Text

Proposed Tech Specs,Reflecting Cycle 2 Reload
	ML17304A832
Person / Time
Site:	Palo Verde
Issue date:	12/14/1988
From:	; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	;
Shared Package
ML17304A830	List: ML17304A826; ML17304A832;
References
	NUDOCS 8812300049
	Download: ML17304A832 (174)
	v • d • e

G. MARKED-UP TECHNICAL SPECIFICATION CHANGE'AGES Index' IV XIX Limiting 'Conditions for Operation and Surveillance Requirements:

3/4 1-21 3/4 1-22, 3/4 1-23 3/4 1-24 3/4 1-25 3/4 10-2 3/4 10-4 Bases for L'imiting Conditions for Operation .and'urveillance Requirements:

B 3/4 1-6 B 3/4 1-7 8812800049 SS1214 4-4 PDR ADOCK 05000530 P PDC

CONTROLLED BY USER INDEX LIMITIHG CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION .PAGE 3/4. 0 APPLICABILITY............................................. 3/4 0-1 3/4.1 REACT VITY CONTROL SYSTEMS 3/4.l. 1 BORATION CONTROL" SHUTOOMN MARGIN - ALL CEAs FULLY INSERTED............. 3/4 1-.1 SHUTDOMN'ARGIN - 'KN " ANY CEA MITHDRAMN........;... 3/4 1-2 1

MODERATOR TEMPERATURE COEFFICIENT..................... 3/4. 1-4 MINIMUM TEMPERATURE FOR CRITICALITY.................. 3/4 1-6 3/4. 1- 2 BORATION SYSTEMS FLOM PATHS " SHUTDOMN.......:... 3/4 1-?

FLOM PATHS " OPERATING............. 3/4 1-8 CHARGING PUMPS " SHUTOOMH. .. . .. . .. . ..- -. 3/4 1"9 CHARGING PUMPS " OPERATING. -... -........ - -. - - - -.' >> - -. 3/4 1-10 BORATED MATER SOURCES " SHUTDOMH. 3/4 1"11 I

BORATED MATER SOURCES - OPERATING...................... 3/4 1-13 BORON DILUTION ALARMS........... 3/4 1->4 3/4.1. 3 MOVABLE CONTROL ASSEMBLIES CEA POSITION................................. 3/4 1-21 POSITION INDICATOR CHANNELS " OPERATING...... ~ ~ ~ ~ ~ ~ ~ t ~ 3/4, 1"25 POSITION INDICATOR CHANNELS' SHUTDOMH........ -....... 3/4 1" 26 CEA DROP TIME---.....-..... 3/4 1-27 SHUTDOMH CEA IttSERTIOH LIMIT. 3/4 1-28 PPP j REGULRTIt(G CEA IHSERTIOt< LIMITS LEGAL TR GE/~ ~'R'MET70/ I LIB/

3/4 1-2e

~/i (-sa PALO VERDE - UttIT 3 IV Ath'Et/C,'1St)T tf0. 2

4i Cl Il

, CONTROLLED BY USER INDEX LIST OF FIGURES PAGE

3. 1-1A SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE....,......... 3/4 1-2a

3. 1-1 ALLOWABLE MTC MODES 1 AND 2............................ 3/4 1-5

3. 1" 2 MINIMUM BORATED WATER VOLUMES......".................... 3/4 1-12

3. 1" 2A PART LENGTH CEA IHSERTION LIMIT VS THERMAL POWER....... 3/4 1-23

3. 1" 2B CORE POWER LIMIT AFTER CEA DEVIATION................... 3/4 1"24

3. 1-3 CEA INSERTION LIMITS- VS THERMAL POWER (COLSS IN SERVICE).................... - ~.......... >>. - -. 3/4 1-31

3. 1-4 CEA INSERTION LIMITS VS THERMAL POWER (COLSS OUT OF SERVICE).........................--.- ~ --. ~ 3/g 1-32 3tS'.

2"1

'fat t.EPL6Tlk CEA W5EB.7?Obl UHll- I5 TREi&+I DNBR MARGIN OPERATING LIMIT BASED OH COLSS

~P 3fj /-

(COLSS IN SERVICE).............................--.-.... 3/4 2-6 302 2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR (COLSS OUT OF..

I SERVICE)...................... 3/4 2-7 3%2 3 REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVELo s ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ * ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-10

3. 3"1 DNBR MARGIN OPERATING LIMIT BASED ON COLSS FOR BOTH CEAC'S INOPERABLE............................. 3/4 3-10

3. 4-1 DOSE E(UIVALEHT I"131 PRIMARY COOLANT SPECIFIC ACTIVITY'IMITVERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY

> 1.0 pCi/GRAM DOSE EQUIVALENT I-131... 3/4 4-27

3. 4-2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 1O YEARS OF FULL POWER OPERATION... 3/4 4"29

, 4'.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST 3/4 7-26 B 3/4.4"1 NIL"DUCTILITYTRANSITION TEMPERATURE INCREASE AS A.

,FUNCTION OF FAST (E > 1 MeV) NEUTRON FLUENCE (550 F IRRADIATION). B 3/4 4-10

5. 1-1 SITE ANO EXCLUSION BOUNDARIES......... 5-2

5. 1-2 LOW POPULATION ZONE ....... 5-3

5. 1-3 GASEOUS RELEASE POIt]TS 5-4

6. 2-1 OFFSITE ORGAttIZATIOt]. 6-3

6. 2-2 ONSITE UttIT ORGAHIZAT'ION..... 6-4 PALO VERDE - UNIT 3 XIX A jEt(0 tEHT NO,

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~ r CONTROLLED BY USER

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REACTIVITY CONTROL SYSTEMS 3/4. 1. 3 HOVABLE CONTROL ASSEMBLIES CEA POSITION LIHITING CONDITION FOR OPERATION 3.1.3.1 All full-length (shutdown and'egulating) CEAs, and all,part-length CEAs which are inserted in the care, shall be OPERABLE with each CEA of group positioned within 6.6 inches (indicated position) of all ather a'iven CEAs in its group. &~~ength-6~roups-r h 11-b i d- ,

APPLICABILITY: MODES 1" and 2".-

ACTION:

a. Mith one or more full"length CEAs inoperab'le due to being immovable as a result of excessive friction',or mechanical interference or known to be untrippable, determine that the SHUTDOMN MARGIN require-ment of Specification 3.1;.1.2's satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. Mith more than one full-length or part-length CEA inoperable or misaligned fram any other CEA in its group by more than 19 inches (indicated position), be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c Mith one or more full-length or part-length CEAs misaligned from any other CEAs in its group by more than 6.6 inches, operation in MODES 1 and 2 may continue, provided. that core power is reduced in accordance with Figure 3. 1-2$ and that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the misaligned CEA(s) is either:

p

1. Restored'o OPERABLE status within its above specified alignment requirements, or

2. Declared inoperable and the SHUTDOWN HARGIN'equirement of Specification 3. 1.1.,2 is satisfied. After declaring the CEA(s)

,inoperable, operation in HODES 1 and .2 may continue pursuant to the requirements of Specification~3. 1.3.6lrprovided:

~d. 3, i.3,g a) Mithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the remainder of the CEAs in the group, with, the inoperable CEA(s) shall be aligned to within 6.6 in-ches of the inoperable CEA(s) while maintaining the

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allow-'nd the THERHAL POVER level TQ5+I+Qrt+ fop . Speci ficetio( 3. 1.,3. 6(dur-ing subsequent operatian. QnCt.3.l ABACA 31 3 (

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"See Special Test Exceptions 3.10.2 and 3.10.4.

PALO VERDE " UNIT 3 3/4 1-21 AMENDMENT NO.

0 0

CONTROLLED BY USER LIHITING CONDITION FOR OPERATION (Continued)'CTION:

(Continued) b) The SHUTOOMN MARGIN requirement of Specification 3.1.1.2 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ..

d. Mith one full-length CEA inoperable due to causes other than addressed by ACTION a., above, but within its above specified align-ment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6.

e. Mith one part-length CEA inoperable and inserted in the core, operation may continue provided the alignment of the inope'rable part length CEA is maintained within 6.6 inches (indicated position) of

.all other part-length CEAs in its groups Ore] 'ltd~ g~ (6 Inaiivkun~I

>VF~ODrl +0 ttlP T8 U)KE(Y}@1& cf-5 Qca+g~ P,/,+,

Mith part length CEAs i erted, beyon sn er son >ms s excep or surveillance testing pursuant to Specific ion 4.1.3.1.2, wi hin 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

Restore the pa length CEAs to within their lim s, or I

2., Reduce THE AL POMER to les than or equal t that fraction of RATED HERHAL POMER wh ch is allowed by art length CEA group positi using Figure 3 -2A.

SURVEILLANCE RE UIREHENTS 4.1.3.1.1 The position of each full-length and part-length CEA shall be deter-mined to be within 6.6 inches (indicated position) of all other CEAs in its group at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then verify the individual CEA positions at least once per 4,hours.

4.1.3. 1.2 Each full-length CEA not fully inserted and each part-length CEA which is inserted in the core shall be determined to be OPERABL'E by movement of at least 5 inches in any one direction at least once per 31 days.

PALO VERDE - UNIT 3 3t4 I-Z2 Ai1ENC'sENT NO. 2

0 0

0

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m 1.00 112.5" C7 m 0.90 o.so 0 ACCEPTABL~

0.70 OPERATION UNACCEPTABL OPER TION O

x 060 0 z

8 050

~OX POWER LINE XJ O

O 0.40 m

0.30 U

O tXI 0.20 t:

fig 0.10 Xl 22.5" 0.00 150 140 130 120 110 10Q 90 8Q 70 60 50 40 . 30 . 20 10 0 PART LENGTH CEA POSITION, INCHES WITHDRAWN FIGURE 3.1-2A PART LENGTH CEA INSERTION LIMIT VS. THERMAL POWER

0 0

GONTROLLED BY USFR 0

~ C g LLI m O 30 LU

~ 20 (60 MlN, 2CYo) g ~

og Q

O~

gb 2acI~

10 I

I ~

LU ~ '0 QO

~O 0 10 20 30 40 50 60 R TIME AFTER DEV)ATION, MINUTES WHEN CORE POWER'IS REDUCED'TO 55% OF RATED THERMAL POWER PFR THIS LIMIT'CURQE, FURTHER REDUCTION IS NOT REQUIRED

( .CORE PO'4ER FIGURE 3.

LIMIT AFTER 1-2 A

CEA DEVIATION" PALO VE'RDE - UNI,T 3 3/4 I-24

'0 POSITION INDICATOR CHANNELS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA:

a. CEA Reed Switch Position Transmitter (RSPT 1) with the capability of determining the absolute CEA positions within 5.2 inches,

b. CEA Reed Switch Position Transmitter (RSPT 2). with the capability of determining the absolute CEA positions within 5.2 inches, and

c. The CEA pulse counting position indicator channel.

APPLICABILITY: MODES 1 and 2.

ACTION:

With a maximum of one CEA per CEA group having only one of the above required CEA position indicator channels OPERABLE, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

a. Restore the inoperable position indicator channel to OPERABLE status,. or

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b. Be in at least HOT, STANDBY, or C. Position the CEA group('s) with he ino erable position indicator(s) at its full'y withdrawn positi while maintaining the requirements of Specifications 3. 1.3. 1> . 1.3. . Operation may then continue provided the CEA group(s) with the inoperable position indicator(s) is maintained fully withdrawn, except during surveillance testing pursuant to the requirements of Specification 4. 1. 3. 1. 2, and each CEA in the group(s) is verified fully withdrawn at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its "Full Out" limit~.

SURVEILLANCE RE UIREMENTS 4.1.3.2 Each of the above required position indicator channels shall be determined to be OPERABL'E by verifying that for the same CEA, the position indicator channels agree within 5.2 inches of each other at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CEAs are ful:ly withdrawn (Full Out) when withdrawn to at least 144.75 jnches.

PALO VERDE - UNIT 3 3/4 1'-25

0 II

CONTROLL"0 BY USER AGi3 pQP~

REACTIVITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION Z,l,z.'I The, o,~I- l~+I ~ clro~ s~ ~d~,~In+i~~

+Laa'p iud I

Ice'r ool.+~erOId ',

<< ~><<M&.> OPERON&

R4-~ The part length CEA groups shall be limited to the insertion limits shown on Figure 3. 1-5 with PLCEA insertion .between the Long. Term Steady State Insertion, Limit and the Transient Insertion Limit restricted to:

) gC < 7 EFPO per 30 EFPD interval, and 0-

$ $. < 14 EFPO per calendgr year.

I. 5W LeR~ ~oremLE.

dt's lenqW CRR group Pottcry Qiuuti'PP@IIIQ &lkI lOV\ $ + t tI"lt7~ f~WQ +4'L~ )8 ~4- T~>ICAf

~ttq 1"LlwuW utah~ Qo~ CH$~ Ol".~ IPodperWI8 p APPLICABILITY: MODELS and 2" ACTION:

as With the part length CEA groups inserted beyond the Transient Insertion Limit, except for surveillance testing, pursuant to Specification 4. 1.3. 1. 2, within two hours, either:

1. .Restore the part length CEA group to within the limits, miami or 2,. ResIuce. QQRAAI POKER a& &./Ious',

a.) One, Ei" ~OTO. Q,ER~ OPZRA3CE I) Reduce THERMAL POWER to less than or equal to that Traction of RATED THERMAL POWER which is allowed by the PLCEA group position using Figure 3. 1-5+ Qr g) I2R. jr'L+ l~s& HEII ATPb I O'I rt ho~,

I) lorn m~~ ~O~E~~LE-Se. ir <+ leaSH HIIT'TPI+tk8$

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.b. With the part length CEA groups inserted between the; L'ong Term Steady State Insertion Limit and the Transient Inserts'on L'imit for intervals > 7 EFPD per 30 EFPD interval. or > 14 EFPO'q)er'calendar year, either:

l. 5 Restore the part length groupqwithin the Long T'dry Steady State Insertion Limit within'two hours, or

2. 'Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.1.3.7 position 5 of the part length CEA grouppshall be deter'mined to be

~

The within'he Transient nsertion Limit at least once per 12 bours/"-;, +

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ddri Y'ei ~

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G~p~ P 'lou~

IA4NAtlD k)tlat haW ye+ 'tenq~ CSR qr0~P paSI+IOn+ ..'.U~

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"See Special Test Exceptions 3. 10. 2 and 3. 10. 4.

PALQ YEROE - UNIT R 3/4 1-33

4l 0

COIMTROLLEO BY USER A~ P~

0 IO 20 30 40 50 R

60 70 TRANSIENT INSERTION LIMIT (75.0 INCHES ) o i/l 80 o CL UNACCEPTABLE RESTRICTED OPERATION OPERATION IOO LONG TERM STEADY STATE INSERTION LIMIT

'II2.5 INCHES)

I30 I40 ISO oo oCl o

o o

o t

o oCP o

eo o

r o

o on o

o>

o o

o oo o

FRACTION OF RATED THERMAL POlYER FIGURE 3.1-5 PART LENGTH CEA IHSERTIOH LIMIT VS THERMAL POMER PALO VERDE - UNIT g Q 3/4 1-34

t C

is 0

CONTROLLED BY USER SPECIAL TEST EXCEPTIONS 3/4. 10. 2 MODERATOR TEMPERATURE COEFFICIENT GROUP HEIGHT INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3 I)i ~>>~$

3.10.2 e moderator temperature coefficient, group height, insertion, and power di tribution limits of Specifications 3.1.1.), 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7,'and the Minimum Channels OPERABLE requirement of I.C.l (CEA Calculators) of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS."provided:

a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85K of RATED THERMAL POWER, and

b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4. 10.2.2 below.

APPLICABILITY: MODES 1 and 2.

ACTION: g,i,s9 With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C.l (CEA Calculators) of Table 3.3-1 are suspended, either:

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or

b. Be in HOT STANDBY within 6 hours.

. SURVEILLANCE RE UIREMENTS r 3 i 5."))

