ML17285A299

Proposed Tech Specs Re Use of Advanced Nuclear Fuels Corp Reload Fuel in Cycle 5

ML17285A299
Person / Time
Site:	Columbia
Issue date:	03/03/1989
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17285A298	List: ML17285A297 ML17285A299 ML17285A300 ML17285A301 ML17285A302
References
TAC-72251, NUDOCS 8903090441
Download: ML17285A299 (27)

Text

UPDATED ATTACHMENT TO WNP-2 CYCLE 5 RELOAD

SUMMARY

REPORT TECHNICAL SPECIFICATION CHANGES cy 0303 0

8903%ADOCH @000397 pDR o p Fig/gg g9Ão'fd i/8P

s

~,

0 i

CONTROLLED COPW INDEX LIST OF FIGURES PAGE SODIUM PENTABORATE SOLUTION SATURATION TEHPERATURE... 3/4 1-21 3.1. 5-2 SODIUH PENTABORATE TANK, VOLUME VERSUS CONCENTRATION, Rf(UIREMENTS......................................... 3/4 1-22 3.2. 1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE BCR183........................ 3/4 2-2

3. 2. 1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (HAPLHGR) VERSUS AVfRAGE. PLANAR EXPOSURE, INITIAL CORf FUEL TYPE 8CR233........................ 3/4 2-3 3.2. 1-3 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVfRAGE BUNDLE EXPOSURE ANF Bx8 RELOAD FUEL.......................'........... 3/4 2-4 3.2. 1-4 MAXIMUM AVERAGE PLANAR I.INEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPE 8CR183...........:.................... 3/4 2"4A 3.2. 1-5 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURf, INITIAL CORE FUEL TYPE 8CR233................................ 3/4 2-4B REDUCED FLOW HCPR OPERATING LIHIT.................... 3/4 2"8

3. 2. 4" 1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE ANF 8x8 RELOAD FUEL.......... 3/4 2-10 3.2. 6-1 OPERATING REGION LIMITS OF SPEC. 3.2.6............... 3/4 2-12 3.2. 7-1 OPERATING REGION LIMITS OF SPEC. 3.2.7............... 3/4 2-14

3. 4. 1. 1-1 THERMAL POWER LIMITS OF SPEC. 3.4.1.1-1.............. 3/4 4-3a 3.4. 6. 1-1 HINIMUH REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (INITIALVALUES)...... 3/4 4-20 3.4.6. 1" 2 HINIMUM REACTOR VESSEL HETAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE (OPERATIONAL VALUES)......... 3/4 4"21

4. 7-1 SAMPLE PLAN 2) FOR SNUBBER 'FUNCTIONAL TEST .......... 3/4 7-15

3. 9. 7-1 HEIGHT ABOVE SFP MATER LEVEL VS. MAXIMUM LOAD TO BE CARRIED OVER SFP..................................... 3/4 9"10 B 3/4 3-1 REACTOR VESSEL MATER LEVEL.......................:... B 3/4 3-8 3iQ, I "6 hlA)can& ooI AVERAGE Pl.ANAIE, c.lNLAEE, HEAT GEuQIEATIou RATE CMAPt.ttGA') VOGIE,rug AVIAAGo IBuuoc.E Expofv+C AuD 9x I 9Ã put g A~F'x9-xg 2- '/C MASHINGTON NUCLEAR - UNIT 2 XX Amendment No. 63 W, 2,'I- 2 I-INBAP HEAT 68uaAATzou RAT& (svG R) Lt8iT VBVu5 '3/p Z "IP A bVSIEAaG PI.AmAIE. r X POOoCS ANF WX9-X y F'uEE.

%24- 9 l.Iuf Ate, /NEAT 6 QAI'GREAT(oaf PI suan. Exposes

~AT5 CI IIGR) CIIIIiV VERSo5 Ix9-9x puEI Pled A vznacr AMF

I CONTROLLED COPM

'AFETY LIMITS BASES THERMAL POWER Low Pressure or Low Flow (Continued) at this flow is approximately 3..35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50K of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure psig is conservative.

below'85 2.1.2 THERMAL POWER Hi h Pressure and Hi h Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage're not directly observable during reactor opera-tion, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the cor e are expected to avoid boiling'ransition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determin~$ using the AHF Critical Power Methodology for boiling water reactors which is a statistical model that combines. all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boil-ing transition is determined using the AHF nuclear critical heat fluxenthalpy XH-3 correlation. The XH-3 correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.

