBY within thegext 6 hours and borated to a SHUTDC',!N MARGIN equivalent to at least (j ak/k at 200*F; restore the above required boric acid pump (s) to OPEPAB E status within the next 7 days or be in COLD SHUTD0i!N within the next 0 hours.
V .
.1
=
bi ,
. 1
; SURVEILLANCE REOUIREMENTS l
. : I f
7 4.1.2.6 No additional Surveillance Requirements other than those required ! g' by Specificatien 4.0.5. p 1
? .
M - e m CALVERT CLIFFS - UNIT 1 3/4 1-13 - g' w - v* w -
i *
) i 7'
R_EACTIVITY C0!! TROL SYSTEMS - l BORATED WATER SOURCES - SHUTDOUit d
~
LIllITIftG C0flDITI0tl FOR OPERATI0ft , 1 a 0 3.1.2.7 As a minimum, one of the following borated water sources shall j be OPEPABLE:
.e a. One boric acid storage tank and one associated heat tracing circuit with the tank contents in accordance with Figure 3.1-1.
- b. The refueling water tank with:
4)
- 1. A minimum contained borated water volume o 9,9 gallons, l
- 2. A minimum boron concentration o an p
- 3. A minimum solution temperature of 35'F.
,] APPLICABILITY: MODES 5 and 6.,
h ' ACTIO!!:
";; With no borated water sources OPERABLE, suspend all operations involving CORE ALTEPATIO:ts or positive reactivity changes unti.1 at least one
.i _ borated water source is restored to OPERABLE status. -
9
) SURVEILLA! ICE REQUIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated 1 OPEPABLE: -
t
- a. At least once per 7 days by:
-' 1.
Verifying the boron concentration of the water. i 2. Verifying the contained borated water volume of the tank, l and s
-t. 3. Verifying the boric acid storage tank solution ~ temperature -
.: when it is the source of borated water. .
- 4 i '
- b. At least once per 24 hours by verifying the RWT temperature when it is the source of borated water and the outstde air temperature is < 35'F.-
- CALVERT CLIFFS - UNIT 1 . 3/4 1-14 Amendment No. 21 eW - w 7
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l' CALVERT CLIFFS - UNIT 1 3/4 1-15 Amendment No. 21 i 4
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REACTIVITY CONTROL SYSTEMS - .- .._. sa BORATED WATER SOURCES - OPERATIRG .g 3 RE
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LIMITING CONDITION FOR OPERATION 5.E
~
3.1. !! . 8 At least two of the following three borated water sources shall ]5 be 0)ERASLE: _
=5'
- a. Two boric acid storage tank (s) and one associated heat tracing @@
~
circuit per tank with the contents of the tanks in accordance EE with Figure 3.1-1 and the boron concentration limited to < St, @[ and =_=
==.
ti The refueling water tank with: ag b.
- 1. A minimum contained borated water volume of 400,000 gj gallons, 7 7,, ,gg,, .;-
- 2. A beron concentration of betwaen SSGG and -:::: gpm, l
- 3. A minimum solution temaerature of 40*F, and fl-
'f
- 4. A maximum solutien temperature of 100*F in MODE 1. ..
.=5 APPLICASILITY: MODES 1, 2, 3 and 4. . . .
g ACTION: 3fo 5 With only one borat water source OPERAELE, restore at least two borated water sources to ERABLE status within 72 hours or be in at least HOT = STANDBY with' .he next 6 hours and borated to a SHUTDOWN KARGIN ecuivalent to at least i ok/k at 200*F; restore at least two borated water sources E i; to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours. [' E
. Ub SURVEILLANCE REOUIREMENTS =
=
'i -
4.1.2.8 At le st two borated water sources shall be demonstrated OPERABLE:
.=
=
- a. At least once per 7 days by:
- 1. Verifying the boron concentration in each water source, g
- 2. Verifying the contained borated water volume in each wat6r source, anc I}
1
- 3. Verifying the boric acid storage tank solution temperature. -
- b. At least once per 24 hours by verifying the RWT temperature gEE when the outside air temperature is < 40 F.
3/4 1-16 Amendment Nc. 27,3 5 CALVERT CLIFFS - UNIT 1
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V. CALVERT CLIFFS - UttIT $ 3/4 2-4 Amendment tio. 9,18 x> .t i,
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s =_=gs _..i=_ =_te_. =. .,tgt;=i=. . as_ =m. _ _ a. = --i= .- 0 20 40 60 80 100 120 140 CORE HEIGHT, INCHES F16 42-1 MGg W4 2-5 . 1 1 1 1 i BALTIMORE Figure GAS & ELECTRIC CO. AUGMENTATION FACTOR
-Calvert Cliffs VS ' Nuclear Power Plant DISTANCE FROM BOTTOM OF CORE 4.2-1
m P POWER DISTRIBUTI0ft LIMITS TOTALPLAtlARRADIALPEAK!ftGFACTOR-F[y LIMITIltG C0!!DITI0ft FOR OPERATIO!! T T 3.2.2 The calc ted value of F*Y, defined as F*Y =F xY(1+T 9 ), shall be limited to <_ ).7 00 l APPLICABILITY: MODE 1*. OO ACTI0ft: \.1 . WithFfy> ithin 6 hours either: [
- a. Reduc 9 THERMAL POWER to bring the ccmbination of THERMAL POWER and F to within the limits of Figure 3.2-3 and withdraw the full Nngth CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6; or
- (, b. Be in at least HOT STAfiDBY.
i SURVEILLAftCE RE0VIREMEttTS
?
l 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. l 4.2.2.2 Ffy shallbecalculatedbytheexpressionFfy=Fxy(1+T)andFfy q shall be detennined to be within its limit at the following intervals: l
- a. Prior to operation above 70 percent of RATED THERMAL POWER ^
after each fuel loading,
- b. At least once per 31 days of accumulated operation in MODE 1, -
and
- c. Within four hours if the AZIMUTHAL POWER TILT (T q ) is > 0.030.
1
*See Special Test Exception 3.10.2.
.O I CALVERT. CLIFFS - UNIT-1 3/4 2-6 Amendment No. 27, 2.4, 32
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s .
-f e CALVERT CLIFFS-UlGT 1 3/4 2-5 Amendment tio. 9, M ,18 i
L
7_ , P-
, POWER UISTRIBUTIOll LIMITS TOTALINTEGRATEDRADIALPEAKI!!GFACTOR-Ff LIMITING C0ilDITI0il FOR OPERATION T T 3.2.3 The cal ' ed value of F , defined as F r
= Fr (1+Tq ), shall be lioited to 1 .571 1.~2 0 0 r l APPLICABILITY: MODE 1*.
d ACT10ti: \1 WithFf>.57 within 6 hours either: g
- a. Be in at least HOT STANDBY, or
- b. Reduc and to within F[y the limits POWER THERIGL of Figure 3.2-3 to and bring withdraw the the full combination lengt1 CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6. The THEFJiAL POWER limit
( .O determined from Figure 3.2-3 shall then be used to establish a revised upper THEFJiAL POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction of RATED THEFliAL F0WER determined by Figure 3.2-3) and subsequent operation shall be maintained within'the reduced acceptable operation region af Figure 3.2-4. SURVEILLAtlCE REQUIREMENTS
'~
4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 Ff 'shall be calculated by the expression Ff = F (1+T q ) and Ff 7 shall be detemined to'be within its limit at the following intervals:
- a. Prior to operation above 70 percent of RATED THERMAL POWER
.after each fuel ~ loading,
- b. At.least once per 31 days of acetr.iulated operation in M00E 1,
,, and- ,
- c. Within four hours if the AZIMUTHAL POWER TILT (T q ) 1s > 0.030.
f- "See Special Test Exception 3.10.2. t .b -
]ALVERT CLIFFS - UNIT 1 '3/4 2-9 Amendment No. 27, 2A ,32, 33, 3-
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(
~ PERIPHERAL AXIAL SHAPE INDEX (Ygl , t i
FIGURE 3.2-4 DNB Axial Flux Offset Control Limits y . , 1
.l- <
l CALVERT CL"TS-UNIT .l . 3/4 2 M Amendment No. 9.I8
. 9 Q:
l }~ .
-: . n - ~ - . h ,: w _ _ ,, 1. n . _ - _
~) W_ u.:
_- n _
- . . ~ p. . -
1 ( i I
)l - . TABLE 3.3-1 J ,
9 REACTOR PROTECTIVE INSTRUMENTATION i G E
" MINIMUM TOTAL NO. CHAtlNELS CHANNELS APPLICABLE 8 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION vi 1. Manual Reactor Trip 2 1 2 1, 2 and
- 1
, g 2. Power Level - High 4 2 3(f) .,2 2#
1 [ 3. Reactor Coolant Flow - Low 4/SG 2(a)/SG 3/SG 1, 2 (e) 2# j
! 4. Pressurizer Pressure - High 4 _ 2 3 1, 2 2#
- S. Containment Pressure - High 4 2 3 1, 2 2#
3
- 6. Steam Generator Pressure - Low 4/SG 2(b)/SG 3/SG 1, 2 2#
- 7. Steam Generator Water
[, $ Level - Low 4/SG 2/SG 3/SG 1, 2 2#
,Y 8. Axial Flux Offsct 4 2(c) 3 1 2#
- 9. Thermal Margin / Low Pressure 4 =
2(a) 3 1, 2 (e) 2# 1 9a. team Generator Pressure l Difference - High 4 2(a) 3 1, 2 (e) 2#
- 10. Loss of Turbine--Hydraulic '
Fluid Pressure - 1.ow 4 2(c) 3 1 2# 4 - i A b b ,
s.