4.10.2.1 The THE HAL POWER squall be determined at least once per hour during PHYSICS TESTS in hich the requirements of Specifications 3. 1. 1.3, 3. 1.3. 1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3,,3.2.7, or the Minimum Channels OPERABLE require-ment of I. C. 1 (CEA Calculators) of Table 3.3-1 are suspended and shall be verified to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits'of 3.2; 1 by monitoring it continuously with the Incore Detector 'pecification Monitoring System pursuant to the requirements of Specifications 4.2.1.2 and 3,3.3.2 during PHYSICS TESTS above 20' of RATED THERMAL POWER in which the requi rements of Specifications 3. 1. 1. 3, 3. 1. 3. 1, 3. 1. 3. 5, 3. 1. 3. 6, 3. 2. 2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of I.C. (CEA Calculators) of Table 3.3-1 are suspended. A 3:l.3.7, PALO VERDE = UNIT 3 3/4 10-2

0 CONTROLLED BY USER j~~~@SPECIAL TEST DCEPTIONS 3/4.10.4'EA POSITION REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLO LEG TEMPERATURE LIMITING CONDITION FOR OPERATION i3 l.3.'I 3.10.4 The, requirements of'Specifications 3.1.3.1, 3.1'.3.6)and-3.2.6 may be suspended during the performance of PHYSICS TESTS to determine the isothermal temperature coefficient, moderator temperature coefficient, and'ower coefficient provided the limits of Specification 3.2. 1 are maintained and determined as specified in Specification 4.10.4.2 below.

APPLICABILITY: MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3. 1.3. 1,II3.1.3.6'fand 3.2.6 are suspended, either:

~,),g,g ) ~'I!I%I l

a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2. 1, or

b. Be in HOT STANDBY within 6 hours.

~ SURVEILLANCE RE UIREMENTS Z,i.E. ~~

4.10.4. 1 The THERMAL POWER shall be determined at least once per our during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1, 3.1.3.6~3Al.SIP and/or 3.2.6 are suspended and shall be verified to be within the test power plateau.

4. 10.4.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring Monitoring System pursuant to the it continuously with the Incore Detector requirements of Specification 3.3.3.2 during PHYSICS TESTS above 20K of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1, 3 . 1.3.6l'and/or 3.2.6 are suspended.

E, l,3 i'I 3, l S.S)

(

PALO VERDE - UNIT 3 3/4 10-4

0 0

CONTROLLED BY- USER.

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES Continued and load maneuvering. Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and .from these analyses CEA insertions are determined and a consistent set of radial peaking factors defined. The Long Term Steady State and Short Term Insertion Limits are deter-mined based upon the .assumed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used. The limits speci--

fied serve to limit the behavior of the radial peaking factors within the bounds determined from analysis. The actions specified serve to .limit the extent of radial xenon redistribution effects to those accommodated in the analyses. The Long and Short Term Insertion Limits of Speeificatio@3. 1.3. are, specified for the plant which has been designed for primarily baseloaded o eration but which has the ability to accommodate a limited amount of load maneuv ring.

~Qadi) ~7 The Tr8nsient Insertion Limits of Specification 3.1.3. and the Shutdown CEA Insertion Limits'f Specification 3.1.3.5 ensure that (1) the minimum SHUT-DOMN MARGIN is maintained, and (2) the potential effects of a CEA egection accident are limited to acceptable levels. Long-, term operation at the Tran-Insertion Limits is not permitted since such operation could have effects 'ient on .the core power distribution which could invalidate assumptions used to deter-mine the behavior of the radial peaking factors..

The PVNGS CPC and COLSS systems are responsible for the safety ancf monitoring functi'ons, respectively, of the reactor core. COLSS monitors the 'DHB Power Operating Limit'(POL} and various operating parameters to help the operator main-tain plant operation within the limiting conditions for operation (LCO). Operat-ing within the LCO guarantees that in the event of 'an Anticipated Operational Occurrence (AOO), the CPCs wi 11 provide a reactor trip in time to prevent un-acceptable fuel damage.

The COLSS reserves the Required Overpower Margin (ROPM) to account for the Loss'of Flow (LOF) transient which is the limiting A00 for the PVNGS pTants.

Mhen the COLSS is Out of Service (COOS), the monitoring function is performed yia the CPC calculation of DNBR in conjunction with a Technical Specification COOS Limit"Line (Figure 3.2-2) which restr'icts the reactor power sufficiently to preserve the ROPM.

The reduction of the CEA deviation penalties in accordance with the CEAC (Control Element Assembly Calculator) sensitivity reduction program has been performed. This task involved setting many of the inward single CEA deviation penalty factors to 1.0. An inward CEA deviation event in effect. would no< be accompanied by the application of the CEA deviation penalty in either t.he CPC.

DNB and LHR (Linear Heat Rate) calculations for hose C As with the reduced penalty factors. The protection for an inward CEA aeviatton event is thus accounted for separately.

PAID VERDE - UNIT 3 8 3/4 1-6

a CONTROLLED BY USER REACTIVITY CONTROL SYSTEHS BASES HOVABL'E CONTROL ASSEHBLIES Continued 7

If an inward CEA deviation event occurs, the current CPC algorithm applies two penalty factors to each of the ONB and LHR calculations. The first, a static penalty factor, is applied upon detection of. the event. The second; a xenon redistribution penalty, is applied linearly as,a function of time after the CEA drop. The expected margin degradation for the inward CEA deviation event

~

-+f exeeeds-t4 Figure 3. l-2E)is required.

a power

'~RWI for which the penalty factor has been reduced is accounted for in two ways.

The ROPM reserved in COL'SS is used to account for some of the margin degrada-reduction in accordance with the curve in In addit'ion, the part length CEA maneuvering is restricted in accordance with Figure 3.1-gA'o justify reduction of the PLR deviation'enalty factors.

The technical specification permits plant operation if both CEACs are considered inoperable for safety purposes after this period.

g-PALO VEROE - UNI,T 3 8 Ri'-7

ATTACHMENT 5 - RESPONSE TIMES FOR RCP SHAFT SPEED A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes the response time of the DNBR -Low Reactor Coolant Pump (RCP) shaft speed trip in Technical Specification Table 3.3-2.

The change is due to redefining the events which take place before the Control Element Assemblies (CEAs) drop into the core. During Cycle 1, the response time of 0.75 seconds was measured from the time a trip condition existed, such as a loss of power to the RCP motors, to the moment the Control Element Drive Mechanism (CEDM) coil breakers opened. During Cycle 2 operation, the response time of 0.3 seconds will be defined from the time a signal is sent down the RCP shaft speed sensor line to the CPCs, to the moment the CEDM coil breakers open.

B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of Technical Specification 3.3.1 is to ensure that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained'o permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

RCP shaft speed is used by the Core Protection Calculators (CPCs) to compute the core coolant mass flow rate. The coolant flowrate is then used by the CPCs in the DNBR calculation.

C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT During the Cycle 1 startup testing, it was found that the projected reactor coolant flow rate trip software, housed in the CPCs, which monitors the RCP shaft speed and projects what the RCS flow will'e in the future, was too sensitive to smal'1 deviations in RCP shaft speeds and caused unnecessary trips of the unit. To correct this problem, the software dealing with the projected flow rate trip was taken out. In its place, software which trips the reactor when the RCP shaft speed slows to 95% of its normal speed will be installed.

Because of this change, the response time, as defined for the RCP shaft speed trip, has been redefined for Cycle 2 to reflect the purpose. of the new trip.

As a result of the redefinition of the response time, the safety analysis for Cycle 2 has, taken credit for the faster time. To ensure that the reactor is operated within the safety analyses, Table 3.3-2 will have to reflect the response time that was used in the safety analyses.

D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining .whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in 5-1

accordance with a proposed amendment would not: '(1) involve a significant increase in the probability or consequences of an accident previ'ousl'y evaluated;, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a si'gnificant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change does not i'nvolve .a significant increase in the probability or consequences of an accident previously evaluated because the changed response time ensures, sufficient margin: for .mitigating the most limiting Design Basis Event (DBE)., The Cycle 2 safety analysis results are still bounded by the reference cycle analysis. Therefore,.

there is no increase in the probability or consequences of an accident previously evaluated.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new or different kind of accident from any accident ,previously evaluated 'because the change maintains the margin of safety. The redefinition of the response time ensures that the results of the Cycle 2 safety analysis will remain within the bounds of the Specified Acceptable Fuel:Design Limits (SAFDLs) and, by maintaining the 0.3 second'esponse time,, the unit will be operated within the realm of the safety analysis. Therefore, the change will not create the possibility of a new or different kind of accident.

Standard 3--Involve a significant reduction in a, margin of safety.

The proposed change does not involve a significant reduction in a margin of safety because the change ensures that the margin of safety for Cycle 2 is, maintained. The analysis results show. that there is sufficient margin to mitigate the most limiting DBE and that the results are bounded by the reference cycle. Therefore,. no reduction in margin will arise.

The .proposed amendment matches the guidance concerning, the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:.

( j.ii) For a nuclear, power reactor, a change resulting from a .nuclear reactor core reloading, if no fuel assembl'ies signifi'cantly different from those found previously acceptable to the NRC for a previous core at the facility in questi'on are involved. This assumes that no significant changes .are made to the acceptable criteria for the Technical Specifications',. the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not signifi'cantly changed, and that NRC has previously found such methods acceptable.

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E. SAFETY EVALUATION FOR THE AMENDMENT RE VEST The proposed Technical'pecification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated'n the FSAR. The proposed change does not change or replace equipment or components which are important to safety. The change refl'ects the actual response time of the trip circuitry.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The change maintains the margin of safety. The redefinition of the response time ensures that the results of the Cycle 2 safety analyses will remain within the bounds of the SAFDLs and, by maintaining the 0.3 second response time, the unit will be operated within the realm of the safety analysis. This does not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the bases for the Technical Specifications. The change ensures that the margin of safety, for Cycle 2 'is maintained. The analysis results show that there is sufficient margin to mitigate the most limiting DBE and that the results are bounded by the reference cycle. Therefore, no reduction in margin will arise.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 3, in accordance with this change, would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental 'Statement (FES) as modified by ,the staff's testimony to the Atomic Safety and Licensing Board; or

2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.

MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions for Operation and Surveillance Requirements:

3/4 3-11 5-3

0 li

TABLE 3.3-2 D') REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TINES

( FUNCTIONAL UNIT RESPONSE TIME C) m I. TRIP GENERATION A. Process

1. Pressurizer Pressure - High < 1.15 seconds

2. Pressurizer Pressure - Low < 1.15 seconds

3. St.earn Generator Level - Low < 1.15 seconds Steam Generator Level - High < 1.15 seconds

5. Steam Generator Pressure Low < 1.15 seconds

6. Containment Pressure - High < 1.15 seconds

7. Reactor Coolant Flow - Low < 0.58 second

8. Local Power Density - Higlt

a. Neutron Flux Power from Excore Neutron Detectors < 0.75 second*

t). CEA Positions < 1.35 second**

c. CEA Posit.ions: CEAC Penalty Factor < 0.75 second**

9. DNBR - Low

a. Neutron Flux Power from Excore Neutron Detectors < 0.75 second"

b. CEA Positions < 1.35 second""

c. Cold Leg Temperature < 0.75 secondH

d. llot Leg Temperature < 0.75 secondÃ

e. Primary Coolant Pump Shaft Speed - seconds

f. Reactor Coolant Pressure from Pressurizer < 0.75 second8h'0

g. CEA Positions: CEAC Penalty Factor < 0.75 second"*

B. Excore Neutron Flux Or30

l. Variable Overpower Trip < 0.55 second*

2. Logarithmic Power Level - High

a. Starts>p and Operating Z' 55 second*

b. Shutdown < 0.55 second*

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ATTACHMENT 6 - PDIL A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE VEST The proposed amendment revises Technical Specifications (T. S.) 3/4.1.3.5 and 3/4.1.3.6 to address the shutdown and regulating Control Element Assembly (CEA) insertion limits specifically for 1 or 2 CEACs out of service. In addition, the CEA insertion limits shown on Figures 3.1-3 and 3.1-4 are revised. Operation of the regulating CEAs during Cycle 2 will be more limited than in Cycle 1. The revisions to the curves will maintain the margin of safety and insure that there will be sufficient shutdown margin to handle the most limiting Anticipated Operational Occurrence (AOO) and limiting fault events.

The clarifications to the'ext of T.S. 3/4.1.3.5 and 3/4.1 3.6 were not part

~

of the Unit 2 Cycle 2 submittal. These changes will be incorporated into the Unit 1 and Unit 2 T.S.

B. PURPOSE OF THE TECHNICAL 'SPECIFICATION The purpose of T.S. 3/4 '.3.5 and 3/4.1.3.6 is to ensure that (1) acceptable power distribution limits are, maintained, (2) the minimum shutdown margin is maintained, and. (3) the potential effects of CEA misalignments are limited to acceptable levels.

'EED FOR THE TECHNICAL SPECIFICATION AMENDMENT The proposed changes made to the CEA Insertion Limits are due to the change in the Cycle 2 core physics. Because of the change to the core, the worth of the CEAs has changed and as a result, the effects of the dropped and ejected CEA events change. To ensure that there is. sufficient margin to mitigate such events, CEA insertion has to be restricted by the insertion limits set forth in the proposed Figures 3.1-3 and 3.1-4. The changes in the text of T.S.

3/4.1.3.5 and 3/4.1.3.6 are to clarify the T.S. with respect to CEACs in or out of service.

BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS 'CONSIDERATION DETERMINATION

1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not:,(1) Involve a significant increase in the probability of consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

A discussion of these standards as they relate to the amendment request follows:

Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.

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0

The, proposed change does not involve a significant increase in the probabi.lity or consequences of an accident previously evaluated because the figures were updated to be consistent with the Cycle 2 safety analyses. Using the CEA insertion limits as given on Figures 3.1-3, and 3.1-4 assures that there is suffi:cient margin for the most limiting Design Basis Events. The analyses performed includes an evaluation of all safety analyses events for which the CEA insertion limit curve serves as an initial condition. The format change was required to clarify the T.S. for the conditions of CEACs in or out of service. Therefore, there is no change in the probability or consequences of an accident occurring.

Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change will not create the possibility of a new 'or different kind of accident from any acci'dent previously evaluated because the proposed changes to Figures 3.1-3 and 3.1-4 are required to make the graphs consistent with the Cycle 2 safety analyses. Operation of the reactor within the limits as shown on the proposed figures will ensure that the SAFDLs will not be exceeded during the most limiting AOO. The Cycle 2 figures were based on meeting the same criteria as the Cycle 1 figures. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated will not be created.

Standard 3--Involve a significant reduction in a margin of safety.

The proposed'hange does not involve a significant reduction in a margin of safety because the proposed change is being made to make the technical specifications consistent with the Cycle 2 safety analyses. Operation of the reactor within the limits as shown on the proposed figures will ensure that the SAFDLs: will not be. exceeded during the most limiting AOO.

The Cycle 2 figures were based on the same design criteria as the Cycle 1 figures.

2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:

A purely administrative change to the technical specifications:

for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

and (iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications,. the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.

6-2

Ck 45

E. SAFETY EVALUATION FOR THE AMENDMENT.RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.,

Therefore, there is no increase in the probability of occurrence or the consequences of an accident occurring.

The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously eval'uated in the. FSAR. The proposed change places limits on the insertion of the CEAs such that the results from any accident occurring, while within the bounds set by: T.S. Figure 3.1-3 and 3.1-4, will have the same consequences as those determined for the reference cycle. Thus, the proposed change is a result .of maintaining the Cycle 2 safety analysis results within the reference cycle bounds and no new or different kinds of accidents will be created.

The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. The proposed change is being made to maintain Cycle 2 margin of safety and sufficient shutdown margin for the most limiting AOO and limiting fault events.

F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed .environmental question because operation of PVNGS Unit 3, in accordance with this change, would, not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the, Atomic Safety and 'Licensing Board; or

'2. Result in a significant change in effluents or power levels; or

3. Result in matters not previously reviewed in the licensing. basis for PVNGS which may have a significant environmental impact.

G'. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions for Operation and Surveillance Requirements:

3/4 1-28 3/4 1-29 3/4 1-30 3/4 1-31 3/4 1-32 6-3

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Cl

SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to at least 144.75 inches.

APPLICABILITY: -

MODES 1 and 2*0.

