ML17285B054

Proposed Tech Specs Re Use of Cycle 6 Reload Fuel in Plant

ML17285B054
Person / Time
Site:	Columbia
Issue date:	02/27/1990
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
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NUDOCS 9003080222
Download: ML17285B054 (29)

Text

ATTACHMENT TO MNP"2 CYCLE 6 RELOAD

SUMMARY

REPORT

SUMMARY

JUSTIFICATION FOR TECHNICAL SPECIFICATION CHANGES

0

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~T... ~PN . Justification INDEX XX Editorial change to reflect addition of LFA NAPLHGR and LHGR curves 1.0 1-2 Editorial change to critical power ratio definition to more accurately describe the critical power ratio 2.0 B 2-1 Editorial change to the BASES introduction to account for use of non-ANF LFA fuel in WNP"2 3/4.2.1 3/4 2"1 Addition of NAPLHGR curves Figures 3.2.1-7 3/4 2-4D and 3.2.1-8 for LFA fuel 3/4 2-4E 3/4.2.3 3/4 2-7 New NCPR values to reflect cycle specific 3/4 2-8 transient analysis and revision to Figure 3.2.3-1 to change title.

3/4.2.4 3/4 2"9 Addition of LHGR curves Figures 3.2.4-4 and 3/4 2-10C and 3.2.4-5 for LFA fuel 3/4 2-10D B3/4.2. 1 B 3/4 2"1 Editorial change to bases to reflect change to Tech. Spec. 3/4.2.1 discussed above B3/4.2.3 B 3/4 2-3 Editorial change to bases to reflect change to Tech. Spec. 3/4.2.3 discussed above 5.3 5-5 Editorial change to Design Features to reflect the use of lead fuel assemblies.

ATTACHMENT TO HNP-2 CYCLE 6 RELOAD

SUMMARY

REPORT TECHNICAL SPECIFICATION CHANGES

CNTROLLED COPY INOEX LIST OF FIGURES FIGVRE PAGE

3. 1. 5" 1 SOOIUM PENTABORATE SOLUTION SATURATION TEMPERATURE... 3/4 1-21

3. 1. 5-2 SOOIUM PENTABORATE TANK, VOLUME VERSUS CONCENTRATION REQUIRPilENTS . 3/4 1-22

3. 2. 1-1 MAXIMUM AVERAGe PLANAR LINEAR HEAT GFNERATION RATE (MAPLHGR) VERSUS AVERAGc PLANAR EXPOSUR, INITIAL CORE FUEL TYPE 8CR183. 3/4 2-2

3. 2. 1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE'FUEL TYPE BCR233. 3/4 2-3 J

3. 2. 1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNDLE EXPOSURE ANF 8x8 RELOAO FUEL................. . . .... ....... 3/4 2-4

3. 2. 1-4 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE BCR183.. 3/4 2-4A

3. 2 1-5

~ MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAl.

CORE FUEL TYPE 8CR233 3/4 2-4B

3. 2 1-6

~ MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE BUNOLE EXPOSURE ANF 9x9-IX ANO 9x9-9X FUEL 3/4 2-4C 6A~

3. 2. 3-1 ..............

~ REOUCEO FLOW MCPR OPERATING LIMIT 3/4 2-8

3. 2. 4" 1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 8x8 RELOAO FUEL.......,.. 3/4 2-10

3. 2. 4-2 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 9x9-IX FUEL.............. 3/4 2-10A

3. 2. 4" 3 LINEAR HEAT GcNERATION RATE (LHGR) LIMIT VERSUS av g~

3. 2. 6-1 AVERAGE PLANAR EXPOSURE ANF 9x9-9X OPERATING REGION LIMITS OF SPEC.

FUEL..............

3.2.6 .........

3/4 2-10B 3/4 2"12

3. 2. 7-1 OPERATING REGION LIMITS OF SPEC. 3.2 7 3/4 2" 14

3. 2. 8-1 OPERATING REGION LIMITS OF SPEC. 3.2.8.... 3/4 2-16

3. 4. l. 1" 1 THERMAL POWER LIMITS OF SPEC. 3.4.1.1-1.............. 3/4 4-3a

3. 4. 6. 1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (INITIAL VALUES)...... 3/4 4-20 WASHINGTON NUCLEAR - UNIT 2 XX Amendment No. 71

INSERT A MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS BUNDLE AVERAGE EXPOSURE SVEA-96 LEAD FUEL ASSEMBLIES MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS BUNDLE AVERAGE EXPOSURE GEll LEAD FUEL ASSEMBLIES INSERT B LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE SVEA-96 LEAD FUEL ASSEMBLIES LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE GE11 LEAD FUEL ASSEMBLIES

I i N II

GQ. s nuu.ru COPY DEFINITIONS CHANNEL FUNCTIONAL TEST

l. 7 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.