< l N
i . 1 4
* ; i s
g ., TABLE 3.3-1 (Continued) , 2 ' TABLE NOTATION *
\
r With the protective system trip breakers in the closed position and 1 the CEA drive system capable of CEA withdrawal. The provisions of Specification 3.0.4 We not applicable. ')N (a) Trip may be bypassed below 10-4 of RATED, THERMAL POW f beautomaticallyremovedwhenTHERMALPOWERis>10-[R;bypassshal of RATED - THERMAL POWER. g (b) Trip may be manually bypassed below psia; bypass shall ting automatically removed at or abov psi - (c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% cf.
't
~
RATED THERMAi. POWER. ' (d) Trip may be bypassed below 10-4% and above 121of RATED 1 RMAL POWER. , ( if ;i 1 e + g (e) Trip may be bypassed during testing pursuant,to 'Special Test Excep-( tion 3.10.3. . 2 (f) There shall be at leastilo decades of overlap between the Wide
; Range Logarit?"nic Neutrots Flu;: Honitoring ChEnnelsLand'the Power Range Neutron Flux Monitoring Channels. f* i ,
' ( 1 .
w ii ACTION STATEMENTS 1 \ L
.f ACTION 1 -
With the number of channels OPERABLE one led than q required by the Mininium Channels OPERABLEUreppitement, y restore the inoperable channel to OPER4BCE status within C
'48 hours or be in HOT STANDBY within the nldt.6 hours I apid/or
, open the protective system tri}i breafsis.
ACTION 2 - With the number of OPERABLE channels one -lesh than the
, Total Number of Channels, STARTUP and/or POVER G?ERATION s
I may proceed provided the following conditions are satisfied:
- a. The inoperable channelkis placed in either the bypassed
- or tripped condition within 1 hour. For the purposes i g
-of testing and mdintenance, the inoperable , channel may r be bypassed-for'up ta 48 hours from time of initial loss E
] of OPERABILITY:. biwever, the inoperable channel shall
! then be either 'r;estored to OPERABLE status Er placed in the tripped conditiond
,h.
p LCALVERT CLIFFS - UNIT 1 ' t 3/4 3-4' W n. -.- , .
-. - A. s
.{yfap - ,
, ,. ,. - .. , - e
3
%{
S$Q L'$ p n- .
/v TABLE 3.3-2 w
"' # , *O -"
~'
j j REACTOR PROTECTIVE INSTRt)MEtlTATI0O RESP 0.'_1SE TIMES FUtlCTIONAL UNIT -RESP 0NSE TIME j 11. . Manual Reactor Trip - flot Applicable
- 2. Power Level - !!igh ' .< 0.40 seconds *# and < 8.0 seconds ## '[ ,
a 3. Reactor Coolant- Flow - Low 1 0.50 seconds
- 4. -Pressurizer Pressure - High 1 0.90 seconds
- 5. Containment Pressure - High 1 0.90 seconds
'6. -Steam Generator' Pressure - Low 1 0.90 seconds .
- 7. Steam Generator Water Level - Lbw 1 0.90 seconds b 8. Axial Flux Offset
~
< 0.40 seconds *# and < 8.0 seconds ##
- 9. Thermal Margin / Low Pressure ggp < 0.90 seconds *# and < 8.0 seconds ##
eam Generator Pressure Difference - High 1 0.90 seconds
- 10. Loss of Turbine--Hydraulic fluid Pressure - Low flot Applicable E 11. Wide Range Logarithmic Neutron Flux Monitor tiot Applicable in
- *tleutron detectors are exempt from response time testing. Response time of the neutron flux signal portion
- of the channel shall be measured from detector output or input of first electronic component in channel.
# Response time does not include contribution of RTDs.
- ##RTD response time only. This value is equivalent to the time interval required for the RTDs' output a to achieve 63.2%.of its total change when subjected to a step change in RTO temperature. .
n TABLE 4.3-1 k . REACTOR PR'OTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS R q h , CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL. SURVEILLANCE m FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED e lg- 1. Manual Reactor Trip H.A. N.A. S/U(1) N.A. [ 2. Power Level - High
- a. Nuclear Power S .D(2),H(3),Q(5) M 1, 2
- b. AT Power S ,
D(4),R M 1
- 3. Reactor coolant Flow - Low S R M 1, .4.
Pressurizer Pressure - High S R M 1, 2
- 5. Containment Pressure - High S R M '1, 2 w ~
O 6. Steam Generator Pressure . Low S R M 1, 2
-7. Steam Generator Water Level - Low S R M 1, 2
- 8. Axial Flux Offset S R M 1
- 9. Thermal Margin / Low Pressure S R M. 1, 2 9a. Steam Generator Pressure Difference -
, .High S R M 1, 2
- 10. Los f Turbine--Hydraulic Fluid Pres e - Low N.A. N.A. S/U(1) N.A.
NDD
i h'. . l TABLE 3.3-3 (Continued) l TABLE NOTATION i O (a) Trip function may be bypassed in this it0DE when pressurizer
; pressure is < 1700 psia; bypass shall be automatically removed when pressurizer pressure is > 1700 psia.
( li (c) Trip function may be bypassed in this MODE b ow 0 psia; bypass shall be automatically removed at or above 0 ps .- i G5 The provisions of Specification 3.0.4 are not applicable. I ACTION STATEMENTS 4 ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 houro and in COLD SHUTDOWN within the following 30 hours. ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided ,_ the following conditions are satisfied: 1 a. The inoperable channel is placed in either the bypassed j or tripped condition within 1 hour. For the purposes j of testing and maintenance, the inoperable channel may a ' be bypassed for up to 48 hours from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or
; placed in the tripped condition.
9 y b. Within one hour, all functional units receiving an input i from the inoperable channel are also placed in the same i condition (either bypassed or tripped, as applicable)
) as that required by a. above for the inoperable channel.
; c. The Minimum Channels OPERABLE requirement is met;
. however, one additional channel may be bypassed for up
; to 48 hours while performing tests and maintenance on j that channel provided the other inoperable channel is q placed in the tripped condition.
J i O AO j l CALVERT CLIFFS - UNIT 1 3/4 3-15 4
~ ~ ~ . . - - _ - - . c 7. w .,,n. , .---..7 .-,,,-,.--#"
.. ,A. A .
^ .
) D )
. TABLE 3.3-4 -
O {" < ENGINEERED SAFETY FEATURE ACTUATI0ft SYSTEM INSTRUMENTATI0ft TRIP VALUES ALLOWABLE
~P FUNCTIONAL UNIT TRIP SETPOINT VALUES
.- 1. SAFETY INJECTION (SIAS) e a. Manual (Trip Buttons) Not Applicable Not Applicable
] b. . Containment' Pressure - liigh 1 4.75 psig 1 4 75 psig
- c. Pressurizer Pressure - Low > 1578 psia
> 1578 psia
. .i
- 2. CONTAINMENT SPRAY (CSAS)
- a. Manua) (Trip Buttons) Not Applicable Not Applicable
- b. Containment Pressure -- liigh 1 4.75 psig 1 4.75 psig
- 3. CONTAINMEtiT ISOLATION (CIS) t# a. Manual CIS (Trip Buttons) Not Applicable Not Applicable
- b. Containment Pressure - liigh 1 4.75 psig 1 4.75 psig
{ u
- 4. MAIN STEAM LINE ISOLATION
- a. Manual (MSIV lland Switches and Feed llead Isolation
- i. Iland Switches) Not Applicable Not Applicab'le .
- 5 '7 d
- b. Steam Generator Pressure - Low > h psia
_ > ha'S'?d . .!
'l i
i (N TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS C
- 1. Manual
.f
; a. SIAS '
d SafetyInjection(ECCS) Not Applicable
- b. CSAS l Containment Spray Not Applicable
, c. CIS Containment Isolation Not Applicable
~
- d. RAS Containment Sump Recirculation Not Applicable
- 2. Pressurizer Pressure-Low a.- Safety Injection (ECCS) < 30*/30**
cn
% J j 3. Containment Pressure-High 1
, a. Safety Injection (ECCS) 1 30*/30**
l- b. Containment Isolation 1 30
- c. Containment Fan Coolers 1 35*/10**
3 j 4. Containment Pressure--Hioh
.f
- a. Containment Spray s
1 0*/60** 6
~
q 5. Contaicment Radiation-High / 6 < j a. Containment Purge Valves Isolation <6
,i C 2
M 4 . CALVERT CLIFFS - UNIT 1 3/4 3-20
= '
~. _,, .
- e
i O 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) SAFETY INJECTION TANKS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with:
- a. The isolation valve open,
- b. A contained borated water volume of between 1113 and 1179 cubic feet of borated water (equivalent to tank levels of between 187 and 199 inches, respectively),
w i n t VMuW 2300 ftY^
- c. A oron concentration of Mt:r '72: x d :2^^ yem., o.iu -
l
- d. A nitrogen cover-pressure of between 200 and 250 psig.