ACTION:

With a maximum of one shutdown CEA withdrawn to less than 144.75 inches, except for surveillance testing pursuant to Specification 4.1. 3.1. 2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a. Withdraw the CEA to at least 144.75 inches, or

b. Declare the CEA inoperable and uppity Specification 3.1.3.1.

~empty wtpp SURVEILLANCE RE UIREMENTS 4.1; 3.5 Each shutdown CEA shall be determined.to be withdrawn to at least 144.75 inches:

Within 15 minutes prior to withdrawal of any CEAs in regulating groups during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter>> s S,fe>>d-Irl+NQls ooherl ketfl Q8H0'6 csfN. s~crr~le

'Ti e. ingOidAlal CGA CIsssrHo~ cd- icsN.t ones.

pe h&~

"See Special Test Exception 3. 10.2.

¹With K ff greater eff than or equal to 1.

PALO VERDE - UNIT 3 3/4 1-28

'0

~O II

CONTROLLED BY USER REGULATING CEA INSERTION LIMITS L.'MITING CONDITION FOR OPERATION

3. 1.3w6 The regulating CEA groups shall be limited to the withdrawa se-quence, and to the insertion limits¹¹ shown on Figure 3.1-3~" when he COLSS

's in service or shown on Figure 3.1-4"" when the COLSS is not i service.

T e CEA insertion between the 'Long Term Steady State Insertion imits and the Tra sient Insertion Limits is restricted to Less than or equal to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interv

b. Less than or equal to 5 Effective Ful.l Power 0 s per 30 Effective Full Power Day interval, and

c. ss than or equal to 14 Effective Full Po er Days per 18 Effective Fu Power Months.

APPLICABILITY: MODES 1* and 2"¹.

ACTION:

a. Mith the regulating CEA groups inserted beyond the Transient

. insertion Limits, except for surveillance testing .pursuant to Specification 4 1.3. 1.2, wit >n 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the rggulating, CEA groups to"within the limits,.or-"-

2. 'educe THERMAL POMER to less than or equal to that fraction of RATER TNERNAL POREP which is allowed by the CEA group position ia ~

using Figures 3.1f-)~or 3.1-4.

b. Mith the regulating EA groups inserted between the, Long Term Steady State Insertion Li its and the Transient Insertion Limits for inter-vals greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, -opera ion may pro-ceed provided ei er:

1. The Short erm Steady State Insertion Limits of Figure 3.1-3 or Figur 3. 1-4 are not exceeded, or

2. Any. sub equent increase in THERMAL POMER is restricted to less than o equal to 5X of RATED THERMAL POMER per hour.

"See Special Tes Exceptions 3.10.2 and 3.10.4.

.;-.".:; 'ith Keff greater than or equal to 1.

"*CEAs are fu Oy withdrawn in accordance with Fioure 3. -3 or .=igu e 3. I-a when-withdrawn o at least 144.75 inches.

¹¹A reacto power cutback will cause ei her (Case 1) Regulating Group 5~or Regulati~ g Group 4 and 5 to be dropped with no seauential inse. -ion of addi-tional,Regulating Groups (Groups 1, 2, 3, and 4) or (Case 2) Regulating Group,5 or. Regulating Group 4 and 5 to be dropped with all or part of the remaining Regulating Groups (Groups 1, 2, 3, and 4) being sequentially in-serted. In either case, the Transien- Insertion Limit and the withdrawal sequence of Figure 3. 1-3 or Figure 3. 1-4 can be exceeded for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

PALO YEROE - UNIT 3 3/4 1-29

0 0

II

CONTROLLED BY USER ACTION: (Continued)

c. With the regulating CEA groups inserted between the Long Term Steady

, State Insertion Limits and the Transient Insertion Limits for/inter vals greater than 5 EFPD per 30 EFPD interval or greater than 14iEFPO per 18 Effective Full Power Months, either:

1. Restore the regulating groups to within the Long Term Steady State Insertion Limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

2. Be in at least HOT STAHOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REOUIREHENTSN

4. 1.3.6 .The position of each regulating CEA groupishall be determined to be within the Transient Insertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the PDIL Aucti'oneer Alarm Circuit is inoperable, then ver-

'fy the individual CEA positions at~least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The accumulated times during which the regulating CEAigroupsf'are inserted beyond the Long Term Steady State Insertion 'Limits but within the Transient Insertion Limits shall be determined .at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:; /

~ . ~

PALO VERDE - UNIT 3 3/4 1-30

0 0

0

REGULATING CEA INSERTION LIMITS L'MITING CONDITION FOR OPERATION O.I.5.5 'tie. ros(~(a4 vvq EEA o(vis~~ ~hat(( (se.frise 'vi'es(.

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~t The regulating CEA groups shall be imited to the wi drawal se-quence, and to the insertion limits¹¹ n on Figure 3. 1-3 when the COLSS is in service or shown on Figure 3. 1- " when the COLSS is not in service.

The CEA insertion between the Long Term Steady State Insertion Limits and the Transient Insertion Limits is restricted to: s S+ Less than or equal to 5 Effective Full Power Oays per 30 Effective Full Power Day interval, and 4'. ~ Less than or equal to 14 Effective Full Power Days per 18 Effective!

Full Power Months.

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APPLICABILITY: MODES ACTION:

a0 With the regulating CEA groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4. 1.3. 1.2, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the regulating CEA groups to within the limits, or

'THEMAEL( 'poLDEP Q5 fa((oco5.!

~e. of- fror e. CEASE~

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using Figures 3. 1-3 or 3. 1-4 ol-

~) Se if! et& (~sH K6 I W&All5B/ soli'r 4OV ~a

4l 0

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With the regulating CEA'roups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion. Limits for inter-vals greater than 5 EFPO per 30 EFPO interval or greater than 14 EFPD per 18 Effective Full Power Months, either:

1. Restore the regulating groups to within the Long Term Steady State Insertion Limits within 2 hours, or

2. Be in at least HOT STANDBY wi,thin 6 hours.

Q. Mi+4 '[he. r u(a4w C~

~Q ~~er&Ml QP~c+ Q ~&eg+g> 4wliMQyg Tnx~ieut& ~~et+In < Livn.i~

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4. 1.3.6 The position of each regulating CEA group shall be determi ed to be lyh~E within the Transient In ertion Limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ex ept during time intervals when th PDIL Auctioneer Alarm Circuit. is inoperable, then ver-times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shal.l be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

See Special Test Exceptions 3. 10.2 and 3. 10.4.

¹With K ff greater than or equal to 1.

~ ~ ~ ~ a ~ M J ~

¹¹A reactor power cutback will cause either 'C= , -.:..'

Regulating Group 4 and, 5 to be dropped with no tional Regulating Groups (Groups 1 2 roup or Regulating Group 4 and 5 to be remaining Regulating Groups (Groups

'd droppea

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scow Ro with ail r "ar:

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or

.-r

."e serted. 1. 2, 3 :na -') '"

Inn either case, the Transient '.nsert sequence of Figure 3. 1-3 or Figure 3.~-4 cn L;mi t ano .he -'70

