ML17279A162

Proposed Tech Specs Re Cycle 3 Reload Summary Rept

ML17279A162
Person / Time
Site:	Columbia
Issue date:	03/27/1987
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
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ML17279A161	List: ML17279A160 ML17279A162 ML17279A164 ML17279A167 ML17279A168
References
NUDOCS 8704030057
Download: ML17279A162 (29)

Text

ATTACHMENT TO WNP-2 CYCLE 3 RELOAO

SUMMARY

REPORT TECHNICAL SPECIFICATION CHANGES 8704030057

'DR 870327 ADOCK 05000397" PDR

I 2-0 0

GGNTRGLLED

..0 SAFETY LIMITS and LIMITING SAFETY SYSTEM SETTINGS

. GGPY BASES g Jgpg~gg~ AJ ucl BAIL Ft/6Ls coAP0RATIDlJCA+$

INTROOUCT ION The fue'l cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated ransients. The fuel cladding integrity Safety Limit is set such that no fuel d m age is calculated to occur if the limit is not violated. Because fuel damage i not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.06 for two recirculation loop ration and 1.07 for single recircula-

.tion loop operation for both GE and fuel. MCPR greater than 1.06 for two recirculation loop operation and 1.07 for single recirculation loop operation represents a conservative margin relative to the conditions required to main-tain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations sig-nal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding integrity Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity safety limit assures that during normal operation and during anticipated operational occurrences, at least 99.9 percent of the fuel rods in the core do not experience transition boi ling (Reference XN-NF-524 (A), Rev. 1).

2. 1 SAFETY LIMITS 2.1. 1. THERMAL POWER Low Pressure or Low Flow For certain conditions of pressure and flow, the XN-3 correlation is not valid for all critical power calculations. The NN-3 correlation is not valid for bundle mass velocities less than .25 x 10'bs/hr-ft~ or pressures less than 585 psig. Therefore, the fuel cladding integrity Safety Limit is estab-lished by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10'bs/h (approximately a mass velocity of .25 x 10'bs/hr-ft~), bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10'bs/h. Full scale ATLAS test data taken at pres-sures from 14.7 psia to 800 psia indicate that the fuel assembly critical power WASHINGTON NUCLEAR - UNIT 2 B 2"1 Amendment No. 28

I GGNTRGLLEB GGPY SAFETY 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 below 585 psig is conservative.

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 are 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

'imit. 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'f considering the the fuel rods in the power core are expected to avoid boiling transition distribution within the core and all uncet tainties.

A<F A mt=

T R R Methodology. for boiling water reactors .. which is a statistical model that combines all of the uncertainties in op rating parameters and the procedures used to calculate critical power. The p obability of the occurrence of boil-ing transition is determined using the nuclear critical heat flux-enthalpy XN-3 correlation. The XN-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 and the nominal values of the core, parameters listed in Bases Table B2.1.2-2.

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 XN-3Rcorrela-tion is given in XN-FN-512(A), Rev. 1 . The power distribution is based on a typical 764 assembly 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 distribution used in the analysis.

a0 XN-NF-524(A), Rev. l.

b. Exxon Nuclear Company XN-3 Critical Power Correlation, XN-NF-512(a),

Rev. l.

WASHINGTON NUCLEAR - UNIT 2 B 2"2 Amendment No. 28

GGNTRGILLEB GGPY REACTIVITY CONTROL SYSTEMS w7~C FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION y+ g, PPIC 4 d < ceo'tp ~ L >o J

~

3.1.3.4 The average scram insertion time from the fully withdrawn position, for the four control rods arranged in a two-by-two array, based on deenergiza-tion 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 0. 455 39 0. 920 25 2. 052 5 3. 706 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a. With the average scram insertion times of control rods exceeding the above limits:

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

2. Perform the Surveillance Requirements of Specification 4. 1.3.2.c 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 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS

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. l.3.2.