APPLICABILITY: MODES 1, 2 and 3.* ACTION: O
- a. With one safety injection tank inoperable, except as a result of a closed isolation, valve, restore the inoperable tank to OPERABLE stetus within one hour or be in HOT SHUTDOWN within the next 12 hours.
- b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT
. SHUTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERABLE: g a. At least once per 12 hours by:
- 1. Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and
- 2. . Verifying that each safety injection tank isolation valve
(.~ is open.
*With pressurizer. pressure > 1750 psia.
9
.CALVERT CLIFFS - UNIT'l 3/4 5-1 Amendment No.? ?
mu ._.m__ _ .
x ,. ~ y 1 g , ; .5
?. ;
W
; EMERGENCY.; CORE COOLING SYSTEMS:
. . h_ = . -~- m& REFUELING WATER TANK- ' i= ' _i. . . LIMITING CONDITION FOR' OPERATION :. a
=
EE::
=a.
3.'5.4 The refueling Water tankishall be OPERABLE with: ~._
- a. A minimum. contained. borated water; volume of 400,000 gallons, [-i
. _. 2300 '26*o EI A boron concentration of between 4+E9 and GiHM ppm, If;
~
' b. l
- c. A' minimum' water temperature of 40*F, and F.
9_. .
- d. : A maximum soiution temperature 'of 100*F in MODE 1..
APPLICABILITY: MODES 1, 2, ' 3 and o. ;;; ACTION: With the refueling water tank' inoperable, . restore the tank to OPERABLE status within 1 hour or be~_in at least HDT STANDBY within 6 hours and in - COLD SHUTDOWN within the fo11owing- 30 hours. --
--.=
'a SURVEILLANCE REOUIREMENTS- -
-91 i:
~
4.5.4 The RWT shall be demonstrated OPEPJ.BLE: l
- a. At least once per 7 days by: ;-j ..
t
- 1. Verifying the contained borated water volume in the tank, and - 3;q
=5
'2. Verifying the_ boron concentration of the water. 5 .
- b. At-least once per 24 hours- by verifying the RWT temperature when the outside air temperature is < 40*F. ' h-f h
y Er;
=_
! :s
. o
, :7
;_.=.
..=._.
CAI.VERTlCLIFFS :- UNIT 1 Amencment No. 17, M i 3/4 5-7
'a
. _1 @D ' SD d b 0
@ u E- '
j
( . 3/4.9 REFUELIrlG OPERATI0ils 1 BORON CONCErlTRATION 9 4 LIMITING CONDITI0tt FOR OPERATION a i 3.9.1 With the reactor vessel head unbolted or removed, the borca concentration of all filled portions of the Reactor Coolant System ani
- the refueling pool shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met:
- a. Either a K of 0.35 or less, which includes a h k/ conser-vativeal18Nnceforuncertainties,or 2300
- b. A baron concentration of > 672 ppm, which includes a 50 ppm conservative allowance for UTFcertainties.
APPLICABILITY: MODE 6*. ACTION: With the requirements of the above specification not~ satisfied, immediately (,* suspend all operations involving CORE ALTERATIONS or positive reactivity 2. 3 M Y changes and initiate and continue boration at > 40 gpm of 20 pm boric acid solution r its equivalent until K ~ is reduce to < 0.95-or the baron concentration is restored to 1 e ppm, whichever is the more restrictive. The provisions of Spec fication 3.0.0 are not applicable. 130c W SURVEILLANCE REOUIREMENTS 3 4.9.1.1 The more restrictive of the above two reactivity conditions
) shall be determined prior to:
a b Removing or unbolting the reactor vessel head, and a. j b. Withdrawal of any full length CEA in excess of 3 feet from its fully inserted _ pcsition. 4.9.1.2 The baron concentration of the reactor coolant system and the 4 refueling pool shall be determined by chemical sneiysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours. t 0 ine reector she>> be meinte4ned in n00e 6 when the reector vessei heed is unbolted or removed. 1 CALVERT CLIFFS - UNIT 1 3/4 9-1
, - ~ ~
. , . ~ .-..c
( 3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN ; f LIMITING CONDITION FOR OPERATION 3.10.1 :The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity'equiva-lent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s). APPLICABILITY: MODE 2. ACTION: ,
- a. With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion 23onb immediately initiate and continue boration at > 40 gpm of ppm boric acid solution or its equivalent untiT the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
- b. With all full length CEAs inserted and the reactor subcritical
.. by less than the above reactivity equivalent, imrediatelv _236~5 ' ( I") . initiate and continue boration at > 40 gpm of(3235 ppm boric acid solution or its equivalent until 'the SHUTDOWN MARGIN re-quired by Speci'iration 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The ' position of each full length CEA required either partially l or fully withdrawn shall be determined at least once per 2 hou s. , 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50*; withdrawn position with-in 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. i]fi CALVERT CLIFFS - UNIT 1 3/4101 Amendment No. 32
(C 3/4.1 REACTIVITY C0tiTROL SYSTEMS BASES 3/4.1.1 BORATIO!! C01 TROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWit MARGIff A sufficient S!4UTDOWri MARGIt! ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticalitv in the shutdown condition.
-StSW0'2" "PGI" "e ;e"ren ts fvc c W c.;;h:Ut c:": P fc ;; ; fat ^r i ja. vi iuel depletien, 0:0 bar:n :: an . Th: .T:<st restrict 1/0 ccnditicr :::ur: c.t :yecion, and RC5 n; T .3d :" r t'r.g pg' t e m pe r:.t u r: , ;o d o n5t.i i d an* re"'I tin: ua:Ont-M ad RR d.N hn -
pc:tu
,:with d st:I'N:ai r li x P d acc4 dent in tS: encip o vi U.i; c;;id:nt, n mi n i r~ m quGTnnw p;om nf 6 A ;'Jk 1s initially requirea to concrul l tb2 "Or t4"ity t"2"-ia-t- Accordingly, the SHUTDOWil MARGIll requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T < 200 F, the reactivity transients result-(D ing from any postulated accidEt2 ~re minimal and a@ak/k shutdown margin provides adequate protection. N 3/4.1.1.3 BORON DILUTIC!l A minimum flow rate of at least 3000 GpM provic'es adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during bcron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 9,601 cubic feet in approximately 24 minutes. The reactivity change rate anscciated with boron concen-tration reductions will therefore be within the capability of operator recognition and control.
3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIEtiT (MTC) The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC duringdach fuel cycle are adequate to confirm the MTC value since thi.s coefficient changes slowly due principally to the reduction in RCS baron concentration associated with fuel burnup. The confinnation that the measured itTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle. o CALVERT CLIFFS - UtlIT 1_ B 3/4 1-1 Amendment No. 27, 32
J INSERT for Page B 3/41-1 . Shutdown margin requirements vary throughout core life as_ a function of fuel depletion, RCS boron concentration and RCS Tavg. The minimum available shutdown margin-for no load operating conditions at beginning of life is 4.1% ak/k and at end of life is 4.3% Ak/k. The shutdown margin-is based on the safety analyses performed for a steam line rupture event initiated at no-load conditions. The most restrictive steam line rupture event occurs at E0C conditions. For the steam line rupture event at beginning of cycle conditions, a minimum shutdown margin ~of less than 4.1% Ak/k is required to control the reactivity transient, and end of cycle. conditions require 4.3% ak/k. 1 j l l 1 1
9
.o REACTIVITY CONTROL- SYSTEMS BASES 3/4.1. 5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 515'F. This Limitation is required to ensure 1) the moderator temperatur.t coefficient is within its analyzed temperature range, 2) the protective
~
instrumentation is within its normal operating range, 3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 4)_the . reactor pressure vessel is above its minimum RT NDT temperature. 3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is,available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging' pumps,-3) separate flow paths, 4) boric acid punps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators. (O With the RCS average temperature above 200*F, a minimum of t.,o separate and redundant boron injection systems are provided to ensur_ single functional capability in the event an assumed failure renders one of the systems -inoperable. Allowable out_of-service periods ensure that minor. component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures bg() during'the repair period. ef y,1 9 01, The boration capability of either system is s ent to provide a SHUTDOWN MARGIN from all operating conditions of .0 ak/k after xenon decay and cooldown to 200*F: The maximum boratio capability requirement 1 71' g
- rs at EOL from full power equilibrium xenon conditions and requires w*b > ~
N 381 lonsLof 7.25% 'c' acid solution from the boric acid tanks or ,20' gallons of ppm barated water from the refueling water tank. owever, to be nsistent with the ECCS requirements, the RW7 is required to have a mini mum contained volume of 400,000 callons during MODES 1, 2, 3 and 4. 0 C 730
.With the RCS temperature below 200*F, one injection system is )
acceptable without single failure consideration .on the basis of the I stable reactivity _ condition of the reactor and the additional restric- l tions prohibiting CORE. ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. , , l
-o i CALVERT CLIFFS.- UNIT 1 B 3/4 1-2 Amendment No. 21 )
i s
r .
)
REACTIVITY C0flTROL SYSTEMS U BASES _ f. The boron capability required below 15' .
?00*F is based upon providing a i " ak/k SHUTDOWN MARGIN after xenon deca 'and cooldown from 200'F to E" boric acid 3 40*F. This condition requires eithe- 3 gallons of 7.
solution from the boric acid tanks or ,, 78 gallons of 72 ppm borated water from the refueling water tank. w ;k g30 0
', The OPERABILITY of one boron injection syltem during REFUELING ensures that this system is available for reactivity control while in MODE 6.