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ATTACHMENT 7 - RTD RESPONSE TIME A. DESCRIPTION OF THE, PROPOSED CHANGE The existing PVNGS Unit 3 Technic al Specifications provide an allowance for entering penalty factors into the Core Protection Calculators (CPCs) to compensate for Resistance Temperature Detector (RTD) response times greater than 8 seconds (but less than or equal to 13 seconds). These CPC penalty factors are provided in Technical Specification Table 3.3-2a and are supported by the Cycle 1 safety analyses. However, the Cycle 2 safety analyses will not support these CPC penalty factors. Therefore, Table 3.3-2a must be deleted and Table 3.3-2 must be revised to remove this CPC penalty factor allowance.
B. PURPOSE OF THE TECHNICAL SPECIFICATION Technical Specification Table 3.3-2 (and associated Table 3.3-2a) provide the allowable response times for instrumentation used 'in the PVNGS reactor protective system. By ensuring that the reactor protective instrumentation meets these response time requirements, the assumptions used in the safety analyses are complied with and the associated protective action (ie., reactor trip) is received within the time frame allowed by the safety analyses.
The RTDs that are the subject of this proposed Technical Speci.fication change measure the Reactor Coolant System (RCS) hot and cold leg temperatures. The temperature measurements are provided as an input to the CPCs for use in the DNBR calculation. Each CPC channel receives temperature inputs from both RCS.
hot legs and from two diametrically opposed RCS cold legs.
C NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT This Technical Specificati'on change is necessary in. order to ensure that the Cycle 2 safety analyses assumptions are complied with during Unit 3, Cycle" 2 operations. The Cycle 2 safety analyses assume a maximum RTD response time of 8 seconds and do not include an allowance to enter CPC'enalty factors to compensate for RTD response times greater than 8 seconds. Therefore, there should not be any allowances in the Technical Specifications for using the CPC penalty factors. For this reason, Technical Specification Table 3.3-2a should be deleted and Table 3.3-2 should be revised to remove the penalty factor allowances.
BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability of consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
7-1
0 A discussion of these standards as they relate to the amendment request follows:
Standard 1 -- Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed Technical Specification change will not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed change involves revising Table 3.3-2 and deleting Table 3.3-2a to remove the allowance which provides for CPC'enalty factors to compensate for RTD response times greater than 8 seconds. The subject RTDs measure the RCS hot and cold leg temperatures and provide an input to the associated CPC channel for use in the CPC DNBR calculation. The response times of these RTDs has no impact on the probability of occurrence of any of the accidents that depend on a CPC low DNBR reactor trip.
This revision to Table 3.3-2 and the deletion of Table 3.3-2a will ensure that the consequences of the analyzed accidents will be no worse than evaluated for the Cycle 2 safety analyses. The existing Cycle 1 safety analyses support the use of CPC penalty factors to compensate for RTD response times slower than 8 seconds. The Cycle 2 safety analyses do not support RTD response times greater than 8 seconds. Thus, during Cycle 2, any RTD response times greater than 8 seconds will be unacceptable and the use of Table 3.3-2a will not be supported by the Cycle 2 safety analyses. Therefore, Table 3.3-2a should be deleted and Table 3.3-2 should be revised to assure that operation of PVNGS Unit 1 is in accordance with the Cycle 2 safety analyses.
Standard 2 -- Create the possibility of a new or different kind of accident from any accident previously analyzed.
This proposed Technical Specification change will not create the possibility of a new or different kind of accident from any accident previously analyzed.
This proposed change, to delete the Technical Specification allowance for degraded RTD response times,, does not affect the operation of the RTDs or the associated CPC channels. With the change, if a RTD response time is greater than 8 seconds, the associated CPC channel must be declared inoperable until repairs and/or retest are successfully completed.
Standard 3 -- Involve a significant reduction in a margin of safety.
This proposed Technical Specification change will not involve a significant reduction in a margin of safety. The basis for the existing Technical Specification Table 3.3-2a is the Cycle 1 safety analysis which analyzed the cases where the RTD response times were greater than 8 seconds but less than 13 seconds. For Cycle 2, there will not be an analysis to support the CPC penalty factors for degraded RTD response times. Therefore, Table 3.3-2a must be deleted since it will have no supporting basis during Cycle 2.
The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:
A change that constitutes an additional limitation, restriction or control not presently included in the Technical Specifications: for example, a more stringent surveillance requirement.
7-2
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E. SAFETY EVALUATION FOR THE PROPOSED CHANGE This proposed Technical Specification change will not increase the probability of occurrence of an accident previously evaluated in the FSAR. The subject RTDs measure the RCS hot and cold leg temperatures and .provide an input to the CPCs for use in the CPC DNBR calculations. The response times of these RTDs have no effect
~
on the probability of occurrence of any of the accidents that rely on a CPC low
~
DNBR trip.
This proposed Technical Specification change will not increase the consequences of any accidents previously evaluated in the FSAR. The existing Cycle 1 safety analyses assure a RTD response time of no greater than 8 seconds. Additional analysis was performed for Cycle 1 to justify the application of CPC penalty factors if the measured RTD response times are greater than 8 seconds but no more than 13 seconds. This additional analysis supported the provisions contained in Technical Specification Tables 3.3-2 and 3.3-2a to apply CPC penalty factors to compensate for degraded RTD response times. The Cycle 2 safety analyses also assumed a maximum RTD response time of 8 seconds. However, no additional analysis was performed for Cycle 2 to support RTD response times greater than 8 seconds.
Therefore, the Cycle 2 safety analyses do not support Table 3.3-2a and it must be deleted to ensure operation of PVNGS Unit 1 within the Cycle 2 safety analyses.
Therefore, this Technical Specification change will ensure that the consequences of any accidents will be no greater .than that of the Cycle 2 safety analyses.
This proposed Technical Specification change will not create the possibility of a new or different kind of accident from any accident previously evaluated. This proposed change, to delete the Technical Specifications allowance for degraded RTD response times, does not affect the operation of the RTDs or the associated CPC channels. With the change, if a RTD response time is greater than 8 seconds, associated CPC channel must be declared inoperable until repairs and/or retest are the successfully compl'eted.
Technical Specification change will not reduce the margin of safety as defined
\'his in the basis for any Technical Specifications. The basis for the existing Table 3.3-2a is the Cycle 1, safety analyses which analyzed the cases where the RTD response times were greater than 8 seconds but less than 13 seconds. For Cycle 2, there is no longer an analysis to support the CPC penalty factors for degraded RTD response times. Thus, Table 3.3-2a must be'eleted since it will have no basis during Cycle 2.
ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 3, in accordance with this change, would not:
Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the Staff's testimony to the Atomic Safety and Licensing Board, or
2. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
7-3
li G. MARKED-UP TECHNICAL SPECIFICATION CHANGES PAGES Index:
Limiting Conditions for Operation and Surveillance Requirements:
3/4 3-12 3/4 3-13 7-4
0 0
CONTROLL'=D BY USFR INDEX LIST OF TABLES PAGE FREQUENCY NOTATION ........... 1-8 1.2 OPERATIONAL MODES.............................,..-....... 1-9 2e2 1 REACTOR PROTECTIVE INSTRUMENTATIOH TRIP SETPOINT LIMI L MITS e ~ ~ ~ ~ ~ e o e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o 2-3 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTIOH DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES.....................,
3.1 1 F OR K ff > 0 98o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0oo ~ ~ ~ e ~ ~ ~ ~ 3/4 1-16 3~1 2 FOR Oe98' K ff > Oo97e e ~ ~ ~ ~ ~ ~ o ~ e ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
3/4 1-17 1-18 3o 1 3 FOR 0 97 > K ff > 0 96 r' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~' ~ ~ ~ 3/4
3. 1-4 0. 96 > 3/4 1"19 FOR K ff > Oe95e ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ e ~ ~ ~ e ~ ~ ~ ~ o
3. 1-5 F OR K ff C 0 95 ~ ~ o ~ ~ eoe ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ e ~ ~ e ~ o ~ ~ 00 ~ o ~ ~ ~ oooo ~ 3/4'-20
3. 3" 1 REACTOR PROTECTIVE INSTRUMENTATION...................... 3/4 3-3 3o3 2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES....... 3/4 3-11
~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ ~ ~ e ~ ~ ~ ~ ~ ~ o ~ ~ e ~ o e ~ ~ ~ ~ o ~ ~ ~ ~
4. 3"1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......,...........,......................... 3/4 3"14
3. 3-3 ENGIHEERED SAFETY FEATURES ACTUATION SYSTEM IHSTRUMENTATION..............; ............................ 3/4 3-18
3. 3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES............... 3/4 3-25
3. 3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES................ 3/4 3-28
4. 3-2. ENGINEERED SAFETY FEATURES ACTUATIOH SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............,... 3/4 3-31
3. 3-6 RADIATION MONITORING INSTRUMENTATION.. 3/4 3-38
4. 3-3 RADIATION MONITORING INSTRUMEHTATIOH SURVEILLANCE R'EQUIREMEHTS.......................... 3/4 3-40 3.3 7 SEISMIC MONITORING INSTRUMEHTATIOH.............,........ 3/4 3"43
4. 3-4 SEISMIC MONITORING IHSTRUMEHTATIOH SURVEILLAHCE REQUIREMEHTS. 3/4 3-44
3. 3-8 METEOROLOGICAL MONITORING IHSTRUMEHTATION............... 3/4 3-46
4. 3-5 METEOROLOGICAL MONITORING IHSTRUMENTATIOH SURVEILLANCE REQUIREMENTS 3/4 3"47
3. 3-9A REMOTE SHUTDOWN INSTRUMENTATION 3/4 3-49
3. 3-98 REMOTE SHUTDOWN DISCONNECT SWITCHES.....,.. 3/4 3"50 PALO VERDE - UNIT 3 XX
TABLE 3.3-2 (Continued}
REACTOR PROTECTIVE INSTRUHENTATION RESPONSE TINES FUHCTIOHAL UNIT RESPONSE TINE C. Core Protection Calculator System
1. CEA Calculators Not Applicable
2. Core Protection Calculators Not Applicable O. Supplementary Protection System Pressurizer Pressure - High < 1. 15 second I I. RPS LOGIC A. Matrix Logic Not Applicable
0. Initiation Logic Not Applicable I I I. RPS ACTUATION DEVICES A. Reactor Trip Breakers Not Applicable
0. Hanual Trip Not Applicable Heut.ron detectors are exempt from response time testing. The response time of the neutron flux signal portion of the channel shall be measured from the detector output or from the input of first electronic component in channel.
AA Response time shall be measured from the output of the sensor. Acceptable CEA sensor response shall be demonstrated by compliance with Specification 3. 1.3.4.
OThe pulse transmitters measuring pump speed are exempt from response time testing. The response time shall be measured from the pulse shaper input.
NNResponse t,ime shall be measured from the output of the resistanc temperature detector (sensor). RTO response time shall be measured at least once per 18 months. The measured response t,ime of the slowest RTO shall be less than or equal to seconds. +djtts+meiAs-
-to-the-CPC addressable-constant~given-in-Table-3&--2a-shal-l-be-made-to-aeeommodahe-.
cur rent-values-of-the-RTO-time-constant< . I.fthe-RTO-time-constant-for-a-CPC~hanne+-
-exceeds-the-value-corresponding-t~-the-pena1-ties-curr ently-in-use ,the-affectedmhannet(g-
-==--s ha ) I-be-dec-I a red-inoper ab e-un t i-1-pena Ries-app ropr i ate-to-the-new-timewons tant ill5+8&Mf".
1 INNResponse time shall be measured from the output of the pressure transmitter. The ar e transmitt.er response time shall be less than or equal to O. 7 second.
II 0
CONTROLLED HY USER TABLE 3.3-2a INCREASES IN BERRO BERR2 ANO BERR LAY M S BERRO BERR2 BERR4 RTD DELAY TIME INCREASE INCREASE INCREASE
~X ~X) ~xj t<80sec 0 0 0 8.0 sec < x < 10.0 sec 2. 2.0 1.0 10.0 sec < x < 13.0 sec .0 4.0 6.0 NOTE: BERR term increases are, not cumulative. For example, if the time constant changes fro~the range of 8.0 < x 10.0 sec to the range 10.0 < x < 13.0, tPe BERRO increase from its o 'ginal (x < 8.0 sec}
value is 6.0 not 2.5 + 6.0.
PALO VERDE - UNIT 3 3/4 3-13
0 4l
ATTACHMENT 8 - DNBR LIMIT A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes references to the calculat ed Departure from Nucleate Boiling Ratio (DNBR) from 1.231 to 1.24 as set forth in Technical Specification (T.S.) 2.1.1.1, Table 2.2-1, Basis 2.1.1, and Basis 2.2.1. The amendment also deletes references to the calculation of additional rod bow penaltie's if the rod bow penalty incorporated into the DNBR limit is not sufficient for any part of the cycle.
B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 2.1.1 is to prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission, products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation 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.
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT During Cycle 1 operation, the rod bow penalty factor was applied to the DNBR in increments. This method provided a means for not penalizing the operational margin unnecessarily during the cycle. As the fuel assemblies approach higher burnup the advantage of the Cycle 1 method no longer exists.
The application of a rod bow penalty factor large enough to provide protection throughout the cycle is now more advantageous. This can be accomplished.
because the 'physics of the Cycle 2 core is such that, by applying a rod bow penalty factor of 1.75% Minimum DNBR (MDNBR) to the DNBR limit, there will be sufficient margin to compensate for the effects of rod bow caused by those bundles with burnups of less than 30,000 MWD/MTU. For those bundles with burnups of greater than 30 GWD/MTU, there is sufficient margin from other factors to offset the small increase in the rod bow penalty.
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
8-1
A discussion of these standards as they relate to the amendment request follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident, previously evaluated.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the proposed change incorporates the reference cycle (Cycle 1) approved fuel rod bow penalty factor into the DNBR limit for fuel assembly burnups of up to 30,000 MWD/MTU. For those assemblies which will reach burnups of greater than 30,000 MWD/MTU in Cycle 2, there is sufficient available margin, due to lower radial power peaks, to offset any increase in the rod bow penalty. Thus, the probability or consequences of an accident occurring during Cycle 2 is the same as the reference cycle.
Standard 2--Create the possibility of a new or different kind of accident from any, accident previously evaluated.
The proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change incorporates the reference cycle approved fuel rod bow penalty factor into the DNBR limit for fuel assembly burnups of up to 30,000 MWD/MTU. For those assemblies which will reach burnups of greater than 30,000 MWD/MTU in Cycle 2, there is sufficient available margin, due to lower radial power peaks, to offset any increase in the rod bow penalty. Therefore, the possibility of a new or different kind of accident will not increase.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change does not involve a significant reduction in a margin of safety because the proposed change incorporates the reference cycle approved fuel rod bow penalty factor in the DNBR limit for fuel assembly burnups of up to 30,000 MWD/MTU. For those assemblies which will reach burnups of greater than 30,000 MWD/MTU in Cycle 2, there is sufficient available margin, due to lower radial power peaks, to offset any increase in the rod bow penalty. Therefore, there is no reduction in the margin of safety.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:
(iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.
8-2
E. SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Spec'ification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace any equipment or components important to safety. The proposed change changes the DNBR margin .by incorporating the:
reference cycle approved fuel rod bow penalty for a burnup of up to 30,000 MWD/MTU. Assemblies which will reach a burnup of greater than 30,000 MWD/MTU in Cycle 2, will not contribute a large enough rod bow .penalty to require a larger penalty factor to be applied to the DNBR 1'imit. The reference cycl'e safety analysis has incorporated into the analysis results. The effects of the higher burnups and, therefore, the DNBR for Cycle 2 is bounded by the reference cycle.
'The proposed Technical Specification amendment will: not create the possibility for an accident or .malfunction of a .different type than any previously evaluated in the FSAR. The proposed change is: bounded by the reference cycle safety analysis because the effects of higher burnups on the fuel rod bow, penalty factor were 'incorporated into the,analysi;s.. Therefoxe, the possibility of a new or different kind of accident, stays the same..
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the .basis for the technical specifications. The proposed change is bounded by the reference cycle safety analysi's because the effects of higher burnups on the fuel rod bow penalty factor were incorporated'nto the analysis. Therefore, the margin of .safety stays the same.
~ ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question, because operation of PVNGS Unit 3, in accordance, with this change, would not:
1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified .by the staff's testimony to the Atomic Safety and Licensing Board; or
2. 'Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.,
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G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Safety Limits and Limiting Safety System Settings:
2-1 2-3 2-5 Bases for Safety Limits and Limiting Safety System Settings:
B 2-1 B 2-2' 2'-6 B 3/4 4-1 8-4
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CONTROLLED BY USER 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR 2.1.1.1 The calculated DNBR of the reactor core shall be maintained greater than or equal to (A%
.APPLICABILITY: MODES 1 and 2.
ACTION:
t;vP Whenever the calculated DNBR of the reactgr has decreased to less than~~,
be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specifi-cation 6.7.
PEAK LINEAR HEAT RATE 2.1.1'.2 The peak linear heat rate (adjusted for fuel rod dynamics) of, the fuel shall be maintained less than or equal to 21 kW/ft.
APPLICABILITY:. MODES 1. and 2.
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kW/ft, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7 ~
REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.
APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.
MODES 3, 4, and 5:
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to ~ithin its l.imi t wi thin 5 minutes, and comply wi.th the requirements of Specification 6.7.
PAI 0 VERDE - UNIT 3 2-1
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TABLE 2.2-1 REACTOR PROTECTIVE IHSTRUMEHTATIOH TRIP SETPOIHT LIMITS FUNCTIONAL UNIT TRIP SETPOIHT ALLOWABLE VALUES I. TRIP GENERATION A. Process
l. Pressurizer'ressure - High < 2383 psia < 2388 psia O
2. Pressurizer Pressure " Low > l837 psia (2) > 1822 psia (2) 0
3. Steam Generator Level - Low > 44.2X (4) > 43.7X (4)
Steam Generator Level - High < 91.0X (9) < 91.5X (9) ZJ
5. Steam Generator Pressure - Low > 919 psia (3) > 912 psia (3) 0
6. Containment Pressure " H'.gh < 3.0 psig < 3.2 psig
7. Reactor Coolant Flow - Low rn U
a. Rate < 0.115 psi/sec (6)(7) < 0.118 psi/sec (6)(7) U3
b. Floor > 11.9 psid(6)(7) > 11.7 psid (6)(7) r
c. Band < 10.0 psid(6)(7) < 10.2 psid (6)(7)
(6
8. Local Power Density - High 21. 0. kW/ft (5) ( 21. 0 kM/fL (5)
I >9' /
9. DNOR - Low > k;Rely-(5) (5)
Zl O. Excore Neutron Flux
1. Variable Overpower Trip
a. Rate < 10.6X/min of RATED < 11.OX/min of RATED THERMAL POWER (8) THERMAL POWER (8)
b. Ceiling < llO.OX of RATED < 111.0X of RATED THERMAL POWER (8) THERMAL POWER (8)
c. Band < 9.8X of RATED < 10.0X of RATED THERMAL POWER (8) ~ THERMAL POWER (8)
~ ~