WASHINGTON NUCLEAR - UNIT 2 3/4 1"8 Amendment No. 28

0GNTRGLLED GGPV 3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2. 1 All A ERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a unction of AVERAGE PLANAR EXPOSURE for GE fuel and average bundle exposure for fuel g ll, ~g@ g$ g i iQ gggn in Figures 3.2.1-1, 3.2.1-2, and 3.2.]-3p Figures 3.2.]-4, 3.2.1-5, and 3.2.]-6~ wRF M IN S'Ih>G L& Goo p oPcRh rI4Q APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater. than or equal to 25K of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits'f Figure 3.2. 1-1, 3.2.1-2, or 3 2 1-3 iM ~Iso aooP initiate corrective action within 15 minutes and restore APLHGR to within 'oPaRhv VIOAVRC 3.Z,l>>7>

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 g > i~ 4g yg /-g 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 />. IH SIJv&L& LooP O teRh'ri os SURVEILl.ANCE RE UIREMENTS 4.2. 1 All APLHGRs shall be verified to be equal to or less than the limits determined from Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-3:

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 ]5K 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. 28

I

'

7 4)0 Z c}op /pe Flog 13.0 12.5-I-12 0 L

C .

o3 O 11.5 o t5 IK LC 11.0 Bundle o Average L

C Exposure MAPLHGR 4 I 10.5- (MWD/MT) ke/ft E I D (5 0 13.0

.E~ 10.0 X I 5,000 13.0 10,000 13.0 lQ 9.5 15,000 13.0 Cl C 20,000 13.0 25,000 11.3 9.0 30,000 9.4 35,000 7.9 8.0-5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 Bundle Average Exposure (MWD/MT)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure NIH!lkhS Figure 3.2.1-3 860599.2A AUF FXP RZL~+0 PVEL

4aoP o peaA 7 I og 10.5 10.0 10.01 10.01- 10.01 .. 10.01:

I L.

L 9.0 ~

O Cl Bundle CD Ol C 8.5- Average 4$ 0 Exposure MAPLHGR CO Q o 8.0 (MWOlMT) kwlft E o DQ 0 10.01 f ~o X

7.5 5,000 10.01 X 10,000 10.01 15,000 10.01 CD CJ 7.0 C

6.5 ~

6.0 0 5,000 1O,OOO ... 15,OOO 2O,OOO 25,OOO aO,QOO 35,OOQ 4O,OOO 45 000 BUNDLE AVERAGE EXPOSURE (HWO/N)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus BUNDLE AVERAGE EXPOSURE Figure 3.2.1-6 AMP FAf'E LYRICO FuEL

, GGN)/PE.gP GGPY MCPR OPERATING LIMITS FOR RATED CORE FLOW 05 L.E 7L ~ pE LLOYD)14& PLAcE E ui nt Status F 0 / AH~E MCPR 100 0

Core Flow eratin L't M6~ Core Flow

1. Normal" 1.27 ENC Fuel 1.27 ENC Fuel 1.28 GE fuel 1.28 GE Fuel

2. Control Rod Ins rtion Bounded by Tech. h pec. 1.32 Both F el 1.32 Both Fuel Limits (3.1.3.4 - 3/4 1-7) Type Types

3. RTP Inoperable, 1.32 E Fuel 1.33 ENC fuel Normal Scram 1.33 Fuel 1.34 GE Fuel This MCPR is based on the ENC re oad safety analyses performed using the control rod insertion time shown below (defined as normal scram).

In the event that surveillanc 4. 1.3.2 shows these scram insertion times may be exceeded, the p. ant thermal limits of Step l. above are to default to the values ip'tep 2. above and the scram insertion times must meet the requji'ements of Tech Spec..3. 1.3.4.

Slowes . Measured Average Control Rod Insertion Time to Specified Notches for Position Inserted From Each Grou of 4 Control Rods Arranged Full Withdrawn in a Two-b .-Two Arra Seconds)

Notch . 404 3P 45'otch

.660 Notch 5 1~504 Note 5 2.'624 WASHINGTON NUCLEAR " UNIT 2 3/4 2-7 Amendment No. 28

l.

2.

3.

4.

5.

RPT

• These 0 MWD MTU 4150 4150 4150 4150 Cycle

~Ex osure

- 4)50 MWD MTU HWD NTU MWD MTU MWD MTU

-

-

-

-

MWD MTU

'.

EOC MWD MTU EOC HWD EOC MTU MID MTU EOC MWD MTU operable and inoperable.