[
- 3/4.1.3 MOVABLE CONTROL ASSEMBL!ES The specifications of this section ensee that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA' ejection accident are limited to acceptable levels.
The ACTION statements which pemit limited variations from the basic 2 requirements are accompanied by additional restrictions which ensure that
- y the original criteria are met.
The ACTI0tl statements applicable to a stuck or untrippable CEA and to a large misalignment (> 15 i'nches) of two or more CEAs, require a 2 prompt shutdown of the reactor since either of these conditions may be 9 indicative of a possible loss of mechanical functional capability of the CEAs and in the event of-a stuck or untrippable CEA, the loss of SHUT-DOWN MARGIN. i For small misalignments (< 15 inches) of the CEAs, there is 1) a small degradation in the peakiiig factors relative to those assumed in l generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a
,i small effect on the time dependent long term power distributions rela-
' tive to those used in generating LCOs and LSSS setpoints for DNBR and ,
.; linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the ejected CEA worth used in the safety :
analysis. Therefore, the ACTION statement associated uith the small l misalignment of a CEA permits a one hour time interval during which 4 attempts may be made to_ restore the CEA to within its alignment require-
.' ments prior to initiating a reduction in THERMAL p0WER. The one hour time limit is sxfficient to (1) identify causes of a misaligned CEA, (2) i l
take appropriate corrective action to realign the CEAs and (3) minimize *
-1 the effects of xenon redistribution.
kaendment No. 21 I CALVERT CLIFFS - UNIT 1 B 3/4 1-3
. - _ - _ _ _. . . _ . ..- .~ .- -~ _ , - - - . .
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POWER DISTRIBUTI0il LIMITS BASES the analysis establishing the Drib Margin LCO, and Thermal Margin / Low PreuWe LSSS setpoints remain valid during7o per9 tion at the various able CEA grcup insertion limits. If-F F or T exceed their basic limitations. operation may continue uf2e,r [he ad8itional restric-tions imposed by.the ACTIC!i statements since these additional restric-tions provide adequate provisiens to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin /Lew Pressure and
~
Local ~ Power Density - High LCOs and LSSS setpoints remain valid. An AZIMUTHAL POWER TILT'> 0.10 is not expected and if it should occur, sub-sequent operation would be restricted to only those operations required to' identify the cause of this unexpected tilt. T JhevalueofT that must be used in the equation F*7 = F*Y (1 + T9 ) and F r =F r (1+Tq ) 95 the measured tilt. T The surveillance requiremen.ts for verifying that Ff ,, F 7 andTq T aye within their limits provide assurance that the actualTv ucs 9 -F f r do not exceed the assumed values. Verifjing F and F., afGr, r and T v each9uelloadingpriortoexceeding75%ofRATEDTHEMALPOWERprovides additional assurance that the core was properly loaded. 3/4.2.4 FUEL RESIDEftCE TIME _ , This specification deleted. 3/4.2.5 Of1B PARAMETERS The limits on the DitB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in.the transient and accident analyses. The limits are consistent with the safety- analyses assumptions and hav been analytically demonstrated adequate to maintain a minimum OfiBR ( 1 / hroughcut each analyzed transient. N g. 2. 3 The 12 hour periodic surveillance of these parameters through_instru-ment readout is sufficient _to ensure that the parameters are restored within their limits following load changes and other expected transient l operation. The'18 month periodic ceasurement cf the RCS total ficw rate 1 l is ~ adequate to' detect flow degradation and ensure correlation of the flow indication channels with measured ficw such' that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour basis. , b CALVERT CLIFFS ,Uti!T I -8 3/4.2-2 Amendment tio. 27,-27, 33,
= ___
r , fr 3/4.9 . REFUELING OPERATIONS BASES 3/4.9.l' GORON CONCENTRATION po o The limitations on minimum boron concentration hppm) ensure that:
- 1) the reactor will remain subcritical during CORE ALTERATI0flS, and 2) a uniform bnron concentration is maintaincd for reactivity control in the water volumes having direct access to the reactor vessel. The limitation en K of no greater than 0.95 which includes a conservative allowme for u8brtainties, is sufficient to prevent reactor criticality during refueling operations. .
f/4.9.2 IllSTRUMENTATI0ff The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core. 3/4.9.3 DECAY TIME The minimum requirement fa,r reactor. subtriticality prior to movement
' of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short i lived fission products. This decay time is consistent with the assumptions used in the accident analyses.
3 i i 3/4.9.4 CONTAINMENT PENElRATI0 tis l 1 ! l The requirements on containment penetration closure and OPERABILITY j ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure j restrictions are sufficient to restrict radioactive material release j from a fuel element rupture based upon the lack of containment pressur- ! l ization potential while in the REFUELING MODE. J I 3/4.9.5 COMMUNICATIONS l The requirement f'or communications capability ensures that refueling i station personnel can be promptly informed of significant changes in the l facility status or core reactivity condition during CORE ALTERATIONS. I
; CALVERT CLIFFS - UNIT 1 B 3/4 9-1
- -~ .
.. - _ . _ .n- _ ,. . _ _ ., _ - . . -
p
l M. REACTOR COOLANT SYSTEM LIMITit.G CONDITION FOR OPERATI0tt (Continued) and the setpoints for the following trips.have been reduced to the. values.specified in Specification 2.2.1 for operation v,ith two reactor coolant pumps operating in the same loop:
- 1. . Power. Level-High
- 2. Reactor Coolant Flow-Low
- 3. Thermal Margin / Low Pressure
- 4. ' Axial Flux Offset .
MOM 3: Ope,gion ,na pacea pv ,. del tm nc.de calanT fo:p ee in cyuation u:th ether be o. juk one. ree etor cxiat funyL') .opedine #in owh Icap , MODES %4.*" and Sr**- Operation may proceed provided at least one reactor coolant loop is in y operation witn an associated reactor coolant pump or shutd <in coolin; pump.* The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
*All reactor coolant pumps anc shutoown cooling pumps may be de-energized .V .
for up to I hour to accommocate transition bet < teen shutcov.n ccoling como and-reactor coolant pump cperation, provided no operations are permitted which could cause dilution of the reactor coolant systen boron i concentration.
**A reactor coolant pump shall not be started <iith one or more of the RCS cold lec terperatures < 275:F unless 1) the pressurizer water volume is less thir. 603 cubic feet or 2) the secondary water temoerature of eacn steam genrator is less than 46 F (34 F v. hen measure: b;. a surface contact instrament) above the coolant temperature in the reactor vessel.
SURVEILLA.'CE REQUIREMENTS 4.4.1 The P.eactor Protective Instrumentation channels specified in the applicable - ACTION statement above shall be verified to have had their-trip setpoints changed to the values specified in Specification 2.2.1 for the applicable number of reactor coolant pumps operating either:
.a. - Within 4 hours af ter switching to a different pump combination if switch-is made while operating, or
- b. Prior: to reactor criticality if switch is made while shutdo.in.
#See Special-Test Exception 3.10,5'.
i CALVERT; CLIFFS-UNIT 2. 3/4.4-2 . Amendment No. 6,16 l
- 10. 0 -. -STARTUP TESTING The following : discussions represent the major startup tests _ proposed for Calvert Cliffs I, Cycle 5. Sufficient data is obtained to verify -
- that' the plant operates in a safe condition within the bounds of the applicable acceptance criteria and, therefore, the safety analysis.
L 10.1 HOT FUNCTIONAL TESTING 10.1.1 CEDM Performance Testing During this testing, the proper functioning of the CEAs, CEDMs, and CEA position indication will be verified through the insertion and withdrawal of- the CEAs. Rod drop . times will be measured and evaluated. Any irregularities shall be analyzed. 10.1.2 HCS Flow Verification RCS flow rates will be verified based on differential pressure measurements obtained across the RCPs and RV. These values will be compared tio those obtained during previous testing for consistency. 10.2 INITIAL CRITICALITY Approach to criticality. will commence with the withdrawal of the Shutdown CEA Groups, followed by the withdrawal, in sequence,- of the ~ Regulating CEA Groups concluding with Group 5 at mid-core.
. Criticality _will be established through baron dilution. The plant will be allowed to stabilize following Critical Baron Concentration, and then' proceed to the Low Power Physics Tests to verify physics design parameters.
10.3 LOW POWER PHYSICS TESTING _1 l 10.3.1. .CEA Symmetry Check l CEAs will be , partially inserted _into the core, and withdrawn from the I
~
.x core f to ' confirm proper latching to their respective CEA extension shafts. A qualitative reactivity change will be apparent for single
~
CEAs and a quantitative reactivity for dual CEAs will be determined
-for the purpose of confirming core symmetry.
- 10.3.2 Crit'ical Baron Concentration Critical Baron Concentrations 'will be determined for ARD, and Groups 1 through 5 inserted.
10.3.3 Isothermal Temoerature Coefficient-By varying the RCS temperature, the ' Isothermal Temperature
. Coefficient will be -determined. CEA Regulating Group 5 will be used to control and maintain flux and reactivity within a defined operating band.