CORE ALTERATION 1.8 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Sus-pension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CRITICAL POWER RATIO g ppg y pg/p7 E 6 gI-Tlc A+

1.9 The CRITICAL POWER RATIO (CPR) shall be that power in the assembly which is calculated by application of the correlation to cause some point in the assembly to experience boiling transition divided by the actual assembly operating power.

DOSE E UIVALENT I-131 1.10 DOSE E(UIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TI0-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with'alf-lives greater than 15 minutes, making up at least 95K of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM .(ECCS) RESPONSE TIME

1. 12 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor unti the ECCS equipment is capable of performing 1

its safety. function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

WASHINGTON NUCLEAR - UNIT 2 1-2 Amendment No.28

C(+TROLLED COPY ej 2.0 SAFETY LIMITS and LIHITIHG SAFETY SYSTEH SETTI%5 BASES INTRODUCTION The fuel cladding, reactor pressure vessel and prfaary system pfpfng are the principal barriers to the release of radioactive aaterfals ta the environs.

Safety Lfafts are established ta protect the integrity of t>se barriers during norsal plant operations and anticipated transients. The fuel cladding integrity Safety Lfaft fs set such that no fuel daaage fs calculated ta occur ff the lfait fs not vfolated. Because fuel daaage fs not directly observable, i step-back approach fs used ta establish a Safety Lfaft such that the KPR is not less than 1. 06 for tvo recirculation loop operation and 1. 07 for single recfrcula-tfon loop operation for pJ QcLBAR fue KPR greater than 1.06 for bra recirculation loop operation and 1.07 or single recirculation loop operatfon represents a conservative aargfn rela-tive to the conditions quired to aafntafn fuel cladding integrity. The fuel iu valp-2 cladding fs one of,the physical barriers whfch separate the radioactive aate-rfals floe the environs. The fntegrfty of this cladding barrier fs related to fts relative freedoe fma perforations or cracking. Although scee corrosion or use related cracking say occur during the life of the cladding, fission product afgration free this source fs fncreaentally cmulatfve and cantfnuausly

~ easurabl e. Fuel cladding perforations, however, can result free theraal stresses which occur free reactor operation significantly above desfgn condi-tions and the Lfnftfng Safety Systel Settfngs. Vhile fission product migration free cladding perforation is just as acasurable as that froa use related crack-fng, the thersally caused cladding perforatfans signal a threshold .beyond which still greater theraal stresses lay causa gross rather than fnc~ntal cladding deterioration. Therefore, the fuel cladding integrity Safety Lfaft fs defined with a Iargfn ta the conditions which would produce onset of transition boiling, KPR of 1.0. These candftfans represent a significant departure fran the con-dition intended by design for planned operation. The }KPR fuel cladding integ-rity safety lfaft assures that during noraal operation and during antfcfpated operational occurrences, at least 99.9 percent of the fuel rods fn the care do not experience transition boiling ( ).

~mSSRI C

2. I SAFETY LIMITS 2.1.1. THILL POIIIER Low ~sure or Low Flow

~

For cartafn ccedftfons of pressure axi flow, the XN-3 carrelatfan fs not valid for all crftfcal power calculations. XN-3 carrelatfon fs not valid for bundle sass velocftfes less thai .25 x IO lbs/hr ft or pressures lass than 585 psfg. Therefore, the fuel claddfng integrity Safety Lfift fs estab-lfshed by other Neans. This fs done by establishing a lfaftfng ccedftfon on care THENhL POKR Nth the fallowfng basis. Since the pressure drop fn the bypass region fs essentially all elevatfon head, the care pressure drop at low power and flows wfll always be greatar than 4.R psf. Analyses show that with a bundle flow of 28 x Io lbs/h (approxfsataly a sass velocfty af .25 x 10 lbs/hr-ft~), bundle pressure drop fs nearly independent of bundle power vfll be greater than 2S x I alnd has a value of 3.5 psf. Thus, the bundle flow wfth a 4.5 psf driving head lbs/h. Full scale ATLAS tast data taken at pres-sures free 14.7 psfa ta 8N psfa indicate that the fuel assumably crftfcal power NSHICTN NXLM - SfIT 2 hwndaent No. 45

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INSERT C (Reference; XN-NF-524(A); Rev. 1; ABB Atom Report UK90-126; GE11 Lead Fuel Assembly Report for Washington Public Power Supply System Nuclear Project No.