TABLE 3.2.3-1 MCPR OPERATING Normal scram LIMITS Equipment Status times**

Control rod insertion bounded by Tech. Spec.

limits (3.1.3.4 -

RPT inoperable Normal scram times RPT inoperable Control rod insertion p

bounded by Tech. Spec.

limits (3.1.3.4 - p 3/4 1-7) 3/4 1-7)

HCPR U

GE

• In this portion of the fuel cycle, operation with the given HCPR operating limits is allowed for both normal and Tech. Spec. scram times and for both to 29 1.32

1. 39

1. 37

l. 43 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 Operating Limit 1064 Core Flow Fuel ANF Fuel

1. 26

1. 30

1. 35 1.35 1.39 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 1-7),

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

>CNtN orsTNrauTmN LvnTsOONTROLLED 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 SR fuel shall not exceed the values shown in Figure 3.2.4-1.

AAJF APPLICASILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or E E EE.

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 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 />.

SURYEILLANCE 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 />, y

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 ROD PATTERN for LHGR.

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

2 CD n

I

))l EXP LHGR 2>>

R7 0 15.62 510 15.621 14 2,580 15.10 5,230 14.71 E

7)940 14.19 O 10,470 14.13 m 12 13,220 14.06 K

C 15,990 14.06 0 18,708 14.00 P4 I

21,590 13.93 CD 10 24,420 13.93 O

G Permissible 27,280 13.08 44 C0 Region of 30,150 12.24 Operation 33)050 11.40 8 35,960 10.47 38,900 9.55 P

41,830 8.65 44,760 7.77 10,000 20,000 30,000 . 40,000 50,000 Average Planar Exposure (MWD/MT)

Linear Heat Generation Bate (LHGR) Limit O

Versus Average Planar Exposure Xgaaw 8 x 8 Fuel RELo40-Figure 3.2A-1 860599.3A

INSTRUMENTATION 3/4. 3. 10 NEUTRON FLUX MONITORING INSTRUMENTATION LIMITING CONOITION FOR OPERATION 3.3.l0 The.APRM and LPRM" neutron flux noise levels shall not exceed three (3) times their established baseline value dp Figuw~ >.3./D -3

~~ ~~~~~+ /y ~~ >y/>~py~ ~z>~

APPLICABIL'ITY: OPERATIONAL CONOITION I with two reactor coo'lant system recir culation loops in operation with THERMAL POWER greater than the limit specified in Figure 3.3. 10-1 and total core flow less than 45K of rated total core flow or with one reactor coolant system recirculation loop not in operation with THERMA POWER greater than the limit specified in Figure 3.3.10-l.

ACTION:

Q.With the APRM or LPRM" neutron flux noise level greater than three (3) times their established baseline noise levels, initiate corrective action within 15 minutes to restore the noise levels 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 or equal to the limit specified in Figure 3.3.10-1 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

SURVEILLANCE REQUIREMENTS 4.3.10.1 The provisions of Specification 4.0.4 are not applicable.

4.3.10.2 With two reactor coolant system recirculation loops in operation, establish a baseline APRM and LPRM" neutron flux noise level value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> upon entering the APPLICABLE OPERATIONAL CONDITION of Specifica-tion 3.3.10 provided that. baselining has not been performed since the most recent CORE ALTEPATION.

4.3.10.3 With one reactor coolant system recirculation loop not in operation, estab 1 i sh a baseline APRM and LPRM" neutron flux noi se 1 evel val ue with THERMAL POWER less than or equal to the limit specified in Figure 3.3.10-1 prior to entering the APPLICABLE OPERATIONAL CONOITION of Specification 3.3.10 provided baselining has not been performed with one reactor coolant system recirculation loop not in operation since the most recent CORE ALTERATION.S 1'h 8lco iw, &m upsy Zo~mi'e~imo ~jtgsrrcZ 3rd-/

&tvMky'vc 4f/IA 4)/@in /cFNznu6~ zd fakcP+ocQ r- g+@ ~wltol

~4 ~ /n/e/-8/ >arMP> <r/Wc fr~ dml~ dA ymyZ dn~eJ . (...l

~pi~ ur'vari(ZAN>.