. 10.3.4 -CEA Group Worth Measurements The RCS will'be diluted / borated while the CEAs are inserted / withdrawn to compensate for a change in. reactivity. These changes will be monitored via the reactivity computer.
10.4 POWER ASCENSION TEST. Two major plateaus..for. testing will be at 50% power and at 100% power. The following specific tests will be performed as shown in order to compare and verify as-built characteristics of the core with their respective predictions. 50% and 100% Plateau Testing.
._ Upon reaching each ' power level, Xenon Equilibrium wiH be established with- Bank 5 withdrawn to- 105 inches. The critical baron concentrations ' and . core power distributions will be determined and verified with predictions. . k E _ Eu. ,
f% " ' Follbwing ~ xenon equilibrium and the ' power distribution measurement,
.TC will be. varied, thereby yielding data for the . isothermal temperature coefficient determination. The Power Coeffcient will be determined by. maintaining T avg constant and varying the l power
-level. In both cases, CEA 5-1 will' be used for reactivity control and maintaining power.
10.5 ACCEPTANCE CRITERIA Acceptance criteria for the above startup testing will be developed consistent with those presented during previous startup. Acceptance Limits: CEA Groups _ Worth + 15% on each group
+ 10% on sum of all groups measured
-Critical Boron Measurements i 100 PPM Temparature Coefficient + .3 X 10-4 ap/oF Power Coefficient + .2 X 10 -#Ap/%P Rod Drops 3.1 seconds Power Distribution Ff,F , and T within q
Technical. Specifications Limits. The measured radial box power distri-butions will be compared with predic-tions. If a measurement varies from a prediction by more than +10% (+15% for fuel assemblies on core periphery then the difference will be resolved and the validity of the safety analy-sis confirmed prior to submittal of ; the summary report of startup . test results. l CEA Symmetry Check < 10% tilt; a tilt of >10% will be resolved prior to exceeding 20% of maximum- allowable thermal power ~ level for the existing RCP I
-combination. - I
If any;. acceptance criteria limits. listed above are exceeded, 'an evaluation 'shall be made to determine first, the applicability of the predicton: to the . precise plant; conditions under which the test was performed; second, the accuracy of the measurement; - finally,.. the validity 'of the physics' data input to the safety analysis for- the entire cycle. Specifically, if any regulating bank worth measurement
- falls outside of. Its acceptance criteria or if the total worth of the regulating banks.. falls outside of its ' acceptance criteria, shutdown
. bank--C. shall -be measured and compared with its acceptance criteria.
If-shutdown bank 'C worth falls outside of its acceptance criteria or if-~ the accumulated total worth of all the banks measured falls: below their total worth acceptance criterion (after appropriate corrections and adjustments) then an evaluation shall be made of the validity of
.the safety analyses for the entire cycle.
A summary ~ report of the results of these tests will be submitted to the NRC within 45 days of completion of the startup program. I l J wr : .. _ _
-a
gs7_ t SECTION
11.0 REFERENCES
l-
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REFERENCES (Chapters.1 thr'ough 6)- .
- 1. - ? Letter,J A. ! E. Llundvall',Jr. (BG&E) to R. W. .;Reid (NRC) dated February 23,
'1979).
~
, 2. (Letter, : A. ' E. Lundvall', Jr. (BG&E):t0 R.1 W. Reid (NRC), " Report af Startup
< Testing . for l Cycle : Four",- Calvert :- Clif fs ' Nuclear Power Plant Unit - No. 1, DocketLNo.-50-317, dated October 15, 1979
- 3. : D.- E. - Bessette, _. et Jal, " Examination . of ' Calvert Cliffs-1 Test Fuel 2
Assemblies at' End of' Cycles 1 and 2",..NPSO-72, September 1978
-4. E..'.J. ~Ruzauskas, et' al, " Examination of Calvert Cliffs-1 Test Fuel Assembly' After ' Cycle 3", NPSO-87, September:1979
- 5. " Baltimore' Gas and Electric Company Calvert Cliffs Nuclear . Power ' Plant Units -l'and.2 Final ~~ Safety Analysis Report", dated January 4,1971
- 6. . ' 8G&E.LCalvert Cliffs I~ Slides Depicting ' SCOUT-1 High Burnup Demonstration
~ Program
- 7. Letter, , A~ - E., ' Lundvall, Jr. to
. B. -C. Rusche, "Second Cycle License
-Application", dated October 1, 1976
; 8. Letter, . J. W.- Gore,1 Jr. -- to E. . G. Case, " Third. Cycle License Application",
' dated December 1,1977, as = modified by letter, A. E. Lundvall, Jr. _ to R.
L W. - Reid, " Request: for Amendment. to 0perating License", dated.May 8,1978
'9.L cCENPD-187, "CENPAN Method' offAnalyzing Creep Collapse of Oval -Cladding",
' dated June'1975
.e
- 10.iCEN-83(B) P, J"Calvert(Cliffs Unit il : Reactor Operation with ' Modified CEA GuidefTubes",fdatedFebruary 8,51978 and letter, A. E. Lundvall,_Jr. to V.
!Stello,JJr., IReactor(Operation with ;Modifled'.CEA Guide Tubes", dated'
~
~
n-. IFebruary ;17, l-1978 .. o
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.lb. CEN-1Cl(B)-P, Letter, A. - E. Lundvall to R. W. Reid, "Calvert Cliffs II Cycle 2 Reload Submittal Update", dated August 28, 1978
- 12. . ' Letter, A. E. Scherer to R. A. Clark, LD-80-019, " Slides from NRC Meeting of April 30, 1980", dated April 30, 1980 13 . - CENPD-139,~"C-E Fue'l Evaluation Model Topical Report", dated July 1974 A 14.- CENPD-153-P, Revision 1, " Evaluation of Uncertainty in the Nuclear Power
~
Peaking Measured t/ the Self-Powered Fixed In-Core Detector System", dated May 1980
- 15. CENPD-161-P, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core", July 1975
- 16. CENPD-162-P-A (Proprietary) and CENPD-162-A (Nonproprietary), " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids Part 1, Uniform Axial Power Distribution", April 1975
- 17. CENPD-206-P, " TORC Code, Verification and Simplified Modeling Methods",
January 1977
- 18. C. Chiu, J. F. Church, "Three-Dimensional Lumped Subchannel Model and Prediction - Correction Numerical Method for Thermal Margin; Analysis of PWR Cores",. TIS-6191,: June 1979
- 19. C. Chiu, P. Moreno, N. E. Todreas, R. W. Bowring, "Enthalpy Transfer Between PWR . Fuel Assemblies in Analysis of the Lumped Subchannel Model",
Nuclear Engineering Design, 53, July 1979, pgs. 165-186
- 20. ~CEN-124(B)-P, " Statistical Combination of Uncertainties, Part 2", January
.1980
- 21. . Supplement 3-P (Propr.ietary) to'CENPD 225P, " Fuel and Poison Rod ~ Bowing",
June 1979 ,
'22. Letter from : D. 8. Vassallo (NRC) to A. E. Scherer (C-E), dated June 12, c1978-m
& t ; ,
' ;*- ;g;: ~ m y
a 3
^
.23.c CEN-124(B)-P, f" Statistical Combination of- Uncertainties, t Part :1", January.:
.1980- ~ ,
. . -s . . ; ..
~24.1 CEN-124(8)-P. s" Statistical'-Combination of Uncertainties, Part-l3", March 3
-1980
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I REFERENCES FOR'SECTION 7 (NON-LOCA TRANSIENT ANALYSES)'- la. '" Statistical Combination of Uncertainties Methodology; Part 1: C-E Calculated Local Power Density and Thennal Margin / Low Pressure LSSS for Calvert Cliffs Units I and II",:CEN-124(B)-P, December,1979. lb. . Statistical Combination of Uncertainties Methodology: Part 2: Combination of System Parameter Uncertainties in Thermal fiargin - Analyses for.Calvert Cliffs Units I and II", CEN-124(B)-P, January,1980, lc. " Statistical Combination of Uncertainties Methodology; -Part 3: C-E Calculated Local Power Density and Departure from Nucleate Boiling Limiting Conditions for Operation for-Calvert Cliffs Units I and II", CEN .124(B)-P, March,1980.
- 2. CENPDd99-P,"C-ESetpointMethodology", April,1976. I
- 3. Letter'from A. E. Lundall to R. W. Retd, dated February 23, 1919.
k 4. Letter from J. N. Gore Jr. to E. G. Case, " Third Cycle License. Application", December 1,1977,- As modified by Letter from A. E. Lundall to R. W. Reid, " Request for Amendmert to Operating License", liay 8,1978, i
- 5. Letter from A. E..Lunda11 Jr. to B. C. Rusche, "Second Cycle License Application", October 1, 1976. -
- 6. CENPD-190A, "CEA Ejection, C-E fiethod for Control Element Assembly Ejection", July,1976.
- 7. GEMP-482, H. C. Brassfield, et al. , f' Recommended Property and Reactor Kinetics Data for Use in Evaluating a Light Water-Cooled Reactor Loss- .
l of-Coolant Incident Involving Zircaloy-4 or 304-SS, Clad U0 "' AP#il' 1968. 2 -
- 8. Idaho Nuclear Corporation, Monthly Report, Ny-123-69, October,1969.