2, Reload 5, Cycle 6). The latter. two references support application of the above established safety limit to GE11 and SVEA-96 LFA fuel in WNP-2.

~I C(+TROLLED COPY 3/4. 2 POWER DISTRIBUTION LIMITS

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3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE l.IMITING CONDITION FOR OPERATION gyp'-gg h)JO GB lg L FA )ge Ze J 7 aRe&C'l 3e'le ) 8 <<>Nf gfhg CfpRC

)

3.2.1 All AVERA E PLANAR LINEAR HEAT GENERATION RATES I(APLHGRe) for each type of fuel as a fun tion of AVERAG PLANAR EXPOSURE for,GE~fuel and average bundle exposure for AN fuel shall not exceed the limits shown in Figures 3.2. 1-1,

~

3. 2. 1-2, 3.2. 1-3, eacf 3. 2. 1-6 when in two loop operation, and Figures 3. 2. 1-3,

3. 2. 1-4, 3.2. 1-5, 3. 2. 1-6when in single loop operation.

. y g, I 7 eR~<l Pe2eI f APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

ACTION: 7,P. z,)" ~ 70e 3 ~ e)g )-f 3 )

Mith an APLHGR exceeding the limits of Figure 3.2. 1-1, 3.2. 1-2, 3.2. 1-3,)N:

3. 2. 1-6~ in two loop operation or Figure 3.2. 1-3, 3. 2. 1-4, 3. 2. 1-5, single loop operation, initiate corrective action within 15 minutes and restore

~ 3. 2. 1-6~ in APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POMER to less than 25K of RATED THERMAL POMER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits de-termined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1-6>3~.1-7 ~~ J

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Mithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POMER increase of at least 15K of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROO PATTERN for APLHGR.

WASHINGTON NUCLEAR - UNIT 2 3/4 2-1 Amendment No. 69.

~ ~ ~ ~

Bundle

~ ~ Average

~ ~ ~ ~ Exposure MAPLHGR (MWD/MTU) (I<W/ft) a a 0 8.90 C

5,000 8.90 10,000 8.90 15,000 8.90 20,000 8.90 25,000 7.74 30,000 6.44 35,000 5.41 C)

I CU 10000 20000 30000 40000 Bundle Average Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure SVEA-96 Lead Fuel Assemblies Figure 3.2.1-7 900052.2 0

12 11 Two Loop and Single Loop Operation 10 Bundle 9 Average Exposure (MWD/MTU) MAPLHGR LLI 0 10.9 I CU 8 5,000 10.9 10,000 10.9 15,000 10.9 20,000 10.9 25,000 9.5 7 30,000 7.9 35,000 6.6 6

5000 10000 15000 20000 25000 30000 35000 Bundle Average Exposure Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure GE 11 Lead Fuel Assemblies Figure 3.2.1-8 900067.5

GQITROLLED COPY Table 3.2.3-1 HCPR OPERATING LIMITS MCPR Operating Limit Uo to 106M Core Flow Cycl e gag 5'VBA '76 Equipment n g LFA Exoosur e Status GE uel ANF Fuel

1. 0 MWD - 3750 MWD .24 1. 24 l,37 RTU HHU

2. 3750 HWD " EOC MWD Normal scram times"" l. 35 1. 31 I,Vg HTU TMU s.36 3, 3750 MWD " EOC HWD Control rod insertion 1.42 I 95 HTU RTO bounded by Tech. Spec.

limits (3.1.3.4-p 3/4 1-8)

4. 3750 HWD " EOC HWD RPT inoperable I55',

FiiJ FR Normal scram times""

5. 3750 MWD - EOC HWD RPT inoperable l. 48

~MU MHU Control rod insertion bounded by Tech. Spec.

limits (3.1.3.4-p 3/4 1-8)

O. 0 MWD -'OC MWD Single loop operation .35 l. 35 t.s I HHU RTO RPT operable Normal scram times""

"In this portion of the fuel cycle, operation with the given HCPR operating limits is allowed for both normal and Tech. Spec. scram <imes and for both RPT operable and inoperable.

""These MCPR values are based on the ANF Reload Safety Analysis performed using the control rod insertion times shown below (defined as normal scram). In the event that surveillance 4. 1.3.2 shows these scram insertion times have been exceeded, the plant thermal limits associated with normal scram times default to the values associated with Tech. Spec. scram times (3. l. 3. 4-p 3/4 1-8),

and the scram insertion times must meet the requirements of Tech. Spec.

3. 1.. 3.4.