TABLE 2. 2-1 (. (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS Trip may be manually bypassed above, 10-~X of RATED THERMAL POMER; bypass shall be automatically removed when'HERMAL POWER is "less than or equal to 10"~X of RATED THERMAL PQMER.
(2) In MODES 3-4, value may be decreashd manually, to a mRimum of 100 psia, as pressurizer pressure is reducedp=-provided the margin between the pres-surizer pressure and this value is.@maintained at less.~Chan or equal to 400 psi; the setpoint shall be increased automaticallj as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; .~ass sha'll be automatically removed whenever pressurize pressure is gpeater than or equal to 500 psia.
'f (3) In MODES 3-4, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or e'gual to 200 psi; the setpoint shall be increased automatically as.steam generator pressure is increased until the trip setpoint is reached.
(4) Irange of the distance between instrument nozzles.
steam generator upper and lower level wide (5) As stored within the Core Protectioh Calculator (CPC}. Calculation of the trip setpoint includes measurement, calculational and processor uncer-tainties~ Ta ip may be manually bypassed below 10-~X of RATED THERMAL POMER; bypass sha13 be automatically removed when THERMAL POWER is greater than or equal to 1Q=~~ of RATED THERMAL POWER.
T he approved DHBR limit is 1.231 which includes a partial rod,bow penalty compensation. If the Coel burnup exceeds,shall be adjusted~
that for which an i"ncreased rod bow penalty is required, the DHBR limit, In this case a DNBR Hip setpoint,of 1.231 is allowed'provided that the -difference is com-pensated by an increase in the CPC addressable const:ant BERRl as follows:
v'here 8 RR1 ld is the uncoa ensatad value go<BBRR1; RB is t fuel rod bow p nalty in DRBRQB is the fuel ~rd bow penalty iry DilBR already ac ounted for in th&DHBR limit; POL: is the power opera'ting limit; and d (v pOL)/d ( . DRBR) is the absolu value of the epst adverse derivative of POL with respect to DHBR.
PALO VERGE - UNIT 3 2-5 AMEHGsMEHT HO. 2
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GONTROLLED BY USER 2.l and 2.2 SAFETY LIMITS AND LIMITING SAFFTY SYSTFM SETTINGS BASES 2:l. l REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would'esult in the release of fission products to the reactor coolant. Overheating of the .fuel cladding is prevented by (1) restricting fuel operation 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, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kM/ft which will not cause fuel centerline melting in any fuel rod.
ized First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location NB for axially uniform and p non-uni form heat flux di stributi ons. The loca the ratio of the predicted DHB heat flux at
.actual heat flux at that location, is indic ti B rati o (DNBR), defined as p ticular core location to the e of the margin to ONB.
minimum value of DNBR duiing normal opera o and design basis anticipated operational occurrences is limited to The ased upon a statistical combination of CE-1 CHF correlation and engineering fa or uncertainties and is estab1ished s a Safety Limit. The DNBR limit of .
~
~ ~ includes a rod bow corn'pensation of on DNBR. -Fo~uel-burnup~lvichmxceedMha~~
~ ~
DNBR-trip-s i~~equi.red-,Mhe&HBR-l-imi~ ~
ha+1-be-ad jus ted~~+i~cas~4e.
'bow-penal
~
etpei nt-of-~31-~Howedm~h~quired&HB~crease-is-
~
-compensated-by-an-increase-of-the-addr~He-conMan~EHB:
Second, operation with a peak 1inear heat rate below that which wouId cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level (i.e., with some melting in the center},
fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods. Volume changes which accompany the solid to liquid phase change are significant and require accommodation. Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting. Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit. To account for fuel rod dynamics (lags}, the directly indicated linear heat rate is dynamica.lly adjusted by the, CPC program.
(~
PALO VERDE - UNIT 3 B 2-1
45 Il
CONTROLLED BY USER BASES Limiting Safety System Settings for the Low ONBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and -High Linear Power Level trips, and Limiting Conditions for Operation on DNBR and kM/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.
2. 1.2 REACTOR COOLANT SYSTfM PRESSURE 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 Coolant System components are designed to Section III, 1974 Edition, Summer 1975 Addendum, of the ASHE Code for Nuclear Po~er PTant Components which permits a maximum transient pressure of 110K (2750 psia) of 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.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reacto~ Coolant System are prevented from exceeding thei r Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allo~able Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
l~ 2p The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Safety Limits of . nd 21 kM/ft, respectively.
Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the ONBR - Low and Local Power Oensity-High trips include the measulement, calculational and processor uncer ainties and dynamic allowances as defined in CESSAR System 80 applicable system descriptions and safety analyses.
PALO VERDE - UNIT 3 B 2-2
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CONTROLLED BY USER SAFETY LIMITS ANQ LIMITING SAFETY SYSTEMS SETTINGS BASES
,.ONBR - Low (Continued)
The ONBR, the trip variable, calculat by the CPC incorporates various uncer-tainties and dynamic compensation r utines to assure a trip is initiated prior to violation of fuel design limit . These uncertainties and dynamic compensa-tion routines ensure that a rea or trip occurs when the calculated core ONBR is sufficiently greater than such that the decrease in. calculated core ONBR after the trip will not result in a violation of the,ONBR'afety Limit.
CPC uncertainties related to ONBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Oynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
The ONBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.
Parameter Limitin Value
a. RCS Cold Leg Temperature-Low > 470'F
b. . RCS Cold Leg Temperature-High < 610'F
c. Axial Shape Index-Positive Not more positive than + 0.5
d. Axial Shape Index-Negative Not more negative than - 0.5
e. Pressurizer Pressure-Low > 1861 psia
f. Pressurizer Pressure-High < 2388 psia
g. 'Integrated Radial Peaking Factor-Low > 1.28
h. Integrated Radial Peaking Factor-High < 4.28
i. equality Margin-Low > 0 Steam Generator Level " Hi h The Steam Generator Level - High trip is provided to protec the turbine from excessive moisture. carry over. Since the turbine is automatically tripped when the reactor- is,.tripped; this trip provides a reliable means for providing protection to the turbine from excesssive moisture carryover. This trip's setpoint does not correspond to a safety limit, and provides protection in the event of excess feedwater flow.. The setpoint is identical to the main steam isol'ation setpoint. 'ts functional capability at the specified tr'.p setting enhances the overall reliability of he reactor protec-',on system, PALO VERDE -'NIT 3 S'-6
0 0
CONTROLLED BY USER'/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 REACTOR COOLANT, LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above~~
/i'/
during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT .STANDBY wi.thin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
In MODE 3, a single reactor coolant loop provides sufficient:heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal.
capability for removing decay heat; but single failure considerations require
'f that at least two loops (either shutdown cooling or RCS) be OPERABLE. Thus, the reactor coolant loops are not OPERABLE, this specification requires that two shutdown cooling loops be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single shutdown cool.ing loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.
The operation of one reactor coolant pump or one shutdowncooling pump provides adequate flow to ensure mixing, prevent strati.fication, and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 4000 gpm will circ~late one equivalent Reactor Coolant System volume of 12,097 cubic feet in approximately 23 minutes.
The reactivity 'change rate associa'ted with boron reductions will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a reactor coolant pump in MODES 4 and 5, with one or more RCS cold legs less than or equal to 255 F during cooldown or 295 F during heatup are provided to prevent RCS pressure transients, caused by, energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than lOO~F above each of the RCS cold leg temperatures.
3/4.4.,2 SAFETY YALYES The pressurizer code safety valves operat'e to prevent tl e RCS from being pressurized above its Safety Limit of 2750 psia. Each safety, valve is designed to relieve a minimum of 460,000 lb per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the .event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure reiier caoaoility and ~il'I prevent RCS overpressurization.
.PALO VERDE - UNIT 3 8 3/4 4-1
0 il
ATTACHMENT 9 - RCS FLOW RATE A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes the Reactor Coolant System (RCS) total flow rate as set forth ig Technical Specification (T.S.) 3.2.5 from gregter than or equal to 164.0 x 10 ibm/hr to greater than or equal to 155.8 x 10 ibm/hr.
PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.2.5 ensures that the actual RCS total flow rate is maintained at or above the minimum value used in the safety analysis.
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT T.S. 3.2.5 is being changed to eliminate an ambiguity in where instrument uncertainty is to be included when comparing measured RCS flow rate against the RCS flow rate used in the safety analysis. As currently worded, actual total RCS f$ ow rate is to be compared against the 100% design flow value of 164.0 x 10 ibm/hr,. The term "actual" implies that the RCS flow rate determined by the Reactor Coolant Pump (RCP) delta-pressure method is to be corrected for pressure transmitter uncertainty. The uncertainty amounts to a maximum of 4S of flow for transmitters within the)r calibration period. The corrected flow rate i:s then compared to 164.0 x 10 ibm/hr. The RCS flow ratg used in the safety analysis, however, is 95% of the dgsign flow or 155.8 x 10 ibm/hr. The 100% design flow rate of 164.0 x 10 ibm/hr conservatively accommodated the maximum instrument uncertainty of 4S, removing the need to correct for instrument uncertainty. The T.S. basis states that the specification is provided to ensure that the actual total RCS flow rate is maintained at or above the minimum value used in the safety analysis. This T.S. change will remove the -ambiguity and permi't any changes in instrument uncertainty to be handl'ed procedurally rather than requiring additional T.S.
changes.
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission has provi'ded 'tandards for determining whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed'mendment to an operating 1'icense for a facility involves no significant hazards consideration. if- operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences .of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards as they relate to the amendment request follows:
9-1
/
i 0
C 41
Standard 1--Involve a significant increase in the probability or consequences of an accident prev'iously evaluated.
The proposed change does not involve a significant increase in the probability or consequenges of an accident previously evaluated because the value of 155.8 x 10 ibm/hr for minimum RCS flow rate is the value used in the reference cycle (Cycle 1) safety analysis. Therefore, the probability or consequences of an accident is the same for Cycle 2 as it is for the reference cycle.
Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change will not create the possibility of a new or different kind of accident from, any accident previously evaluated because the same value was used for both the reference cycle and Cycle 2 safety analysis.
Therefore there is no possibility of creating, a new or different kind of accident with the reduced RCS total flow.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change -does not involve a significant reduction in the margin of safety because no changes have been made to the safety analysis. The proposed value in the T.S. is the value used in both the reference cycle and Cycle 2 safety analysis. Therefore, the margin of safety is the same for Cycle 2 as i.t is for the reference cycle.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:
(i) A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.
SAFETY EVALUATION FOR THE AMENDMENT RE VEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components. important to safety.
The safety analysis for the proposed change is the same as the reference cycle and, 'therefore, the probability of occurrence or the consequences of an accident is the same.
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0 il
The proposed technical specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The Cycle 2 safety analysis for the proposed change uses the same value for RCS minimum flowrate as for the reference cycle and therefore, the possibility for an accident is the same.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the bases for the technical specifications. No changes have been made to the safety analysis. The proposed value in the T.S.. is the value used in both the reference cycle and 'Cycle 2 safety analysis. Therefore, there is no reduction in the margin of safety.
F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 3, in accordance with this change, would not:
1. Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
~ ~
G. MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions for Operation
~
and Surveillance Requirements:
3/4 2-8 Bases for Limiting Conditions for Operation and Surveillance Requirements:
B 3/4 2-4 9-3
Cl 0
0
CONTROLLED BY USER POMER DISTRIBUTION LIMITS 3/4. 2. 5 RCS FLOW RATE LIHITING CONDITION FOR OPERATION ESSES 3.2.5 The actual Reactor Coolant. System total flow rate shall be greater than or 'equal to .3&~ x 106 ibm/hr.
APPLICABILITY: MODE 1.
ACTION:
Mith the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THERHAL POMER to less than 5X of RATED THEQSL POMER wi,thin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
SURYEILLANCE RE UIREHENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than or equal to its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
PALO VERDE - UNIT 3 3/4 2-8
4l 0
CONTROLLED BY USFR POMER DISTRIBUTION LIMITS BASES 3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the safety analyses.
3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold'eg temperature is maintained withi'n the range of values used in the safety analyses.
3/4.2.7 AXIAL SHAPE INDEX This specification is provided to -ensure that the actual value of the core average AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.
3/4.2.8 PRESSURIZER'RESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within .the range of values used in the safety analyses.
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ATTACHMENT 10 - LINEAR HEAT RATE A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment changes the Linear Heat Rate (LHR) limit as defined in Technical Specification (T.S.) 3/4.2.-1 from 14.0 kw/ft to 13.5 kw/ft. The change also provides information for the appropriate methods of monitoring LHR and formats the T.S. with regard to human factors.
B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3/4.2..1 is to limit Linear Heat Rate which will ensure that, in the event of a Loss of Coolant Accident (LOCA), the peak temperature of the fuel cladding will not exceed 2200 F.
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT In support of the Unit 3 reload, the re-analysis of the Safety Analyses resulted in a change in the Linear Heat Rate limit to ensure the peak fuel clad temperature is not exceeded.
In addition to changing the references to LHR, the amendment also delineates how LHR is to be monitored. By providing more detail of the monitoring of LHR, assurance is provided that, the LHR will be maintained below the specified limit. The amendment also changes the format of the ACTION statement in such a way as to facilitate assessment of the actions required if the limit should be exceeded.
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50..92. A proposed'mendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety..
A di'scussion of these standards as they. relate to the amendment request follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change: does not involve a significant increase in the probability or consequences of .an accident previously evaluated because the safety analysis of the proposed change is bounded by the safety limits set forth by 10CFR50.46. Changing the LHR limit will ensure that there is sufficient margin for the most limiting Design Basis Event (DBE). The format changes to the LCO and Action statements further 10-1
0 define and clarify the actions required to be taken to ensure maintaining the LHR .below the limit. Therefore, there will be no increase in the probabil'ity or consequences of an accident.
Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change will not create the possibility of a new or different kind of accident from any accident .previously evaluated because the safety analysis results of the proposed change are bounded by the safety limits set forth by 10CFR50.46. The format modification changes the presentation of information within the T.S. but does not delete required actions and adds additional restrictions. Therefore, there will be no increase in the possibility of a new or different kind of accident.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change does not involve a significant reduction in a margin of safety because the safety analysis. results of the proposed change are bounded by the safety limits set forth by 10CFR50.46. Changing the LHR limit will maintain sufficient margin for the most limiting DBE.
Therefore, there will be no reduction in the safety margin.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by examples:
(i) A purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error or a change in nomenclature.
and (iii) For a nuclear .power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no significant changes are made to the acceptable criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.
SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The safety analysis results of the proposed change are bounded by the safety limits set forth by 10CFR50.46 and do not change or replace equipment or components which are important to safety.
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The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a di.fferent type than any previously evaluated in the FSAR. The safety analysis results of the proposed change, are bounded by the safety limits set forth by 10CFR50.46. The proposed change to the LHR is more conservative than the LHR allowed by the reference cycle (Cycle 1), thus reducing the consequences, of an event but not creating any new or different accident or malfunction. The format modification changes the presentation of information within the T.S., but does not delete required actions and adds additional restrictions.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the technical specifications. The safety analysis results of the proposed change are bounded by the safety limits set forth by 10CFR50.46. Changing the LHR limit for Cycle 2 will maintain sufficient margin for the most limiting DBE, thus maintaining the margin of safety.
ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 3 in accordance with this change, would not:
Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement .(FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions for Operation and Surveillance Requirements:
3/4 2-1 Bases for Limiting Conditions for Operation and Surveillance Requirements:
B 3/4 2-1 10-3
Cl ll
CONTROLLED BY USER
,,".,".";-'3/4. 2 POWER DISTRIBUTION LIMITS 3/4 2.1 LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION (III,(- Og (2 ..=.
3.2.1 The linear heat rate W/ft&~4Jl G< /VAN(lgfW16d A'lnf .
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APPLICABILITY: MODE 1 above 20K'f RATED THERMAL. POWER.
ACTION:
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minutes initiate corrective action (TtdL(. I"o.
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to reduce the is cvtv(-
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~ea IIMh linear heat rata to within the limits and either:
a. Restore the linear heat rate to within its limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or
b. Reduce THERMAL POWER to less than or equal to 20K of RATED THERMAL .
POWER within the next 6 hours.
SURVEILLANCE REQUIREMENTS