- 9. Idaho Nuclear Corporation, Monthly Report, Hai-127-70, March, 1970.
m - - r"r 4
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' References 1 (Section .8) -
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. ; ;4 'l ;s
- 1. LAcceptance Criteria ifor EmergencpC'o re Co.olin9 Systens' -
for Light- Watar
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- ? dc ;s
. . .. c -c .. >
//W '
~
, - Cooled Nuclear Power Reactors 1<- MFeddral Register, -Vol .' ~39,- gflo. -r
. '3 . ' ' < ,
. Q.
Friday;; January:4,11974. ^ k : , n i
- 2. CENPD-132, " Calculative Methods 1for;the = CE Large Break LOCA Evaluation I
M'odel"., August 19_74; (Pro'prietary).
, ~f -
3 V'-- 7, N
'y .
.-CENPD-132i Break LOCA Evaluation.Model", Supplement -December 1, '.'1974 Updated (Proprietary). Calculative Methods f 4' ,
/ k CENPD-132, . Supplement 2, . "Ca'lculational . Methods for the CE L'arge Bre'akL
LOCA, Evaluation Model", ~ July 1975 (Proprietary) . I{
- 3. .CEMPD-135, "STRIKIN-II,= A' Cylindrical Geometry Fuel Rod Heat Transfer Program", April .1974 (Proprietary). '
CENPD-135.. Supplement- 2,"STRIKIN-II, A Cylindrical Geometry Fuel Rod
- Heat' Transfer Progrom (ibdificition)", Febru:ry 1975 (Pecprietary).
'CENPD-135, Supplement'4, "STRIKIN-II, A Cylindrical Geometby Fuel Red Heat Transfer.?cogram", August 1975 (Prcprket .cy) . -
t CEtiPD-135, Supplement 5, "STRIKIN-II, A Cylindrical Gech,etry Fuel T,cd Heat Transfer- Program", April 1977. (Proprietary) .
- 4. CENPD-139,'"CE Fuel Evaluation Model", July 1974 -(Proprietary). o;
- 5. CENPD-137, " Calculative Methods for the CE Small Break LOCA Evaluation
" o , - ' <- nt.
.j Model", Combustion Engineering Propr,ietary Report, AugIfs '. 1974. %
.. ~ ,- . <
(Proprie tary) . CENPD-137, Supplement 1, " Calculative Methods for the-CE Sm'alk BreSkl Evaluation M:1t,-Janbary 1977-(Preprictary). ki: '6' . To be supplindi/ 0G?!EL(Calvert Cliffs II Cycle II ECCS Analvsis). (7. !To be supplied by BG&E (Calvert Cliffs -I Cycle IV ECCS Analysis'). I 9
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9.r ;-CENPD-138, . "PARCHL- A FORTRAN-IV -Digital Program. to3 Evaluate' Pool . N' f Boiling,fAxial: Rod- aiid Coolant lleatup",' August-1974f(Proprietary). j w [[C'ENPDil38, Supplement (2 2 P January 1977.-(Proprletary).- - l k Y
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1.0 INTRODUCTION
This document describes the Reactor Protection System (RPS) Asymmetric Steam Generator Transient Protection Trip Function-(ASGTPTF) and its design bases. The ASGTPTF is designed to. protect against Anticipated Operational Occurrences (A00s) associated with secondary system malfunctions which result in
-asymmetric primary loop temperatures. The most limiting event is the loss of load to one Steam Generator-(LL/1SG)-caused by a single Main Steam Isolation Valve (MSIV) closure.
The Baltimore Gas and Electric (BG&E) RPS presently employs an analog thermal l 7 margin trip calculator as part of the-Thermal Margin / Low Pressure (TM/LP) trip function. To provide a reactor trip for asymmetric design basis events, pressure in each of the two steam generators will be monitored and-these signals input to the thermal margin calculator. Secondary pressure imbalan:es between the two generators will be calculated and a corresponding factor applied in the TM/LP Calculator to generate a trip signal. Protection against exceeding the DNBR and maximum Kw/Ft Specified Acceptable Fuel Design Limits (SAFDL's) during the LL/ISG event is presently provided by the-Low Steam Generator Level reactor trip in conjunction with sufficient initial margin maintained by the Limiting Conditions for Operation (LCO's). The ASGTPTF will result in a reactor trip sooner than the Low Steam Generator Level trip and, hence, will produce a smaller margin degradation during this
. event. The additional margin gain allows full advantage to be taken of r
margin recovery progra'ms designed to achieve stretch power, or 18 month fuel cycles for future BG&E reload cycles by assuring that the asymmetric 4 y transients would not be limiting A00's for establishing the LCO's. 4 J 9
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2.0 SYSTEM DESCRIPTION 2.1 General "he ASGTPTF-consists of:
- 1. Existing steam generator pressure sensors (one for each steam generator per channel) and associated process equipment.
- 2. Existing Thermal Margin / Low Pressure (TM/LP) calculators modified to include a bistable with an input of the absolute value of the pressure difference between the two steam generators lPSG1 - PSG2l. The output cf the bistable signals the TM/LP calculation when trip conditions are reached.
- 3. Existing RPS . trip logic and Reactor Trip Switchgear.
Modifications to the existing TM/LP calculator will be discussed next as the rest of the system consists of previous?y licensed installed equipment. A functional block-diagram of this portial of the system is provided in Figure 1.
-2.2 TM/LP Modifications A steam generator pressure signal is input to each of the TM/LP calculators from each steam generator. In each TM/LP calculator, the difference between the two pressure signals is calculated. If the difference exceeds a set amount, a bias is input to the TM/LP calculation. This will result in a reactor trip. The additional bias input to the TM/LP calculation is the asymmetric factor signal (Fas). This will raise the setpoint to a high enough level to ensure a trip if the steam generator pressures differ by more than the setpoint value. Figtme 2 illustrates the functional relationship between the absolute value of the pressure difference, lPSG1 - PSG2l,andthe asymmetric factor signal, Fas.
- i. .
m 3.0 DESIGN BASES . 3.1 Design Basis Events The ASGTPTF is designed to provide a reactor trip for those design basis events associated with secondary system malfunctions which result in' asymmetric primary loop coolant temperatures. The most limiting event is the LL/ISG caused by a single Main Steam Isolation Valve (MSIV) closure. 3.2 Design Criteria The ASGTPTF is designed to the following criteria to ensure adequate performance of its trip function:
- a. The trip function is designed in compliance with the applicable criteria of the General Design Criteria for Nuclear Power Plants, Appendix A of 10 CFR 50, July 15, 1971.
- b. Instrumentation, function and operation of the trip logic conform to the requirements of IEEE Standard 279-1968, Criteria for Protective Systems for Nuclear Power Plants.
- c. The trip function is designed consistent with the recommendations of Regulatory Guide 1.53, Application of the Single-Failure Criterion to Nuclear Power Plant Protective Systems, and Regulatory Guide 1.22, Periodic Testing of Protection System Actuation Functions.
- d. Four independent measurement channels are provided.
- e. The protective system ac power is supplied from four separate vital instrument buses,
- f. The ASGTPTF can be tested with the reactor in operation or shut down.
- g. Trip signal is preceded by n pretrip alarm to alert the operator of undesirable operating conditions in cases where operator action can correct the abnormal condition and avoid a reactor trip.
~
- h. The ASGTPTF components which will be used are of the same type presently in use at BG&E, and will meet the same industry standard as applied to the original RPS (i.e., IEEE-279, August 1968). The operation of the ASGTPTF is not required.during or subsequent to any Design Basis Event which significantly alters the. containment environment (LOCA, Main Steam Line Break or Feedwater Line Break). Therefore, it is not required that additianal in-containment equipment installed specifically for the ASGTPTF be qualified for the adverse environments associated with these events.
- 1. The trip . function is designed so that protective action will not be initiated due to normal operation of the generating station.
- j. All equipment will be designed in accordance with the QADM. Vendor quality control will be in accordance with C-E Procedure WQC 11.1, Revision D. '
- k. Modification to the TM/LP Calculator for the ASGTPTF will not jeopardize previous qualification of this equipment.
)
3.3 Performance Requirements The selection of a trip setpoint is such that adequate protection is provided when all sensor and processing time delays and inaccuracies are taken into account. Final determination of an equipment setpoint is based on equipment characteristics, operating environment, NSSS performance and safety analysis. .The nominal setpoint, uncertainties and response time are provided in Table 1. e e 4 e e o i I LE_ _
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. ASYMMETRIC STEAM GENERATOR TRANSIENT PROTECTION TRIP FUNCTION NOMINAL CHARACTERISTICS
. 1 .
Nominal System. Accuracy 35 psi An-lysis Setpoint +175 psid Nominal Equipment Setpoint +140 psid . Nominal Pretri'p Setpoint . +100 psid Nominal System Response Time <.9 seconds Nominal: Expected value only. Final values are to be determined
. later, and included in the plant Technical Specifications as appropriate.