Slowest measured average control rod insertion times to specified notches for all operable control rods for each Position Inserted From group of 4 control rods arranged in a Full Withdrawn a two-b -two arrav (seconds)

Notch 45 . 404 Notch 39 . 660 Notch 25 l. 504 Notch 5 2. 624 WASHINGTON NUCLEAR - UNIT 2 3/4 2-7 Amendment No. 69

INSERT D The GE11 LFA fuel, the ANF LFA fuel and the GE initial core fuel are also monitored to the ANF Sx8 fuel MCPR Operating Limits (Reference; Power Distribution Limits, Bases, 3/4.2.3, Minimum Critical Power Ratio,

p. B 3/4 2-3).

it Two Loop Operation MCPR Total Core Operating Flow Rate Limit 100 1.07 90 'l.13 80 1.19 70 1.26 60 1.34 50 1.45 40 1.59 CO I

CU 20 30 40 50 60 70 80 90 100 110 Total Core Flow (% Rated)

Reduced Flow MCPR Operating Limit This Curve ls Applicable to ANF Reload Fuel, GE Initial Core Fuel, ANF 9 X 9 LFA Fuel, GE11 LFA Fuel, and SVEA-96 LFA Fuel Note: This curve ls also applicable to FFTR operation when approved.

Figure 3.2.3-1 900067.2

CITROLI ED COI POWER DISTRIBUTION LIMITS 3/4:2.4 LINEAR HEAT GENERATION RATE c ~

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gp (1'Th<

0 LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERAT ON RATE (LHGR) for GE fuel shall not exceed 13.4 kW/ft. The LHGR for ~

Figures 3.2.4-1, 3.2.4-2, 3.2.4-3z fuel shall not exceed the values shown in s.z. V-Vn~~l APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25~ or RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the nex't 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REOUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15~ of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a LIMITING CONTROL ROO PATTERN for LHGR.

WASHINGTON NUCLEAR - UNIT 2 3/4 2"9 Amendment No. 69

t 12 Exposure LHGR (MWD/MTU) (kw/ft)

~ ~

E 0 to 40,000 11.6 I

6$

11 K

C

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0

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Ql 10 C)

CU I Ql CU X:

C$

C 0 10000 20000 30000 40000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure SVEA-96 Lead Fuel Assemblies Figure 3.2.4R 900052.1

14 EXP LHGR 13 0 13.1 510 13.1 2,580 12.7 12 5,230 12.3 7,940 11.9 10,470 11.8 11 13,220 11.8 15,990 11.8 18,708 11.7 10 21,590 11.7 24i420 11.7 27)280 11.0 9 30,150 10.3 C)

CO 33,050 9.6 I

35,960 8.9 CV 6 38,900 8.0 41 1830 7.3 44,760 6.5

?

6 0 5000 10000 15000 20000 25000 30000 35000 40000 45000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure GE 11 Lead Fuel Assemblies Figure 3.2.4-5 900067.4

CITROI LED COPY 3/4. 2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50. 46.

3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. For GE fuel, the peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR.

times 1.02 is used in the heatup code along with the exposure dependent steady-state gap conductance and rod-to-rod local peaking factor. The Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATF {API HGR) for GE fuel 1s this LHGR of the highest powered rod divided by its local peaking factor which results in a calculated LOCA PCT much less than 2200'F. The Technical Speci-fication APLHGR for ANF fuel is specified to assure the PCT following a postu-lated LOCA will not exceed the 2200'F limit. The limiting value for APLHGR 15 shown in Figures 3. 2. 1-1 and 3. 2. 1-2 for two recirculation loop operation and Figures 3.2. 1-4 and 3.2. 1-5 for single. loop operation. Figures 3.2. 1-3, aa4-3.2. 1-6 apply to both single and two loop operation.

calculational procedure used to establish the APLHGR shown on Figures The

3. 2. 1-1, 3, 2. '.

1-2, 3 1-3, 3. 2. 1-4, 3. 2. 1-5, and 3. 2. 1-6 is based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. These models are described in NEDO-20566P or NN-NF-80-19, Volumes 2, 2A, 2B and 2C, Rev. 1. 7A~ ~ cC4 o4< ~o R as7+eci>Him' 6'C 4FA APL/A 6R Vnc vaS S'RocVM JM Fi g v~cS 7.2./-7 4'l 1-'3 avw g tv~~ i~ VH'9'0-/24 a~ei C4c gpgg l <eel Fv&c.

p >ZCr Cc.y ECeoa7 po 4'Ha C Ta~ /'veuc PocuCa Svt ec y TE-H 5 y'S TE Pfv~g ~p y. ItRyg gg g /VAN, Q ASl.dA 0 5 Cy<< z WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-1 Amendment No. 7l

ONTROLLED COPY POWER DISTRIBUTION LIMITS BASES 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial condi-tion of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specifica-tion 2.2.