4. 2. 1. 1 The provisions of Specification 4. 0. 4 are. not applicable.
4.2.1.2 The linear heat rate shall be determined to be within its limiQ when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that the linear heat rate, as indicated on mhb OPERABLE Loca'I Power Density channe~tis dj)g tel tt5 l( i'<Li 4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operati'ng 1 ~
ft~ f~~~ ~v PAl 0 VERDE - UNIT 3 3/4 2-'
Cl II 0
-'-. '3/4'. 2 POMER OISTRIBUTION LIMITS BASES 3/4.2. 1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Oensity channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The COLSS performs this function by continuousIy monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate. Reactor operation at or below this calculated power level assures that, the limits of are not exceeded.
~
/3,+
kM/ft The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steady-state opera-tion. Normal reactor power transients or equipment failures which do not require a reactor. trip may result in this core power operating limit being exceeded. In the event this occurs, COLSS alarms will be annunciated. If the event which causes the COLSS limit to be exceeded results in conditions which approach the corp safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation. The COLSS calculation of the linear heat rate includes appropriate penalty factors which provide, with a 95/95 probability/
confidence level, that the maximum linear heat rate calculated. by COLSS is conservative with respect to the actual maximum linear heat rate existing in the core. These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, engineering heat flux uncertainty, axial densification, software algorithm modelling, computer processing, rod bow, and core power measurement.
Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB, and total core power are also monitored by the CPCs by utilizing PC~4p channel .
factors plus those associated with the The a ove CPC tP~
Therefore, in the event that the
~ I listed uncertainty and penal'ty COLSS startup test acceptance criteria r
is are also inc uded in the CPCs.
[ i ylC~p.t Ada~ J5LP PALO VEROE - UNIT 3 B 3/4 2-1
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ATTACHMENT .11 - DNBR MARGIN MONITORING DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment will revise Technical Specifications (T.S) 3/4.2.4, Table 3.3-1, Bases 3/4.1.3 and Bases 3/4.2.4. The changes are as follows:
T.S. 3.2.4 - (1) Replaces the T.S. with a new format which addresses the specific conditions for monitoring DNBR with or without COLSS and/or the CEACs, (2) delineates by a new format what ACTIONS should be taken, (3) removes reference to the DNBR Penalty Factor table used in T.S. 4.2'.4.4,, and (4) replaces the current Figures 3.2-1 and 3.2-2 for the DNBR limits with Figures 3.2-1, 3.2-2 and '3.2-2a addressing DNBR operating limits for the conditions mentioned in (1) above.
T.S. 3.3 (1) Removes references to the operation of the reactor with both CEACs inoperable and with or without COLSS in service, and (2) deletes the graph of DNBR margin operating limit (Figure 3.3-1) based on COLSS for both CEACs inoperable. This is a result of this information being incorporated into the proposed T.S. 3/4.2.4.
Bases 3/4.1.3 - (1) Removes references to Cycle 1 specific information, and (2) modifies Bases due to T.S. 3/4.2.4 changes.
Bases 3.2.4 - Modifies Bases due to the T.S. 3/4.2.4 changes.
These changes are due, in part, to ensuring operation of Cycle 2 within the approved safety analysis and to improving the Technical Specifications from a human factors point of view.
The difference between the Unit 3 Cycle 2 changes and the Unit 2 Cycle 2 changes is that this submittal adds a specific reference to the full and part-length CEA T.S. (3/4.1.3.5, 3/4.1.3.6, and 3/4.1.3.7) to Table 3.3-1.
This reference will also be incorporated into the Unit 1 and Unit 2 T.S.
PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3/4.2.4 is to ensure the limitation of DNBR, as a function of AXIAL SHAPE INDEX, will be within the conservative envelope of operating conditions consistent with the safety analysis assumptions and which have been analytically demonstrated adequate to maintain an acceptable minimum DNBR throughout all anticipated operational occurrences. Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.
The purpose of T.S. 3/4.3.1 (Table 3.3-1) is to ensure that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, redundancy is maintained to permit a channel to be out of service (3)'ufficient for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
11-1
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT The proposed changes are due to (1) ensuring operation of the reactor within approved safety analysis for Cycle 2 by modifying the T.S. graphs, (2) increasing operator reliabi.lity by placing DNBR operating limits in one place, and (3) eliminating superfluous information to reduce confusion and the possibility of misuse. (i.e., eliminating the Table in T AS. 4.2.4.4)
BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards as they relate to the amendment request follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to the graphs of T.S. 3/4.2.4 does not involve a significant increase in the probability or consequences of an accident previously evaluated because the Cycle 2 safety analyses have shown that when COLSS is in service and at least one CEAC is operable, Specification 3.2.4a provides adequate margin to DNB to accommodate the limiting Anticipated Operational Occurrence (AOO) without violating the Specified Acceptable Fuel Design Limits (SAFDL). For the case when neither CEAC is operable but COLSS is in service, the CPCs assume a preset CEA configuration because they can not obtain the required CEA position information to ensure that the SAFDL or DNBR will not be violated during an AOO. Thus, as a result of the reevaluation of the limiting AOOs for Cycle 2, Specification 3.2.4.b requires that core Power Operating Limit (POL - as calculated by COLSS) be reduced as shown on Figure 3.2-1. This ensures the limiting AOO will not result in a violation of SAFDLs. The proposed revision to Figure 3.2-2 .accounts for the situation when COLSS is out-of-service but at least one CEAC is operable. In this case, the Cycle 2 safety analysis has shown that, by maintaining the CPC calculated DNBR above the value shown in the figure, the limiting AOO will not result in a violation of the SAFDLs. When COLSS is out of service and both CEACs are inoperable, there must be additional margin in the CPC DNBR value to ensure that the limiting AOO will not result in exceeding a SAFDL. An evaluation of the Cycle 2 core design has shown that by maintaining the CPC calculated DNBR above the limits shown in the proposed Figure 3.2-2a, the SAFDLs will not be exceeded during the most limiting AOO. Therefore, the proposed change will not significantly increase the probability or consequences of any accident previously evaluated.
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II
The proposed change to the format of T.S. 3/4.2.4 and Table 3.3-1 does not involve a significant increase in the probability or consequences of an accident previously evaluated because consolidation of the DNBR operating limits within one Technical Specification will increase the operator's ability to ensure proper operation of the reactor. The proposed format change still contains the same Limiting Conditions for Operations (LCO), ACTIONS and surveillance requirements as the original Technical Specifications. Therefore, the change will not significantly increase the probability'r consequences of any accident previously evaluated.
The proposed change to eliminate the DNBR penalty factors table of T.S.
4.2.4.4 does not involve a significant increase in the probability or consequences of an accident previously evaluated because the penalty is an allowance for rod bow and has been incorporated into the DNBR value for Cycle 2. This can be done because the burnup of the reactor core in Cycle 2 will reach the value for applying the maximum rod bow penalty and the table will no longer be needed (see Attachment 12). Therefore, the change will not significantly increase the probability or consequences of any accident previously evaluated.
Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change to the graphs of T.S. 3/4.2.4 will not create the possibility of a new or different kind of accident from any accident previously evaluated because they are required to make the graphs consistent with the Cycle 2 safety analyses. Operation of the reactor within the limits as shown on the proposed Figures 3.2-1, 3.2-2 and 3.2-2a will ensure that the SAFDLs will not be exceeded during the most limiting AOO. Since the Cycle 2 figures were based on meeting the same criteria as the Cycle 1 figures, the possibility of a new or different kind of accident from any accident previously evaluated will not be created.
The proposed change to the format of T.S. 3/4.2.4 and Tabl'e 3.3-1 will not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change reduces the possibility of human error by consolidating closely related allowable operations into a single specification and clearly identifying each allowable operation. The contents of the proposed T.S. are the same as those of T.S. 3/4.2.4 and Table 3.3-1, thus the only change is in regard to the human factors element. By keeping the same contents but arranging them so as to reduce human error, the proposed change will not create the possibility of a new or different kind of accident not previously evaluated.
The proposed change to eliminate the DNBR penalty factors table of T.S.
4.2.4.4 will not create the possibility of a new or different kind of accident from any accident previously evaluated because the possibility of misusing the table is eliminated.
11-3
Standard 3--Involve .a significant reduction in a margin of safety..
The proposed change to Figures 3.2-1, 3 '-2 and 3.2-2a does not involve a significant reduction in a margin of safety because the change is required'o make the T.S. consistent with the 'Cycle,2 safety analyses.
Operation of the reactor .within the limits as. shown on the proposed Figures 3.2-1, 3.2-2 and 3.2-2a will ensure that the SAFDLs will not be exceeded during the most limiting AOO. The Cycle 2 figures were based on the same design criteria as the Cycle 1 figures. Therefore, the proposed change does not reduce the margin of safety.
The proposed format change to T.S. 3/4.2.4 and Table 3 '-1 does not involve a significant reduction in a. margin of safety because the contents of the Technical Specifications have remained the same, only a rearrangement of information has taken place. Therefore, the proposed change does not reduce the margin of safety.
The proposed change to eliminate the DNBR penalty factors table of T.S.
4.2.4.4 does not involve a significant reduction in a margin of safety because the maximum rod bow penalty factor has been applied to the DNBR value for Cycle 2'nd, therefore, the table is no longer needed and the margin of safety has been maintained for Cycle 2.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists "(51FR7751) by examples:
A purely .administrative change,to Technical Specification: for example, a change to achieve consistency throughout the Technical Specifications, in correction of an error, or a change in nomenclature.
and (iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved.. This assumes that no significant changes are made to the acceptable
.criteria for the Technical Specifications, the analytical methods used to demonstrate conformance with the Technical Specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable.
SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed change to the graphs of T.S. 3/4.2.4 does not involve a significant increase in the probability or consequences of an accident previously evaluated because the Cycle 2 safety analyses have shown that when COLSS is in service and at least one CEAC is operable, Specification 3.2.4a provides adequate margin to DNB to accommodate the limiting Anticipated Operational Occurrence (AOO) without violating the Specified Acceptable Fuel Design Limits (SAFDL). For the case when neither CEAC is operable but COLSS is in service, the CPCs assume a preset CEA configuration because they can not obtain the required CEA position
4l 0
information to ensure that the SAFDL or DNBR will not be violated during an AOO. Thus, as a result of the reevaluation of the limiting AOOs for Cycle 2, Specification 3.2.4.b requires that core Power Operating Limit (POL - as calculated by COLSS) be reduced as shown on Figure 3.2-1. This ensures the limiting AOO will not result in a violation of SAFDLs. The proposed revision to Figure 3.2-2 accounts for the situation when COLSS is out-of-service but at least one CEAC is operable. In this case, the Cycle 2 safety analysis has shown that, by maintaining the CPC calculated DNBR above the value shown in the figure, the limiting AOO will not result in a violation of the SAFDLs ~ When COLSS is out of service and both CEACs are inoperable, there must be additional margin in the CPC DNBR value to ensure that the limiting AOO will not result in exceeding a SAFDL. An evaluation of the Cycle 2 core design has shown that by maintaining the CPC calculated DNBR above the limits shown in the proposed Figure 3.2-2a, the SAFDLs will not be exceeded during the most limiting AOO. The proposed change to the format of T.S. 3/4.2.4,and Table 3.3-1 consolidates the DNBR operating limits within one Technical Specification and will increase the operator's ability to ensure proper operation of the reactor. The proposed format change still contains the same Limiting Conditions for Operations (LCO), ACTIONS and surveillance requirements as the original Technical Specifications. The proposed change to eliminate the DNBR penalty factors table of T.S. 4.2.4,.4 does not involve a significant increase in the probability or consequences of an accident
.previously evaluated because the penalty is an allowance for rod bow and has been incorporated into the DNBR value for Cycle 2. This can be done because the burnup of the reactor core in Cycle 2 will reach the value for applying the maximum rod bow penalty and the table will no longer be needed (see 2). Therefore, these changes will not significantly increase the probability or consequences of any accident previously evaluated.
The proposed change to the graphs of T.S. 3/4.2.4 will not create the possibility of a new or different kind of accident from any accident previously evaluated because they are required to make the graphs consistent .with the Cycle 2 safety, analyses. Operation of the reactor within the limits as shown on the proposed Figures 3.2-1, 3.2-2 and 3.2-2a will ensure that the SAFDLs will not be exceeded during the most limiting AOO. The Cycle 2 figures were based on meeting the same criteria as the Cycle 1 figures. The proposed change to the format of T.S. 3/4.2.4 and Table 3.3-1 reduce the possibility of human error by consolidating closely related allowable operations into a single specification and clearly identifying each allowable operation. The contents of the proposed T.S. are the same as those of T.S. 3/4.2.4 and Table 3.3-1, thus the only change is in rearranging them so as to reduce human error. The proposed change to eliminate the DNBR penalty factors table of T.S. 4.2.4.4 eliminates the possibility of misusing the table. Therefore the possibility of a new or different kind of accident will not be created.
The proposed change to Figures 3.2-1, 3.2-2 and 3.2-2a does not involve a significant reduction in a margin of safety because the change is required to make the T.S. consistent with the Cycle 2 safety analyses. Operation of the reactor within the limits as shown on the proposed'igures 3.2-1,, 3.2-2 and 3.2-2a will ensure that the SAFDLs will not be exceeded during the most limiting AOO. The Cycle 2 figures were based on the same design criteria as the Cycle 1 figures. The proposed format change to T.S. 3/4.2.4 and Table 3.3-1 does not involve a significant reduction in a margin of safety because 11-5
0 0
0
the contents of the Technical Specifications have remained the same, only a rearrangement of information has taken place. The proposed change to eliminat the DNBR penalty factors table of T.S. 4.2.4.4 does not involve a significant reduction in a margin of safety because the maximum rod bow penalty factor has been applied to the DNBR value for Cycle 2 and, 'therefore, the table is no longer needed and the margin of safety has been maintained for Cycle 2.
F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 3, in accordance with this change, would not:
1'. Result in a significant increase in any adverse environmental impact previously evaluated in .the Final. Environmental Statement (FES) as modified by the staff's testimony,to the Atomi:c .Safety 'and Licensing Board; or
2. Result in a significant change in effluents or power level's; or 3,. Result in .matters not previously reviewed in the licensing basis for PVNGS which may have a si'gnificant environmental impact.
'G. 'MARKED-UP TECHNICAL 'SPECIFICATION CHANGE PAGES Index:
XIX Limiting Conditions for Operation and Surveillance Requirements:
3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-,7a 3/4 3-7 3/4 3-8 3/4 3-9 3/4 3-10 Bases for Limiting Conditions for Operation and Surveillance Requirements:
B 3/4 1-6 B 3/4 1-7 B 3/4 2-3 11-6
CONTROLLED BY USER
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LIST OF FIGURES PAGE
3. 1-lA SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE............ 3/4 1-2a
~ 3. 1"1 ALLOWABLE MTC MODES 1 AHD 2............................ 3/4 1-5
3. 1" 2 MINIMUM BORATED WATER VOLUMES.......................... 3/4 1-12
3. 1-2A PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER....... 3/4 1-23
3. 1-28 CORE POWER LIMIT AFTER 'CEA DEVIATION........ 3/4 1-24 3o 1 3 CEA INSERTION 'LIMITS VS THERMAL POWER ohio
.,(COLSS IN SERVICE)..................................... 3/4 1-31
3. 1-4 CEA INSERTION LIMITS VS THERMAL POWER (COLSS OUT OF SERVICE).....................-.----.----.- 3/4 1-32 hLLo~aecE Fow ~TH 3.2-1 ~r DHBR ~au) W WAR&N OPERATING LI IT LB@ ~t<
1.0 pCi/GRAN DOSE EQUIVALENT I"131..... 3/4 4"27
3. 4" 2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS OF FULL POWER 0 PERATIONo ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 4-29
4. 7-1 SAMPLING PLAN FOR SNUBBER FUHCTIONAL TEST...........'... 3/4 7-26 e 3/4.4-1 HIL"DUCTILITYTRAHSITIOH TEMPERATURE INCREASE AS A FUNCTION OF FAST (E > 1 MeV} NEUTRON FLUENCE (550~F IRRADIATION} . B 3/4 4-10
5. 1" 1 SITE AHD EXCLUSION BOUHDARIES 5-2
5. 1-2 LOW POPULATION ZONE 5" 3
5. 1" 3 GASEOUS RELEASE POINTS 5-4
6. 2" 1 OFFSITE ORGANIZATION. 6" 3
6. 2" 2 OHSITE UNIT ORGAtiIZATION.. 6-4 PALO VERDE - UNIT 3 XIX ANEHOMEt(T HO. 2
0 0
CONTROLLED BY USER POWER DISTRIBUTION LIMITS 3/4. 2.4 ONBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The ONBR margin shall be maintained by operating withi the Region of Acc ptable Operation of Figure 3.2-1 or 3.2-2, as applicable or in accordance with the requirements of Action 6 of Table 3.3-1.
APPLICABILITY: MODE 1 above 20K of RATED THERMAL POWER.
~
Mith operation outside of the region of acceptable op ation, as indicated by
~ ~
either (1) tAe COLSS calculated core power exceeding he COL'SS calculated core power operating limit based on ONBR; or (2) when th COLSS is not, heing. used, any OPERABLE Lo ONBR channel below the ONBR limit within 15 minutes initiate corrective action to restore either the ONBR cor power operating limit or the ONBR to within he limits and either:
a. Restore the ONBR core power operating limit or DNBR to within its limits withi 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or
b. Reduce THERMAL OWER to less than or equal to 20K of 'RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREMENTS 4.2.4.1 The provisions of Spec>(ication 4.0.4 are not applicab The DNBR shall be determin d to be within its limits when THERMAL le'.2.4.2 POWER is above 20K of RATED THERMAL OWER by continuously monitoring the core power distribution with the ore Operating Limit Supervisory System (COLSS) or, with the COLSS out o service, by verifying at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that the DNBR margin, a indicated on all OPERABLE DNBR margin channels, is within the limit hown on Figure 3; 2-2.
4.2.4.3 At least once per 3 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POW R level less than or equal to the core power operating limit based on D BR.
4. 2. 4.4 The following 0 BR or equivalent penalt factors shall be- verified to be included in the COLS and CPC ONBR calculations at least once per 31 EFPD.
(GWD)
DNBR. Penalt (X)"
0-10 0 5 10-20. 1.0 20"30 30" 40 40-50 2.0i 3.5 5.5 i
The penalty for each batch will be determined from, the batch's maximum burnup assembly and applied to the batch's maximum radial power peak. assembly. A single get penalty for COLSS and CPC will be determined from the penalties associated with each batch accounting for the offsetting margins due to the loweriradial power peaks in the higher burnup batches.