The required input to the statistical model are the uncertainties listed in Bases Table B2.1.2-1, The bases for the uncertainties in the core parameters are given in XN-NF-524(A), Rev. 1 and the basis for the uncertainty in the XH-3 correla-tion is given fn XH-NF-512(A), Rev. 1 (b) . The power distribution is based on a typical 764 asseably core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle.

would not be as severe as the distrfbution used in the. analysis.

a. Exxon Nuclear Critical Power Methodology for Soiling Water Reactors, b.

XH-HF-524(A), Rev. 1.

Exxon NucTear Company XH-3 Rev. 1.

Crftfcal Power Correlation, XH-HF-512(A), i

~

WASHINGTON NUCLEAR - UHM 2 B 2-2 Amendment No.

59

~ o O.

1 REAGTI YITY CONTROL SYSTEMS 0

FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LIMITIHG CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time of all operable control rods from the fully withdrawn position, for the four control rods arranged in a two"by-two array, based on deenergization of 'the scram pilot valve solenoids as time.

zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser Full Withdrawn tion Time Seconds 45 39 25 5

APPLICABILITY: OPERATIOHAL COHDITIOHS 1 and 2.

ACTION:

With the average scram insertion times of control rods exceeding the above limits: 'I

1. Declare the control rods with the 'slower than average scram insertion times inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and

~ 0',

2. Perform the Surveillance Requirements of Specification 4.1.3.2.c 'L ~

at least once per 60 days when operation is continued with an average scram insertion time(s) in excess of the average scram insertion time limit.

Otherwise, be in at least HOT SHUTDOWN within the next U. hours.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMEHTS 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.

WASHINGTON HUCLEAR UNIT 2 3/4 1-8 Amendment Ho. 45

3/4. 2 POWER OISTRIBUTION LIQQ 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE

~ LIMITING CONDITION FOR OPERATION

~~el g,2. I-3.2. 1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE for GE fuel and average bundle exposure for ANF fuel all not exceed the limits shown in Figures 3.2.1-1,

3. 2. 1-2, ee4- 3. 2. 1-3> when in two loop operation, and Figures 3. 2. 1-4, 3. 2. 1-5, and 3.2. 1-6 when in single loop operation. g, g,l-3 APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than

~A REE f RATER TRERPIAE PftlER.

ACTION: 3g I 3g With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, m 3.2.1-3 o>>~'-~

in two loop operation or Figure 3..1-4, 3.2.1-5, or 3.2.1-6 in single loop operation, initiate corrective action within 15 minutes and restore 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 POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

~

'URVEILLANCE'RE UIREMENTS 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.

a0 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%%uo 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 ROD.PATTERN for APLHGR.

WASHINGTON NUCLEAR - UNIT 2 3/4 2-1 Amendment No. 45

~.

11.0 10.5 10.0 C 5 Cl ~g Q. Pg 9.0 Bundle 4 a Average Exposure MAPLHGR 8.5 (Mwo/MT) kw/ft I >7>

8.0 5,000 11.2 C 10,000 11.2 7.5 15>000 11.2 20>000 11.2 7.0 25>000 9.7 30,000 8.1 6.5 35,000 6.8 6.0 5>000 10>000 15,000 20,000 25,000'0,000 35,000 Bundle Average Exposure (MWDIMT)

ANF 9 X 9 - lX AND 9 X 9 - 9X Reload Fuel Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure Figure 3.2.1-6

~ o 4

0NTROLLED COPY Table 3.2.3-1 MCPR OPERATING LIMITS MCPR Operating Limit U 'to 106K Core Flow Cycl e Equipment

~Ex osure Status GE Fuel ANF Fuel

l. 0 MWD " 3750 MWO MTU MTU

2. 3750 MWD - EOC MWO Normal scram times"" l. 31 MTU MTU

3. 3750 MWD - EOC MWD Control rod insertion 1. 38 bounded by Tech. Spec.

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

4. 3750 MWD " EOC MWD RPT inoperable MTU MTU Normal scram times

5. 3750 MWD - EOC MWD RPT inoperable -:L-.55" MTU MTU Control rod insertion t.46 l.%2, bounded by Tech. Spec.

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

~

~;-

0 MWO " EOC MWD Single loop operation 1. 40 1. 37 MTU MTU RPT operable Normal scram times"~

"In this portion of the fuel cycle, operation with the given MCPR operating limits is allowed for both normal and Tech. Spec. scram times 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 (3.1.3.4-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 arra seconds Notch 45 .404 Notch 39 .660 Notch 25 1. 504 Notch 5 2. 624 WASHINGTON NUCLEAR - UNIT 2 3/4 2-7 Amendment No. 62

Two Loop Operation 20 30 40 50 60 70 80 90 100 110 Total Core Recirculating Flow (% Rated)

Reduced Flow MCPR Operating Limit Figure 3.2.3-1 890040

4o POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) for GE fuel shall not exceed 13.4 kW/ft. The LHGR for ANF fuel shall not exceed the values shown in Figure 3.2.4-1~ A a < 2~ ct c/ 9.2 APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or

~21 22 2 RATER TIIERIRI PRIIER.