I _ _ _ _ _ _ _ _ _ _ - - - _ _ - _ _ _ _ -_ - -- k o _ - -
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2500'- - Fas + - .3 1 E. . Setpoint psid + lP3gy - PSG2l 4 4 FIGURE 2 ras ve. lPSG1 - PSG2l ASYMMETRIC FACTOR SIGNAL __ _ _ _ _ _ _ _ _ _ _ ,m d s A [ w J $ - APPENDIX B - Method for Calculating Space-Time Scram ReactivitLes for-Use in Calvert Cliffs Unit 1 -Cycle 5 Safety Analysis l i i l .c s b f '^ a ~ i' ? L 4; , . l ,, -; A'. ', k. ;. .. - ^ '~ q d . L.U i ' , c ,y 1.0~ INTRODUCTION ' It has been.C-E's standard practice to calculate the time-dependent' scram . ~ reactivity I'nsertion by._ calculating the critical:eigenvalue as a function of rod position. This' type of calculation -assumes that the neutron flux ' ' shape during a scram rod insertion is equal to the critical flux shape at ~ each! rod position. -Such an acsumption is conservative.- The critical (or . static) _ flux shape tends to' shift away' from the rods more than the space-time-calculated - flux shape does, and- hence results in 'a smaller reactivity' insertion. ! _ The primary reasons for the difference between the flux shapes is that the space-time calculation accounts for the effects of delayed neutrons, while the . static calculation ignores them. The delayed neutron > precursors are distributed according to the initial neutron flux shape. ! As the neutron population decreases, the importance of the delayed neutrons' ' increases. The delayed neutron precursors provide a source of , neutrons which tends to tilt the space-time-calculated neutron flux shape more toward the rods than the static methods predict. Accounting; for the delayed neutron precursors leads to greater reactivity insertion with the space time method at intermediate rod positions. This effect becomes larger at the end of a reactor cycle when the power shape tends to be axially flat. At certain intermediate rod positions, for example, the reactivity Lchange' predicted by static methods may be only a halff or -a ' third of that predicted by space-time methods.' At full _ rod ; insertions, both methods yield nearly the same reactivity. This E append.tx) describes FIESTA, a one-dimensional two-group space-time . kinetics compdter code ' for PWR scram reactivity calculations which ~ . ; account ;for -the axial - ~ space-time variations in the neutron flux -(Reference 'l).- .The FIESTA code is used -for scram reactivity calculations -l'nstead of' the- HERMITE code c(Reference' 2), previously approved by NRC,
- mainly.1for - two . reasons: ~'(1).1 FIESTA calculations , are more rapid than HERMITE1 calculations ,- and (2) FIESTA produces kinetics parameters which ,
~ are used for point k.inetics calculations. . ~ / f [',__L-_a' # _ _' =:- . -? a, _ , .~ _ , .~ , -r . ( 1 L' yn .nx . ., W LThe FIESTA code 'has been Lverifled ~ witn HERMITE.. ~ItJ has. also been N fjdstifledIby' comparisonLto : standard benchmark cxperiments. . ' Although - the '~ { code has been (written: toLincorporab feedback on fuel- temperature and u , moderator dens'ity, e such . feedbacks are not . needed for : scram reactivity .cq1culations. ~A~ comparison of FIESTA and HERMITE evaluations of a core's responsef tof a' scram nis : provided. This- shows that the. FIESTA-generated ~ , . scram data is ~a Lvalid equivalent of the scram data caJeulated by HERMITE. . Scram data . intended .to be conservative for_ Cycle 5 and all ' subsequent ~ cycles of the .Calvert Cliffs ~ units have bedn ' developed with FIESTA. The generic] power distrib0tions used ' cover all times in life for a variety of reload ~ cores. This. data has'been used in the' safety analysis of the loss - of flow ~ transient ' and - any other accident for which a reactor scram is ~ -initiated. 'The: Use of 'this gc.,eric data .in the Cycle 5 safety analysis is conservative. 1.1
SUMMARY
~
The FIESTA program solves the time dependent neutron diffusion equation using. the1 space time s factorization method. This method .is identical to
~
the improved L quasistatic method presented :in Reference 3. The procedure entails L the . simultaneous solution _ of the flux amplitude and. reactor kinetics - equations. The simultaneous solution requires that _ reactivity insertion ivs. l time ~ calculated' by FIESTA ' will necessarily b'e consistent
- with Lpower . level : vs. time : calculated - by FIESTA. . To demonstrate that FIESTA Tcalculates~ the- space time' scram correctly, its. solution must be
~
s compared - to that : of f anz approved code. IERMITE is used to :nake - this i comparison.'- ;The _ currently ' approved version of HERMITE 'does not explicitly.l calculate _ reactivity; _ it does calculate the effect' of
- reactivity l insertion on_the transient.being analyzed.-
; ; Af comparison Eoff a J FIESTA-generated . power--levelvs. . time trace and a
~
HERMITE-generated power level :vs. time: tracei for a - typical ~ reactor: scram kisigivent in Figure l1h Thej flux shape solutions of ~the two computer codes
' n (during$different stages l.off the scram transientiare_ compared in Figures;2 7 '
, ,? 3through J7. "These ifigures ? ~ illustrate : the . good ' agreement between~ -the-FG
~
evaluations off the:same-reactor scram by the(two codes. .
- i g ' f*i y ,
h , v -
' : a f,a The power level vs. time trace generated by; the FIESTA calculation is essentially; the same as that generated by HERMITE. . The flux shapes are also essentially identical. Therefore, the reactivity insertion vs. time trace calculated by FIESTA is consistent with the HERMITE solution.
A comparison of cycle specific reactivity insertion vs. time with generic data was . performed for a typical initial power distribution- to illustrate the degree of conservatism inherent in the use of generic data in the Cycle 5 safety analysis. This comparison showed that the use of cycle specific data would have inserted at least 7% more negative reactivity during the first second of the scram. Figure 8 displays the results of-the use. of generic vs. cycle specific reactivity insertion data for the Loss of Flow transient. It shows that the impact of the use of generic scram reactivity data reduces the minimum DNBR reached in this transient by 0.03 DNBR units. Since the use of the generic data is conservative, this data is used in the accidents analyzed in this license amendment where a scram is initiated. 1.2 C0tCLUSIONS The following conclusions can be drawn with respect to the use of FIESTA in generating scram reactivity data. (1) - A FIESTA power trace and a HERMITE power trace for the same scram transient are nearly identical.
-(2) FIESTA calculates reactivity simultaneously with power. Since it calculates-the power correctly, it must calcuate the reactivity correctly as well.
~
~
l
-(3) -The use' of generic scram reactivity insertion data instead of cycle '
specific data is conservative for the transients analyzed in this license amendment. l
~ REFERENCES t.
1._ CEN-133(8), " FIESTA-~ -1 A One-Dimensior.a1 Two-Group Space-' Time . Kinetics Code for Calculating PWR Scram Reactivities", November 1979
- 2. CENPD-188-A, '"HERMITE - A 3-D Space . Time Neutronics Code With Thermal Feedback", July 1976
- 3. :K. O. Ott, D. A.r Mensley, ' " Accuracy of the Quasistatic Treatment of Spatial Reactor Kinetics", NSE-36 pp402-411'(1969)
V
r FIGURE 1
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y CYCLE 5Scm1 DATA 1.26 2.9 U 1,7 GEf1ERICSCRAftDATA 1.23 3.0 _ d 5 g 1.6 U y 1.5 s D - - 1 , 14 2 z g 1,3 - CEA's Begin to Drop Into Core 1.1 1 I I i 1,0 0 2 4 6 8 10 TIME, SECONDS GAS ELE T IC CO. ' calvert ciith MINIMUM HOT CHANNEL CE-1 DNBR v s TIME 8 Nuclear Power Plant
O APPENDIX C DESCRIPTION OF MODEL USED TO SIMULATE NSSS BEHAVIOR DURING STEAf1 LINE RUPTIJRE EVENT l i l i g
The Steam Line Rupture (SLB) event analysis for Cycle 5 included the effects of manually tripping the Reactor Coolant Pumps (RCP's) on Low Pressurizer Pressure and the automatic initiation of auxiliary feedwater on low steam generator water level. To accurately predict the NSSS behavior.during this event with manual trip of RCP's and automatic initiation of auxiliary feedwater, modelin NSSS simulation code CESEC (Reference C-1)g . changes The major were model made to the improvements necessary to perform this analysis are:
- 1) A more detailed _ thermal hydraulic model of the reactor coolant -
system in order to simulate the effect of temperature tilt in the reactor core during a SLB event with coastdown of RCP's.
- 2) An explicit representation of the reactor vessel closure head to more accurately predict the RCS pressure during this event.
- 3) A flow model which is able to explicitly calculate the time dependent reactor coolant mass flow rate.
- 4) An improved primary-to-secondary heat transfer model together with a steam generator dry out model to more accurately predict the RCS cooldown after auxiliary feedwater flow delivery to a potentially dry steam generator.
- 5) An improved model of the boren transport in the primary coolant ~ and of the Safety Injection System to calculate the reactivity effect -
of baron -injected via the High Pressure Safety Injection Pumps (HPSI's), Low Pressure Safety Injection Pumps (LPSI's) and the Safety Injection
. Tanks (SIT's).
These model improvements are discussed below and were incorporated in
- a. version of CESEC called CESEC-SLB. The CESEC-SLB code was used to simulate the NSSS behavior during a SLB event with manual trip of RCP's and automatic initiation of auxiliary feedwater for Calvert Cliffs Unit 1, Cycle 5.