PALO VERDE " UNIT 3 3/4 2-5
CONTROLLED BY USER POWER DISTRIBUTION LIMITS ll'Bii 3/4. 2. 4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by one of the following methods:
a. Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR (when COLSS is in service, and either one or both CEACs are operable); or
b. Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR decreased by the allowance shown in Figure 3.2-1 (when COLSS is in service and neither CEAC is operable); or
c. Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC'hannel (when COLSS is out of service and either one or both CEACs are operable); or
d. Operating within the region of acceptable operation of Figure 3.2-2A using any operable CPC channel (when COLSS is out of service and neither CEAC is operable).
APPLICABILITY: MODE 1 above 20K of RATED THERMAL POMER.
ACTION:
Nith the DNBR not being maintained:
l. As indicated by COLSS calculated core power exceeding the appropriate COLSS calculated power operating limit; or
2. Mith COLSS out of service, operation outside the region of acceptable operation of Figure 3.2-2 or 3.2-2A, as applicable; within 15 minutes initiate corrective action to increase the ONBR to within the limits and either:
a. Restore the DNBR to within its limits within 1 hour, or
b. Reduce THERMAL POMER to less than or equal to 20M of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE RE UIREHENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicab1e.
4.2.4.2 The ONBR shall be determined to be within its limits when THERHAL POWER is above 20K of RATED THERHAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that the ONBR, as indicated on any OPERABLE ONBR channel, is within the limit shown on Figure 3.2-2 or Figure 3.2-2A.
4.2.4.3 At least once per 31 days, the COLSS Hargin Alarm 'shall be verified to actuate at a THERHAL POMER level less than or equal to the core power operating limit based on ONBR.
PALO VERDE UNIT P g 3/4 2-S
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CONTROLLED BY USER 100 REGION OF 0 ACCEPTABLE OP ERATlON 80 U
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<.o REGlON OF LQ K 40 UNACCEPTABLE tQ 0 ~ OP ERATlON 08 <
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20 40 60 80 100 PERCENT OF RATED THERMAL POWE fIGU RE '3. 2- 1 DHBR HARGIH OPERATIttG LItlIT BASED Oti COLSS (COLSS IH SERVICE)
PALO VERDE - UHIT 3 3/4 2-6
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CONTROLLED BY USER
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CD CL CC CD 48 I CZ LI CD I
Q CV COLSS ONBR POMER OPERATING LIHIT REOUCTIOU
(% QF RATED THERMAL POMER)
FIGURE 3. 2" 1 COLSS DNBR POWER OPERATIHG LIMIT ALLOMAHCE FOR BOTH CEACs INOPERABLE PALO YEROE - UNIT,A 3/4 2-6 ~B BHBR-He .
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CONTRot LED BY USER COl SS OUT OF SERVICE QNBR l IyIT l INE 2.4 ACCEPTABLE OP ERATION 2.3 CEACs INOPERABLE (.2.2.38).
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m o 22 z
(-.2,2.13)
UNACCEPTABLE QP E RAT ION 2.8
-8.3 -8.2 "8.1 8.1 8.2 CORE AVERAGE ASI FIGURE 3.2-2A
'OH CORE PROTECTION CALCULATORS DHBR MARGIH OPERATIHG LIMIT BASED (COLSS OUT .OF SERVICE, CEACs IHOPERABLE)
PALO VERDE - OMIT g 3/4 2-7a
'0 CP Cl P
CONTROLLED BY USER TABL'E 3. 3-1 (Continued)
REACTOR PROTECTIYE INSTRUMENTATION ACTION'TATEMENTS
3. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator Level 1-L'ow (ESF Steam Generator Level 2-Low (ESF
4. Steam. Generator Level - Low Steam Generator Level - Low (RP>>
(Mide Range) Steam Generator Level 1-Low (ESF Steam Generator Level 2-,Low (ESF
5. Core Protection Calculator Local Power Density - High (RPS)
'ONBR. " Low (RPS)
STARTUP and/or POMER OPERATION'ay continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP and/or POMER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of .ACTION 2 are satisfied; ACTION 4 With the number of channels OPERABLE one. less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5 Mith the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, STARTUP and/or POMER OPERATION may continue provided the reactor trip breaker of the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, the trip breaker associated with the inoperable channel may be closed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing per Specification 4.3. 1. 1. &8 I'e>>lutfErA n45 +
~p? C=14~4lap) 4.l 5, l, AI Q le%
1 ACTION 6 a. Mith .one CEAC inoperable operation may continue for up to 7 days provided that After 7 days, operation continue provided that the conditions of Action Item may 6.b ~~ are met.
b. With both CEACs inoperable , operation may continue provided that:
1. Within 1 hour:
a Operatic is restricted to the linits sheers in Figure~ . 3-1. The DNBR margin require by Specification 3.2.4 is replaced by th'>s restriction when both CEAC's are i operaole nd COLSS is i operation.
b) The Linear eat Rate Margin r uirea by Specifica on 3.2. 1 is main ined.
c) The Re tor Power Cutback ystem is plac d out of service.
PALO VEROE - UNIT 3 3/4 ~-7
CONT. ROLLED BY USER ~i.%i'vl We
'Isnd,'tw cf SPecdgicdd*ddnS TABLE 3. 3-1 (Continued)
'EACTOR PROTECTIYE INSTRUMENTATION Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the DNBR required by Specification 3.2.4b (COLSS in service) or
2. Wi thin 4 a) margin hours:
dd
~++&cut
~~d ACTION STATEMENTS All full-length and part-length CEA groups~
"pos-Ww~ except during survei I lance testing pursuant to the requirements of Specifica-tion 4. 1.3.1.2g@r may-be insertNVNo further than 127.5 inches
/77VNF $ ics 3.2.4d (COLSS out of service) wi thdrawn. >~H~o 3 I.'5 6 h etio~> cER @~tjf 5 2
b) The "RSPT/CEAC Inoperable addressable constant Reactor Power Cutback System in the CPCs is set to indicate that both CEACis , >,g are inoperable ~Pcs-fgsa-~~~9,ld 3 ( $
is disabled, and c) The Control Element Drive Mechanism Control m 9 5:7
~
System (CEOMCS) is placed in and subseque ly maintained in the "Standby" mode enduring CEA~~'motion permi tted excel by~-ebevW when the CEDMCS may be operated in either the "Manual Group" or "Manual Individual" mode.
KA P <~on <urvd.( Qnr ft
~old rcgu) Ferifltrt) $ ~dddl~ d ~ ' m ddMAI i d d dd d- d-d
+CI Ad- othio r) 5 during survei'lienee testin ursuant to Specification
.2 r dur>ng >nsertigo of SEA croup 8 as
'k. l . 3, Cu ~~
~ permi tted by 2. a) above .. then verify at least once LI. I per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the inserted CEAs are aligned within 6.6 inches (indicated position) of all other CEAs in its rou .
4. Foll ing a CEA misalignm t with both CEAC's i perable and COLSS in peration, operation may ontinue provided th within 1 hour:
The power is redu d to 85K of the pr isaligned power but need t be reduced to le~ than 50" of RATFD THERMA OWER. This power restriction replaces the power r striction of Speci+cation 3.1.3.1, Figure 3. -28,'therwise Specrification 3.1.3.1 re ins applic e.
C. .With bot CEACs inoperable and COLSS out"of-serv',
operat' may continue pryrided that:
1. ithin 1 hour:
a) The existipg CPC value of th~PC addressable constan+8ERR1" is multiplied by 1.19 and in resultjog value is re-en red into the CPC b) The eactor Power Cut ck System is pl d out c) o service The COLSS out of ervice Limit ure 3.2-2 of S cification 3.
', on Fig-
, is not appli-cable to this mode of operation.
PALO VERDE - UNIT 3 3/4 3-8
0 CONTROLLED BY USER TABLE 3.3 1 Continued)
REACTOR PROTECTIVE INSTRUMENTATION ACTION STATEMENTS
2. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:
Al.l
/
full )ength and papt length CEA +oops are withdrawn to and subse'quently maintained at the "Full OFjt" positianr except duringsurveil.lance testing pursuant td the requirements of Specifi cation 4.1.3.1.2/or for control'hen CEA grou 5 may be inserted~no further t an 127.5 inche withdrawn.
b) The "RSPT/CEAC Inoperab e" addressabl constant in the CPCs is set to ndicate that oth CEAC's are inoperable.
c) The. ontrol Element Drive Mec nism Control Sy em (CEDMCS) is placed i and subsequent y m intained in he "Standby'ode except dUring EA group 5 otion permi ed by a) above, when the CEDMCS ay be oper ed in either he "Manual
3. 't length Group" or "Manual In 'vidual" mode.
least o ce CE s per 4 ho s, all full 1 ngth and part are verif ed fully with awn except
. during rvei 1-lance esting pursuant to. Specifi tion 4 .3. 1.2 or ur ing inserti n of CEA gro 5 as
~
perm'ed by 2 ' above, then erify at lea once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months tha the inserte CEAs are alj ned within 6 inches (i dicated posit. on) of all Cher CEAs in
'ts group.
4. Following a CEA misali ment with oth CEAC's an COLSS inoperab e, operatio may contin provided tha within 1 hour.
The power is redu ed to 85% f the pre-mis ligned pow r bu need not be/educed to ess than 50% of RATER T RIIAL PP ER. This power restri tion repla'ce the power restriction f Specific ion 3.1.3.1,/Figure 3.1 2B,
~otherwise Speci.fication 3. 1.3. 1 remains applica le, ACTION 7 With three or more auto restarts, excluding periodic auto restarts (Code 30 and Code 33), of one non-bypassea calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST wi hin the next, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
s ACTION 8 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore an inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open an affected reactor trip breaker within the next hour.
PALO VERDE - UNIT 3 3/4 3-9
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CONTROLLEO BY USER REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES Continued and load maneuvering. Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analyses CEA insertions are determined and a consistent set of radial peaking factors defined. The Long Term Steady State and Short Term Insertion Limits are deter-mined based upon the .assumed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used. The limts speci-serve to limit the behavior of. the radial peaking factors within the bounds
-'ied determined from analysis. The actions specified serve to limit the extent of radial xenon redistribution effects to those accommodated in the analyses. The Long and Short Term Insertion Limits of Specification 3.1.3.6 are .specified for the plant which has been designed for primarily base loaded operation but which has the ability to accommodate a limited amount of load maneuvering.
The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion Limits'f Specification 3. 1. 3. 5 ensure that (1) the minimum SHUT-DOWN'ARGIN is maintained, and (2) the potential effects of a CEA exsection accident are limited to acceptable levels. Long"term operation at the Tran-sient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to deter-mine the behavior of the radial peaking factors.
The PVNGS CPC and COLSS systems are responsible for the safety and monitoring functions, respectively, of the reactor core. COLSS monitors the DHB Power Operating Limit (POL) and various operating parameters to help the operator main" tain plant operation within the limiting conditions for operation (LCO). Operat-ing within the LCO guarantees that in .the event of an Anticipated Operational Occurrence (AOO), the CPCs will provide a reactor trip in time to prevent un-acceptable fuel damage.
The COLSS rese es the Loss'of Flow (LOF) transient/'Lll+c
~ ~op@~Qovl Required Overpower Margin (ROPM) to account for
'When the COLSS is Out of Service (COOS), the monitoring function is performed
~
via the CPC calculation of DNBR in conjunction with~Technical Specification COOS Limit Line<>(Figure>s3.2"g) which restricts the reactor power sufficiently to preserve the ROPM.
The reduction of the CEA deviation penal ies in accordance with the CEAC (Control Element Assembly Calculator) sensitivity reduction program has been performed. This task involved setting many of the inward single CEA Ceviatioll penalty factors to 1.0. An inward CEA deviation event in e'feet ovid rot be accompanied by the application of he C=A devia ion penalty in ei:her the CPC .
ONB and LHR (Linear Heat Rate) CalCulatiOnS fOr hOSe CEAS ~ith he reauCea penalty factors. The protection ,or an inwara CEA cev.iation event is hus accounted for separately.
PALO VEROE - UNIT 3
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0 CONTROLLED BY USER POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TII T T (Continued) ptllt/Puntllt is the ratio of the power at a core location in the presence of a tilt to the power at that location with no tilt.
The AZIMUTHAL POWER TILT allowance used in the CPCs is defined as the value of CPC addressable constant TR-1.0.
3/4.2.4 ONBR MARGIN The limitation on DNBR as a function'f AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analy-sis assumptions aed which have been analytically demonstrated adequate to main-tain an acceptable minimum ONBR throughout all anticipated operational occur-of the core with a ONBR at or above this limit provides assurance that an accept-able minimum ONBR will be maintained in the event of a loss of flow transient.
Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power
(- distribution and are capable of verifying that the DNBR does not violate its limits. The COLSS performs this function by .continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR. ~ctm~
Weve-'I-assures-thethe-1-Bii-'ts-of Recure-3 . The COLSS calculation of core power operating limit based on ONBR includes appropriate penalty factors which provide, with a 95/95 probability/confidence level, that the core power limits calculated by COLSS (based on the minimum DNBR Limit) ~are.
conservative with respect to the actual core power limit. These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, engineering heat flux, state parameter measurement, software algorithm modelling, computer processing, rod bow, and core power measurement.
Parameters required to maintain the margin to ONB a total core power are also monitored by the CPCs. Therefore, in the even that the COLSS is not being used, operation within the limits of Figur83. 2-2 can be maintained by utilizing a predetermined DNBR as .a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPCs which assume a minimum core power of 20K of RATED THERMAL POWER. The 20Ã RATED THERMAL POWER threshold is due to the neutron flux detector system being 'ccurate below 20~ core power. Core noise level at low power is too lar e to obtain usable detector readings.
l Mi~(pi 5'+he ONBR penalty factory gg QX5~ gases- to accommodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly. Fuel tcu4Hon'~ assemblies that incur higher average burnup will experience a greater magnitude f rod bow. Conversely, lower burnup assemblies will experience less rod bow.
e penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum int.e-Zia G4Q~ grated planar-radial power peak. A single net penalty for COLSS and CpC is then
~ht d(6lth5 determined from the penalties associated with each batch, account;ing for the off-setting margins due to the lower radial power peaks in the higher burnup batches.
PALO VERDE - UNIT 3 'i ih
~ I iO
ATTACHMENT 12 - AZIMUTHAL TILT A. DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment expands the operating limits o'f az imuthal tilt with COLSS in service. The azimuthal tilt limits will be a step function of power with the upper limit of 0.20 at 20$ power, stepping down to 0.15 at 308 power and 0.10 at 40% power, where it remains to 1008 power.
B. PURPOSE OF THE TECHNICAL SPECIFICATION The limitations on the azimuthal power margins are maintained.
tilt are to ensure that design safety C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT During a reactor power cutback event in Unit 1 the plant was unable to go above 208 power because the azimuthal tilt limit would have been exceeded.
They were required to remain below 208 power for approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months until xenon burned out. This delay could have been prevented and the azimuthal tilt corrected if the plant had been allowed to increase power. This would cause the xenon to burn out faster thus restoring the plant within the limits sooner. By implementing the proposed change such delays could be avoided.
D ~ BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility involves. no significa'nt hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) involve a significant
-'increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
A discussion of these standards as they relate to the amendment request follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because a reevaluation of the safety analysis pertaining to azimuthal tilt was conducted and the results of the reanalysis show that, for the conditions of azimuthal tilt as defined in the new Figure 3.2-1A, the safety analysis of the referenced cycle (Cycle 1) is bounding. Therefore there is no change to the probability or consequences of an accident previously evaluated in the FSAR.
12-1
0 Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change will not create the possibility of a new or different kind of accident from any accident, previously evaluated. The results of the reanalysis were found to be bounded by the reference cycle safety analysis. Relaxing the azimuthal power tilt limit at, lower power levels will not create any new or different kinds of accidents.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change does not involve a significant reduction in the margin of safety. A reanalysis was performed using the proposed azimuthal tilt limits and it was found that the results of the reanalysis were bounded by, the reference cycle safety analysis. Therefore the margin of safety is maintained.
2. The, proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:
(vi) A change which either may result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan: for example, a change resulting from the application of a small refinement of a previously used calculation model or design method.
SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace equipment or components important to safety.
The change is bounded by the existing safety analysis and will not increase the probability of occurrence or consequences of an accident.
The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. By determining that the results of the reanalysis were bounded by the reference cycle safety analysis the field of accidents or malfunctions have not changed. Therefore, there is no increase in the probability for an accident or malfunction of a different type than any previously evaluated in the FSAR.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. To determine the impact performed.
of the change to the azimuthal The tilt limits, a reanalysis was results of the reanalysis were bounded by the reference cycle safety analysis and therefore the margin of safety has been maintained.
12-2
F. ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an, unreviewed environmental question because operation of PVNGS Unit 3, in accordance with this, change, would not:
Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement ,(FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions for Operation and Surveillance Requirements:
3/4 2-4a 12-3
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I I I I I 20 30 50 GO 70 0 100 PERCENT OF R."-.TED THERt'.AL PQWER FIGURE 3.2-1A AZIMUTHAL POWER TILT LIMIT
. VS THERMA POWER (COLSS IH SERVICE)
PALO VERDE - UNIT 3 3/4 2-4a AMEHDMEt)T t<0. 1" 11
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0 0
ATTACHMENT 13 - REFUELING WATER STORAGE TANK DESCRIPTION OF THE TECHNICAL SPECIFICATION AMENDMENT RE UEST The proposed amendment ensures the Refueling Actuation Signal (RAS) trip value of the Refueling Water Storage Tank for recirculation is maintained at the midpoint of the allowable operational values by removing the "greater than" sign from the trip value as set forth in Technical Specification (T.S.) 3.3.2 Table 3.3-4.
B. PURPOSE OF THE TECHNICAL SPECIFICATION The purpose of T.S. 3.3.2 is to ensure that (1) the associated Engineered Safety Features Actuation action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
C. NEED FOR THE TECHNICAL SPECIFICATION AMENDMENT The proposed change to T.S. 3.3.2 Table 3.3-4 will eliminate an ambiguity concerning the level setpoint in relation to the allowable range.
D. BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
1. The Commission h a s pr ov id ed standards for de termznx n g whether a significant hazards consideration exists as stated in 10CFR50.92. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
A discussion of these standards, as they relate to the amendment request, follows:
Standard 1--Involve a significant increase in the probability or consequences of an accident previously evaluated.'he proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because, by maintaining the RAS trip value at the midpoint of the allowable band, the proposed change is more restrictive. This, in turn, limits the operation of the Refueling Water Storage Tank such .that a maximum assurance of protecting the pumps from cavitating is provided. Since the change is still within the limits of the allowable values, the possibility of consequences of an accident previously evaluated will not be increased.