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 next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 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 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 on a LIMITING CONTROL R00 PATTERN for LHGR.

WASHINGTON NUCLEAR " UNIT 2 3/4 2-9 Amendment No. 45

~ o 0

14 Bundle Average 13 Exposure LHGR (MWDIMt) kw/ft 0 13.7 5,000 13.7

~ ~ 10,000 13.7

~

E 11 15,000 13.7 20,000 13.0 6 30>000 11.5 40,000 10.0 05 10 50,000 8.5 60,000 7.0 C

0 9 70,000 5.5 I

t5 6 8 G

I tD X',

Cl g) 6 5

0 5,000 10,000 20,000 30,000 40,000 50,000 60,000 ?0,000 80,000 Average Planar Exposure (MWD/MT)

ANF 9 X 9- IX Reload Fuel Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure Figure 3.2.4-2 890040.3

~.

0 14 Bundle Average 13 Exposure LHGR (MWD/MT) kw/ft 12 0 13.1 5>000 13.1 10,000 13.'f 15,500 13.1 11 20>000 12.5 30,000 11.2 40,000 9.9 10' 8.6 50,000 60,000 7.3 70>000 6.1 8

7 6

0 5,000 10,000 20,000 30.000 40,000 50.000 60.000 70,000 80>000 Average Planar Exposure (MWD/MT)

ANF 9 X9-9X Reload Fuel Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure Figure 3.2.4-3 890040.4

0 e'

3/4. 2 POWER

& CONTROLLED COF 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 RATE (APLHGR) for GE fuel is 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 is 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. Figures3.2.1-3 appl ~

~

y'o both single and two loop operation. awl32)

JPal 4 3.2.1-1, 3.2.1-2, 3-2.1-3, 3.2. 1"4, ~

The calculational procedure used to establishjthe APLHGR shown on Figures 3.2. 1-5~>s based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which 2

Rev. 1.

2 2 eaOo.xoSC'c P 2...,

are consiste'nt with the requirements of Appendix K to 10 CFR Part 50. These 1 2,,22 WASHINGTON NUCLEAR " UNIT 2 8 3/4 2-1 Amendment No. 62

a 0-4

OESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES gofL<

gggg 5VO<

5.3.1 The reactor core shall contain 764 fuel assembIies with each fuel assembly containing 62 fuel rods and two water rods clad with Zircaloy-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 weight

- percent U-235. Reload fuel shall be similar in physical design to the initial core loading> s'xcawr v~+v @zan Rcco~o wvGc HAg Qpfpcv Q ARRAY or- F LA@ E. RoOS CONTROL ROO ASSEMBLIES 5 .3.2 The reactor core shall contain 185 control rod assemblies, each consisting of a crucifo array of stainless steeI tubes containing 143 inches of boron carbide, 84C, g@gr surrounded by a cruciform shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM Qy OESION PRESSURE ANO TEMPERATURBg 5.4. 1 The reactor coolant system esigned and shall be maintained:

a. In accordance with the cod equirements specified in Section 5.2 of the FSAR; with allowance for ormal degradation pursuant to the appl'icable surveillance requir ts,

b. For a pressure of:

l. 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 steam volume of the reactor vessel and recirculation system is approximately 22,539 cubic feet at a nominal steam dome saturation temperature of 545'F.

MASHIHGTON NUCLEAR - UNIT 2 5-5

~ o

~,

ATTACHMENT TO WNP-2 CYCLE 5 RELOAD

SUMMARY

REPORT'UMMARY JUSTIFICATION FOR TECHNICAL SPECIFICATION CHANGES

Tech. S ec. No'. Justification B2.1.2 Editorial change only to reflect change submitted and approved as Technical Specification Amendment No. 28.

3/4.1.3.4 This change corrects a previous oversight in the Tech.

Specs. (see cover letter).

3/4.2.1 Addition of NAPLHGR curve for 9x9 fuel.

3/4.2.3 New MCPR values to reflect cycle specific transient analysis.

3/4.2.4 Addition of LHGR cur ves for 9x9-IX and 9x9-9X fuel.

83/4.2.1 Editorial change to bases to reflect change to Tech.

Spec. 3/4.2.1'iscussed above.

5.3 Editorial change to Design Features to reflect the use of 9x9 lead test assembly fuel.

e'