WASHINGTON NUCLEAR UNIT 2 3/4 3-102 Amendment No. 16

GQNTRQLLED GQPY INSTRUMENTATION NEUTRON FLUE MOHITORING IHSTRUM'EHTATIOH SURVEILLANCE REQUIREMENTS Continued) 4.3,10.4 The APRM and LPRM" neutron flux noise levels shall be determined to be less than or equal to the limit of Specification 3.3.10 when o crating within the

a. 't APPLICABLE OPERATIONAL CONDITION least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and of Specification 3.3.1:

b. Within 30 minutes after completion of a THERMAL POWER increase of at least 5X of RATED THERMAL POWER.

d4 ~ ~~krp~r/~ 8l~ >Aa!Ib~ V~ikez &Be gu&id~

af Hgzuwe Z-g/d-g.

"Detector levels A and C of one LPRM string per core octant plus detector levels A and C of one LPRM string in the center of the core should be monitored.

PThe baseline data obtained in Specification 4.3. 10.3 is applicable to opera-tion with one reactor coolant system recirculation loop not in operation and THERMAL POWER greater than the limits specified in Figure 3.3. 10-1.

4 WASHINGTON NUCLEAR - UNIT 2 3/4 3-103 Amendment No. 16

I ~

) ~

00NTRGLLEB GQPY REACTIVITY CONTROL SYSTEMS BASES 3/4. 1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the safety analyses, and (3) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic re-quirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The re-quirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will/ be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent, operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted=to be taken out of service provided that those in the nonfully inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable, could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

y'pr, cyca 6 S'pgc(F tc 7gausiau> AN4~YS>>

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than he fuel cladding safety limit during the core wide transient analyzed in This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives 'as given in the specifications, provide the required protection and MCPR remains greater than the fuel cladding safety limit. The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reac-tor for long periods of time with a potentially serious problem.

The scram discharge volume, is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3. 1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may sti'll be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

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

~ ~

GGNTRGLL.ED QQPV 3/4. 2 POWER OISTRIBUTION 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

~N 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 Specification for fuel is specified to assure the PCT following a postulated LOCA A< P APLHGR 2>>F;; T Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3 for two recirculation loop operation.

These values shall be multiplied by a "factor of 0.84 for single recirculation loop operation. This multiplier is determined from comparison of the limiting analysis between two recirculation loop and single recirculation loop operation.

The calculational procedure used.to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3 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 Reference 1 or XN-NF-80-19, Volumes 2, .2A, 2B and 2C, Rev. 1.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2"1 Amendment No. 28

~ ~

GGNTRGLLED GGPV POWER OISTRIBUTION 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.

To 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"l.

cycle 5'rEelric.. rhea ssao~ A MA> ystp go r oR7 A4'F The evaluatio I f a given transient begins with t nuptial param-eters shown in that are input to a ore dynamic behavior tran-sient computer program. The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle.

The codes and methodology to evaluate pressurization and nonpressurization events are described in XN-NF-79-7 The principal result of this evaluation is the reduction in MCPR caused by fthe transient.

4~0 xe-asF - re so5'j'4? ~

The purpose of the MCPRf of Figure 3.2.3-1 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow the required MCPR is the maximum of the rated flow MCPR 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 will 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. 28

PLANT SYSTEMS BASES In the event that portions of the fire suppression systems are inoperable, alternate backup fire fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.

When the inoperable fire fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

The surveillance requirements provide assurances that the minimum OPERABILITY requirements of the fire suppression systems are met. An allowance is made for ensuring a sufficie'nt volume of Halon in the Nalon storage tanks by verifying the ~eight and pressure of the tanks.

In the event the fire suppression 'water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant~>~

3/4. 7. 7 FIRE" RATEO ASSEMBLIES The OPERABILITY of the fire barrier's'nd barrier penetrations ensure that fire damage will be limited. These design features minimize the possibility of a single fire involving more than one fire area prior to detection and extinguishment. The fire barriers, fire barrier penetrations for conduits, cable trays and piping, fir e windows, fire dampers, and fire doors are periodically inspected'o verify their OPERABILITY.

3/4.7.8 .AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY.

3/4. 7.9 MAIN TURBINE BYPASS SYSTEM z y QpQ ppQQ(rf 4, 8 N 4 4y'5/So The main turbine bypass system is required to be OPERABLE consist with the assumptions of the feedwater controller failure analysis of The main turbine bypass system provides pressure relief during the feedwater controller failure event so that the safety. limit MCPR is not violated.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 7-4