'o assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature. decrease. The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Table 3.2.3-1.

The evaluation of a given transient begins with the system initial param-eters shown in the cycle specific transient analysis report that are input to an ANF core dynamic behavior transient computer program. The outputs of this program along with the initiaI MCPR form the input for further analyses of the thermally limiting bundle. The codes and methodology to evaluate pressurization and nonpressorization events are described in XN-NF-79-71(P) and XN-NF-84-109(A).

The principal result of this evaluation is the reduction in MCPR. caused by the transient.

The purpose of the HCPRf of Figure 3.2.3-1 is to define operating limits at other than rated core flow conditions. At less than 100K of rated flow the required MCPR is the maximum of the rated flow HCPR determined from Table 3.2.3-1 and the reduced flow MCPR determined from Figure 3.2.3-1, MCPRf assures that the Safety Limit MCPR wi 11 not be violated. MCPRf is only cal-culated for the manual flow control mode. Automatic flow control operation is not permitted.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2"3 Amendment No. 45

INSERT E Lead Fuel Assemblies (LFA's) from Advanced Nuclear Fuels (ANF), General Electr ic (GE) and ABB Atom (ABB) reside in the reactor core. Analyses performed by the three vendors'ndicate that the transient CPR changes for the LFA's are greater than the CPR change calculated for the dominant ANF Bx8 fuel,,

due primarily to the shorter thermal time constants of the smaller diameter rods. All vendors state that their LFA's have inherently higher thermal margins (by design) than the dominant 8x8 fuel. Each vendor chose to address the CPR limit in a slightly different fashion. These methods are discussed as follows.

GE concludes that the inherent high thermal margin of the LFA's is sufficient to compensate for the larger CPR change associated with the shorter time constant and that the ANF Bx8 limits can be conservatively applied to the GE11 LFA's.

The XN-3 CHF correlation used by ANF in the analysis was developed for the ANF Bx8 fuel. A review of the correlation and comparison to CHF data obtained for the 9x9 LFA's concluded XN-3 is conservative when applied to the ANF LFA's and that the LFA's should be conservatively assumed to have a CPR performance at least equal to that calculated for an Bx8 assembly for the same power and inlet conditions. In addition, due to the water canister in the interior of the bundle,. ANF modified the S-factors for the 9x9 LFA's to improve the XN-3 predictive capability. 'hese S-factors were used in the analyses and were provided for use in monitoring the LFA's.

ABB Atom chose to take a more conservative approach and performed analyses which established conservative and unique NCPR values for the SVEA-96. The resulting NCPRs are included in the Technical Specifications.

In addition to the conservatisms discussed, the Supply System has commited to load the LFA's in core locations which have been analyzed to have sufficient margins such that the LFA's are not expected to be the limiting assemblies in the core on either a nodal or a bundle power basis. This approach is to prevent the possibility of the LFA's from ever being the limiting fuel bundle and adds additional margin to the CPR in the event of a plant transient.

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CO/TROLLED-.C OESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3. 1 The reactor core shall contain 764 fuel assemblies with each initial core fuel assembly containing 62 fuel rods and two water rods clad with Zi rcaloy-2. Each fuel rod shall have a nominal active fuel length of 150 inches.

The initial core loading shall have a maximum average enrichment of 1.90 ueight percent U-235. Reload fuel shall be similar in physical design to the initial

'ore loading except that the reload fuel may employ a 9 x 9 array of fuel rods.

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CONTROL ROO ASSEMBLIES co~+sgvva croup n.l~ ~cc,ocucJ.

5.3.2 The reactor core shall contain 185 control rod assemblies, each cons-isting of a cruciform array of stainless steel tubes containing 143 inches of boron carbide, S4C, powder surrounded by a cruciform shaped stainless steel sheath.

5. 4 REACTOR COOLANT SYSTEM DESIGiV PRESSURE AND TEMPERATURE 5.4. 1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable surveillance requirements,

b. For a pressure of:

1. 1250 psig on the suction side of the recirculation pump.

2. 1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3. 1550 psig from the discharge shutoff valve to the jet pumps.

c. For a temperature of 575 F.

VOLUME

5. 4.2 The total water and s.earn volume of the reactor vessel and recirculation system is approximately 22,539 cubic feet at a nominal steam dome saturation temperature of 545'F.

WASHINGTON NUCLEAR - UNIT 2 5-5 Amendment No. 69

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