RCS Thermal-Hydraulic Hodel In'CESEC-SLB, the nodal model used is given in Figures C-1A and C-1B. Each channel represents one half of the reactor vessel inlet downcomer section, the lower plenum, the core region, the bypass flow, and the upper plenum. These two symmetrical loops are linked by cross flow at the reactor - vessel inlet and outlet sections and by flow' mixing within the reactor vessel _ lower and upper plenums. ~ The mixing factors specified are based
- on test data. No cross flow is assumed between parallel regions-in '
the core. The selection'~of the ' temperature used for the moderator reactivity calculation irdetermined by a linear.. extrapolation based on the two parallel core
' node' temperatures obtained from the solution of the energy, mass balance and constant volume thermal-hydraul x ecuations.
o; r - , a t
,s ,
Closure Head Node ,
. During the. rapid contraction o'f. the- primary coolant.which takes place
.as- a. result 'of assteam line break, the pressurizer empties and voids begin to form in'the RCS. Since- flow through the closure head is only
.a.small fraction of the-RCS flow, the temperatures in the closure head remain high and voiding first occurs there. To some extent, the.
closure ~ head itself then begins to perform the function of a pressurizer. Therefore, the reactor vessel closure head region is explicitly modeled in CESEC-SLB to more accurately predict.the RCS pressure. The coolant flow from theipper plenum nodes to the vessel head node is specified by user input fractions. It is assumed that the vessel head fluid returning into the outlet nodes'is evenly distributed between :n* two loops. . Flow Model The CESEC-SLB flow'model calculates the mass flow rate (lbm/sec) at the i pump outlet for each reactor coolant system steam generator loop. The model includes explicit. simulations of the reactor coolant pumps and of the effects of natural circulation flow. The :alculations are based on a 1olution of the one-dimensional momentum equation for each RCS loop. The loops are divided into a number of. nodes whose densities, temperatures, and flows are obtained from the CESEC thermohydraulics model. The flow model utilizes this nodalization.of the loop to calculate the sum of the various' forces around the loop. The forces acting on the fluid volume ; consist'of (1) : gravitational forces due to. density and elevation changes around the: loop, (2) viscous forces due to wall friction and geometric expansions and contractions of the flow path, (3) forces , due to .the RCS pumps. s-The' force,or head due to the pumps is a function of pump speed and flow. It is derived from input tables of homologous pump data. Pump [ ! speed >is calculated transiently using 'the pump inertia and net torque
.on the impeller. . The net torque is the sum of hydraulic, eletrical, friction and windage torques. The hydraulic' torque, like pump head,
.is a function 'of. pump speed'and flow and is derived from input tables of pump speed and flow and fron input tables 'of homologous pump .
~ data. Electrical torque is obtained from an input table of torque versus speed. -Friction'is' assumed vary linearly and windage is assumed to vary quadratically with pump' speed.
-primary-To-Secondary Heat Transfer In CESEC-SLS, the ' average primary-to-secondary temperature difference is
! calculated using algorithms . based on a polynomial fit of the nodal mid
' point temperatures to obtain'.the spatial: variation of' the primary side
-temperature.: In addition, the heat transfer coefficients are calculated on a: node-center-to-node-center basis ~rather than for the steam generator as a whole, s
The; polynomial algorithm in CESEC-SLB. selects'the mid-point temperatures lof theLnodes simulating the'steamigenerator tubes and the inlet and outlet plenum to,obtain1the spatial' variation of the primary side temperature along the steam generators tubes. lThe difference between
- thi,sitemperature variation and the secondary side temperature fis. integrated
.: i _ M-
between he node mid-points and divided by the distance between these points to obtain the average primary-to-secondary temperature difference. This temperature difference is then used to calculate the steam generator heat transfer rate on a node-center-to-node-center basis at each time step. The heat transfer correlations used in CESEC-SLB are the same as used in Reference C-1, ex. cept, the heat transfer coefficients are calculated on a ' node-center-to-node-center. basis'rather than for the whole steam generator. For low heat flux predictions, the secondary-side heat transfer coefficient is limited to a constant input value, rather than being allowed to go to zero with the heat flux. This minimum value is also used for the secondary-side heat transfer coefficient for conditions of reverse (secondary-to-primary) heat transfer. Steam Generator Dry Out Model In CESEC-SLB, the product of the overall heat transfer coefficient with heat transfer area, UA, depends on steam generator liquid inventories. For steam. generator inventories greater than 5000 lbm, the total heat transfer area is used together with a primary-to-secondary heat transfer coefficient calculated using the same correlations reported in Reference C-1. For liquid inventories less than 2500 lbm the product of heat transfer coefficient and heat transfer area is assumed to be just sufficient to raise the enthalpy of any incoming feedwater to that of the saturated liquid. For liquid inventories between 2500 lbm and 5000 lbm the product UA is scaled linearly between values calculated for inventnries greater than 5000 lbm and inventories less than 2500 lbm. In all cases steam is allowed to escape through the ruptured steam line with a critical flow ve16 city as long as the steam pressure is above atmospheric pressure and the steam generator liquid inventory is greater than 2500 lbm. " Safety Infection System The borated safety injection water from the high and low pressure pumps is injected into each cold. leg downstream of the reactor coolant pumps. The N borated injection flow rates versus pressure are specified by input tables. Once the safety injection flow reaches the cold leg, it is assumed to mix homogeneously.with the reactor coolant inventory. The boron is transported through the RCS by solving at each time step the continuity equation for.each coolant node for the baron concentration. The boron concentration for the reactor. core node is used to calculate the reactivity contribution due to boron via an input inverse baron
~
worth. A time delay is input to account for the time required to start the diesel generator and/or to bring the safety injection pumps to full speed. An additional time delay is input to account for the-time required for the unborated water.in the safety injection line (from the outlet of the safety 3' injection pumps to the injection nozzles) to be swept out before borated water from the refueling water tanks enters the cold legs.
7 , 1 CESEC-SLB solves an orifice equation to determine the rate of safety , injection flow from the safety injection tank into the RCS as a function of time. In computing the safety injection flow rate by means of an orifice equation, the code takes into account .the effect of piping friction,' turning losses, and expansion / contraction losses through the
'use of a single equivalent loss coefficient 'tSch is based on the
' minimum cross-sectional flow area. The input parameters are the initial nitrogen pressure, volume of water, volume of gas, flow coefficient, flow area, water specific volume, and elevation head of the safety. injection tanks. In addition to the nitrogen pressure within the safety injection tank, the static head of fluid within the safety injection piping is accounted when calculating the instantaneous pressure difference across the _ orifice. The nitrogen expansion process is assumed to be isentropic.
Reference C CENPD-107, "CESEC Digital Simulation of a C-E Nuclear Steam Supply System', April, 1974. ' r
.4.
r zoM C SPRAYS = SAFETY VALVES e_ n = u iME
,a GG ai y Og ( .
uG@o 27 22 S O rn o O a ,o 23 = 22 = 25 26 10 11 w o i Ri o a o o a E M 24- 21 = 20 = 19 : 7 = 8 = 9 12 a
> o a o
BB
- o a = 18 6 :
20 m 1 r i r
, p .
3 5 16 17 5 4 m l" Q . o o a o a a O 5 2 15 3 [ LETDOWN = = SPRAYS o . o n y PUMP -
\ PUMP
[L,)
! = 13 = 14 = 2 : 1 :
L_, RC hl . of ' a n l
,1 c VUHARGING - SAFETY INJECTION PUMPS AND TANKS a a
-CHARGING V
N00E PHYSICAL DESCRIPTICN 1 COLD LIG 2 UPSTREM HALF 0F. INLET PLENUM (SEFORE FLOW MIXING) 3 00WNSTREAM HALF 0F INLET PLENUM (AFTER FLOW MIXING)
'4 BYPASS FLOW S CORE 6 UPSTREAM HALF OF OUTLET PLENUM 7 00WNSTREAM HALF 0F OUTLET PLENUM 8 HOT LEG '
9 STEAM GENERATOR INLET PLENUM 10- UPSTREAM HALF 0F STEAM GENERATOR TUBES 11 DOWNSTREAM HALF OF STEAM GENERATOR TUBES
-12 STEAM GENERATOR CUTLET PLENUM ~
13 SA'ME AS T IN OTHER STEAM GENERATOR LOOP' , 14 SAME AS 2 IN OTHER STEAM GENERATOR LOOP q 15 SAME AS 3 IN OTHER STEAM GENERATOR LCOP 16 SAME AS 4 IN OTHER STEAM GENERATOR LOOP 17 SAME AS 5 IN OTHER STEAM GENERATOR LOOP 18 SAME AS 6 IN OTHER STEAM GENERATOR LOOP 19 'SAME AS 7 IN OTHER STEAM GENERATOR LOOP 20 SAME AS 8 IN OTHER STEAM GENERATOR LOOP 1 21 SAME AS 9 IN OTHER STEAM GENERATOR LOOP 22 SAME AS 10 IN OTHER STEAM GENERATOR LOOP , 23 SAME AS 11 IN OTHER STEAM GENERATOR LOOP l 24 .SAME.AS 12_IN OTEER STEAM GENERATOR LOOP l 25L l REACTOR VESSEL CLOSURE HEAD 25 5 URGE LINE l 27 PRESSURIZER l 1
. l i
~ ' '
3 GAS E T! C O. PHYSICAL DESCRIPTION OF PRIMARY .ConLANT NODES s ! n ual ccivert citris .FOR CESEC - SL3 C-1B ! Nuclecr. Pcwer Plant a * [:}}