13-1
Standard 2--Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed change will not create the possibility of a new .or different kind of accident from any accident previously evaluated because, by maintaining the trip value at the midpoint of the allowable band, the proposed change is more restrictive. Since the .change reduces the allowabl'e values of the trip to a single value, which was part of the original safety analysis, the possibility of a new or different kind of accident from any accid'ent previously evaluated will not be created.
Standard 3--Involve a significant reduction in a margin of safety.
The proposed change does not involve a significant reduction in a margin of safety because, by maintaining the trip value at the midpoint of the allowable band, the proposed change is more restrictive. By restricting the allowed operation of the Tank even further within the allowable trip values, the Unit does not experience as many possible accidents as before.
Therefore, the change will not reduce the margin of safety.
2. The proposed amendment matches the guidance concerning the application of standards for determining whether or not a significant hazards consideration exists (51FR7751) by example:
(i) A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.
E. SAFETY EVALUATION FOR THE AMENDMENT RE UEST The proposed Technical Specification amendment will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR. The proposed change does not change or replace .equipment or components important to safety.
The change only limits the allowable values of the trip to a single value and is more restrictive by maintaining the trip value at the midpoint of the allowable band. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
The proposed Technical Specification amendment will not create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR. The proposed change is more restrictive by maintaining the trip value at the midpoint of the allowable band. Since the change reduces the allowable values of the trip to a single value which was part of the original safety analysis, the possibility of a different accident or malfunction will not be created.
The proposed Technical Specification amendment will not reduce the margin of safety as defined in the basis for the Technical Specifications. The proposed change is more restrictive by maintaining the trip value at the midpoint of the allowable band. By, restricting the allowed operation of the Tank even further within the allowable trip values, the Unit does not experience as many possible accidents as before. Therefore, the change will not reduce the margin of safety.
13-2
'll 0
0
ENVIRONMENTAL IMPACT CONSIDERATION DETERMINATION The proposed change request does not involve an unreviewed environmental question because operation of PVNGS Unit 3, in accordance with this change, would not:
Result in a significant increase in any adverse environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the staff's testimony to the Atomic Safety and Licensing Board; or
2. Result in a significant change in effluents or power levels; or
3. Result in matters not previously reviewed in the licensing basis for PVNGS which may have a significant environmental impact.
MARKED-UP TECHNICAL SPECIFICATION CHANGE PAGES Limiting Conditions for Operation and Surveillance Requirements:
3/4 3-26 13-3
( TABLE 3.3-4 (Continued)
ENGINEERED SAFETY FEATURES ACTUATIOH SYSTEH INSTRUMENTATION TRIP VALUES ESFA SYSTEH FUNCTIONAL UNIT TRIP VALUES ALLOMABLE VALUES V. RECIRCULATION (RAS)
A. Sensor/Tr ip Units Refueling Water Storage Tank - Low 7.4X of Span 7.9 > X of Span > 6.9 B. ESFA System Logic Not Applicable Not Applicable C. Actuation System Not Applicable Not Applicable VI. AUXILIARY FEEOWATER (SG-1)(AFAS-1)
A. Sensor/Trip Units
1. Steam Generator 5'I Level - Low > 25.8X MR( ) 25.3X MR
2. Steam Generator h Pressure- < 185 psid < 192 psid SG2 > SGl B. ESFA System Logic Hot Applicable Not Applicable C. Actuation Systems Not Applicable Hot Applicable VII. AUXILIARY FEEOWATER (SG-2)(AFAS-2)
A. Sensor/Trip Units
l. Steam Generator k2 Level - Low > 25 BX WR 25.3X MR( )
2. Steam Generator b, Pressure- < 185 psid < 192 psid SG1 > SG2 B." ESFA System Logic Not Applicable Not Applicable C. Actuation Systems Not Applicable Hot Applicable VIII. LOSS OF POWER A. 4. 16 kV Emergency Bus Undervoltage (Loss of Voltage) > 3250 volts > 3250 volts B. 4. 16 kV Emergency Bus Undervoltage 2930 to 3744 volts 2930 to 3744 volts (Degraded Voltage) with a 35-second with a 35-second maximum time delay maximum time delay 1X. CONTROL POOH ESSENTIAL FILTRATION < 2 x 10- pCi/cc < 2 x 10- pCi/cc
II 0
II
ATTACHMENT 14 TECHNICAL JUSTIFICATION FOR THE PROPOSED BASES CHANGES Page B 2-2: Update reference to CPC/CEAC Functional Specifications Page B 2-3: Update reference to current revision of CEN-286 Pages B 3/4 3.1, B 3/4 3.2: Delete reference to Cycle 1 specific analysis The above listed administrative changes are required to ensure clarity and conciseness. The change to page B 2-2 provides the correct document number for reference to the CPC and CEAC functional specifications. The change to page B 2-3 gives the correct revision of CEN-286(V) for reference to the method of calculation for the trip variables for low DNBR and High LPD trips. The change to pages B 3/4 3.1 and B 3/4 3.2 deletes results of analyses performed specifically for Cycle 1,.
which are no longer applicable to Cycle 2.
Pages B 2-5, B 2-6: Change CPC Low Pressurizer Pressure Auxiliary Trip Setpoint from 1861 psia to 1860 psia The CPC low pressurizer pressure auxiliary trip setpoint has changed from 1861 psia to 1860 psia for Cycle 2. This change makes the value of this setpoint the same as all other CE CPC plants. This change is insignificant because the difference of 1 psia between the Cycle 1 and Cycle 2 values is small compared to the uncertainties included for pressure measurement and other conservatisms included in the safety
'nalyses.
Page B 2-6: Change CPC High Integrated Radial Peak Auxiliary Trip Setpoint from 4.28 to 7.00 The CPCs assure that the Specified Acceptable Fuel Design Limits (SAFDLs) on DNB and centerline fuel melt limit are not exceeded during an Anticipated Operational Occurrence (AOO). This is accomplished by the use of CPC generated low DNBR and high LPD reactor trip signals. The CPCs also include auxiliary trip functions to meet additional design bases. One of the auxiliary trip functions is a range trip on several parameters which assures that the core conditions are within the analyzed CPC operating space. Technical specification bases section B 2.2.1 for the DNBR-low trip lists the parameter ranges of validity for the CPC DNBR algorithm and states that operation outside these limits will result in a CPC trip., One of the parameters listed is the 'high limit for the integrated radial peaking factor (F ) ~ ANPP is proposing to change this range limit from 4.28 to 7.00 in order to reduce the possibility of unnecessary plant trips.
CE has determined that the CPC F range limit is not credited to provide a reactor trip in any safety analyses. TKe only setpoint analysis that uses this limit is the CPC overall uncertainty analysis such as that described in the reference. CE has determined that a change from 4.28 to 7.00 will not impact any of the results of the uncertainty analysis since the Fr will not exceed 4.28 in any normal or CPC design transient conditions included in 'this analysis. However, because of the conservative design of the CPCs for Cycle 2 operation, the on-line calculated CPC F could exceed 4.28 under two conditions and result in an unnecessary reactor trip. The two conditions and justification supporting the change are described as follows.
14-1
4i iO
ATTACHMENT 14 (continued)
1. Low Power Operation f,
The CPCs synthesize a core average and- pseudo hot pin axial power distribution based primarily on inputs from excore neutron flux detectors and CEA position information. In order to preserve the integrity of the power distribution synthesis at low power levels (below approximately 15-17% power), a, conservative pre-calculated axial power distribution is used. In order to determine the hot pin power distribution, the pre-calculated axial power distribution is adjusted by, the appropriate Radial Peaking Factor (RPF) for the present CEA configurati:on. For Cycle 2 operation, the pre-calculated axial power distribution and the RPFs for heavily rodded CEA configurations have significantly increased with respect to Cycle 1 values. Thus at low power, heavily rodded operation (ie. reactor startup) the calculated integrated radial peaking factor can exceed 4.28 and result in an unnecessary reactor trip on CPC F' CE has concluded that the CPC DNBR is conservative for F up to 7.00 and thus an auxiliary trip at 4.28 is not required for this r'ondition.
2. Transient Conditions (CEA drops)
For CEA drop events where a significant increase in F can occur, CE has concluded that a CPC DNBR trip is expected to occur when needed r independent of the F r range limit.
The range limit for .Fr is impl'emented in CPCs as a reload'ata block constant.
Therefore, changing thzs constant form 4.28 to 7.00 does not require a CPC software modification.
As a result of the evaluation described'bove, CE and ANPP have concluded that the CPC DNBR calculation remains conservative at a 95%/95% probability/confidence level with on-1'ine calculated intregrated radial peaking factors as high as 7.00. This conclusion will be confirmed on a cycle by cycle basis as part of the normal setpoint verificati'on process.
Reference:
CEN-356(V)-P, Revision 01-P, "Modified Statistical 'Combination of Uncertainties", July 1987 14-2
0 iO II
CONTROLLED BY USER BASES Limiting Safety System Settings for the Low DNBR, High Local, Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and Limitihg Conditions for Operation on DNBR and kM/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design, basis anticipated operational occurrences.
2. 1.2 REACTOR COOLANT SYSTEM PRESSURE 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 Coolant System components are designed to Section III, 1974 Edition, Summer 1975 Addendum,'f the ASHE Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110K (2750 psia) of 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.
2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected'o ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the, Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift,allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Safety Limits of 1.231 and 21 kw/ft, respectively.
Since thes'e trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore .the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density-High trips include the measurement, calculational and processor uncertainties de ~jul-,5'-
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PALO VERDE - UNIT 3
'8 2-2
i
CONTROLLED BY USER BASES REACTOR TRIP SETPOINTS (Continued)
The methodology for the calculation of the PVNGS trip setpoint values, plant protection system, is discussed in the CE Document No. CEN-286(V)gated
+us+ 2q, fflfl, /Rsvp 2p Manual Reactor Tri The Manual reactor trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Variable Over ower Tri A reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions. This trip function will trip the reactor when the indicated neutron flux power exceeds either a, rate limited setpoint at a great enough rate or reaches a preset ceiling. The flux signal used is the average of three linear subchannel flux signals originatiltg in each nuclear instrument safety channel. These trip setpoints are provided in Table 2.2-1.
g- Lo arithmic Power Level - Hi h The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an. unplanned criticality from a shutdown condition. A reactor.
trip is initiated by the Logarithmic Power Level - High trip unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER level is above 10-4X of RATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level decreases to 10-~X of RATED THERMAL POWER.
Pressurizer Pressure - Hi h The Pressurizer Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coo1ant System protection against overpressurization in the event of loss of load without reactor, trip. This trip's setpoint is below the nominal lift setting of the pressurizer safety valves and its operation minimizes the undesirable opera-tion of the pressurizer safety valves.
Pressurizer Pressure - 'Low The Pressurizer Pressure - Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a decrease jn Reactor Coolant System inventory and in the event of an increase in heat PALO VERDE - UNIT 3 B 2-3
CONTROLLED BY USER BASES Local Power Oensit - Hi h (Continued)
a. Nuclear flux power and axial power distribution from the excore flux monitoring system;
b. Radial peaking factors from the position measurement for the CEAs;
c. Delta- T power from reactor coolant temperatures and coolant flow measurements.
The local power density,.(LPD}, the trip variable, calculated by the CPC incorporates uncertainties and dynami'c compensation routines. These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peakLPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the Peak Linear Heat Rate Safety Limit. CPC uncertainties related to peak LPO are- the same types used for DNBR calculation. Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature tfelays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
ONBR - Low ]8AD The ONBR - Low trip is provided to prevent the ON) in the limiting coolant channel in the coze from exceeding the fuel 6'sign limit in the event of design bases anticipated operational occurrences The DNBR - Low tr ip incorporates a low pressurizer pressure floor of psia. At this pressure a ONBR - Low trip will automatically occur. The ONBR is calculated in the CPC util'izing the following information:
a~ Nuclear flux power and axial power distribution from the excore neutron flux monitoring system;
b. Reactor Coolant System pressure from pressurizer pressure measurement; C. Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements; Radial peaking factors from the .position measurement for the CEAs;.
e. Reactor coolant mass flow rate from reactor coolant pump speed; Core inlet temperature from reactor coolant cold leg temperature measurements.
PALO VERDE - UNIT 3 8 2'-S
0 CONTROLLED BY USER SAFETY LIMITS ANO LIMITING SAFETY SYSTEMS SETTINGS BASES DNBR - Low (Continued)
The DNBR, the trip variable, calculated by the CPC incorporates various uncer-tainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits. These uncertainties and dynamic compensa-tion routines ensure that a reactor trip occurs when the calculated core ONBR is sufficiently greater than 1.231 such that the decrease in calculated core DNBR after the trip will not resul't in a violation of the DNBR Safety Limit.
CPC uncertainties related to ONBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
The DNBR, algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.
Parameter Limitin Value
a. RCS Cold Leg Temperature-Low > 470'F
b. . RCS Cold Leg Temperature-High < 610'F
c. Axial Shape Index-Positive Not more positive than + 0.5
d. Axial Shape Index-Negative Not more ne ative than - 0.5
e. Pressurizer Pressure-Low
f. Pressurizer Pressure-High < 2388 psia
g. Integrated Radial Peaking Factor-Low > 1.28
h. Integrated Radial Peaking Factor-High < 4-.28 '7,00
i. guality Margin-Low > 0 Steam Generator Level - Hi h The Steam Generator Level - High trip is provided to protect the turbine from excessive moisture carry over, Since the turbine is automatically
'tripped when the reactor is tripped, this trip provides a reliable means .for providing protection to the turbine from excesssive moisture carryover. This trip's setpoint'does not correspond to a safety limit, and provides protection in the event of excess feedwater flow. The setpoint is identical to the main
. steam isolation setpoint. Its functional capability at the specified trip setting enhances the overall rel'iability of the reac or protec-ion system.,
PALO VERDE - UNIT 3 8 2-6
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CONTROLLED BY USER 3/4. 3 INSTRUHENTATION BASES 3/4.3. 1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATION The OPERABILITY of the reactor protective and, Engineered Safety .Features Actuation Systems instrumentation and bypasses ensures that (1} the associated Engineered Safety Features Actuation action and/or reactor trip will be initiate when the parameter monitored by each channel or combination thereof -reaches its setpoint, (2} the specified coincidence logic is maintained, (3} sufficient redundancy is maintained to permit a channel to be out of service for testing, or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is, required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.
Response time testing of resistance temperature devices, which are a part of the reactor protective system, shall be performed by using in-situ loop current test techniques or another NRC approved method.
The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications of power level,
'RCS flow rate, axial flux shape, radial peaking factors and CEA deviation penalties. Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3. 1 and 6.8. 1) ensur'e that inadvertent misloading of addressable constants into the CPCs is unlikely.
The design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable. If one CEAC is in test or inoperable, verification of CEA position is performed at least every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the second CEAC fails, the CPCs in conjunction with plant Technical Specifications will use DNBR and LPD penalty factors and increased DNBR and LPO margin to restrict reactor operation to a power level that will ensure safe operation of the plant. If the margins are not maintained, a reactor trip wi11 occur.
The value of the DNBR in Specification 2. 1 .is conservatively compensated for measurement uncertainties. Therefore, the actual RCS total flow rate determined by the reactor coolant pump differential pressure instrumentation or by calorimetric calculations does not have to be conservatively compensated for measurement uncertainties.
An an ysis was done to specify a minimum power levy) below which a addi-tional wer reduction i unnecessary even if there is cEA misaliggn t with CEACs out of service.
PALO.VERDE - UNIT 3 8 3/4 3-1
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CONTROLLED BY USER INSTRUMENTATION BASES
.REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM NS RUMENTA ION Continued w
The analysis determined a Power,Operating Limit (POL) power and assumed a CEA misalignment occurred from this~power level. yThe power penalty factor would accommodate changes in radial peaks and one hour xenon redistribution that would occur if there<were a CEA~misalignment wi'th CEACs out of service. The quotjdnt of the POL'ower and~the CEA misalignment Power Penalty factor is the maximum power (50% power) at~which DNBR SA<FDL violation will occur even if 8 tinware is a CEAgiisalignment from POL conditians. Below this~power, extra thermal margip will be available to the, plant. Thus, for CEA misalignment, power reducti'on below this limiti'ng power is unnecessary,.
/
vt'
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The~ owest core .power for a POL was calculated to'e 70% of rated po~er.
This was based on thb following wor'st COLSS fluid conditions.
H~h Temperatur e 580 F Low ASI Un P essure erf low fraction:
-.3
0. 865
/ full 1785 ps>a /
I w Flaw 9', of flow igh Radial Peak: .70 (garlk 5+4+PLR; PDIL =,40% Power)
TPe'surveillance/requirements sp cified for these systems ensure that the overa) 1 system functional capability is maintained~comparable to the original design standards. /The periodic surveillance tests performed at the minimum fre(uencies are sufficient to demonstrate this capability.
The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the safety analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable. The response times in Table 3.3-2 are made up of the time to generate the trip signal at the detector (sensor response time) and the time for the signal to interrupt power to the CEA drive mechanism (signal or trip delay time). The-response-t-iime~r~-taken-from-the-sequence=of-events-Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.
3/4. 3. 3 MONITORING INSTRUtlENTATION 3/4. 3. 3. 1 RADIATION t10NITORIttG IttSTRUMEttTATIQtt The OPERABiLITY of the radiation monitoring channels ensures that:
(l) the radiation levels are continually measured in the areas served by the PALO VERDE - UN IT 3 8 3/4 3-2
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ML17304A832

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