ML20149H174

Proposed Tech Specs Involving Reduction of Required RCS Total Flow to 387,600 Gpm

ML20149H174
Person / Time
Site:	Catawba
Issue date:	02/10/1988
From:	DUKE POWER CO.
To:
Shared Package
ML20149H172	List: ML20149H168 ML20149H174
References
NUDOCS 8802190162
Download: ML20149H174 (10)

Text

. .

(! IABif 2.2.-l y RfACIOR IRIP SYSilM INSIRUMENTA110N IRIP SEIPOINIS fT cn

• 101Al $[NSOR All0WANCE ERROR j[ FUNCIl0NAl UNIT (IA) 2 (S) TRIP SETPOINT ALLOWABLE VALUE

[k 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

'-' 7 m co 2. Power Range, Neutron flux

• • $$ a. liigh Setpoint 7. 5 4.56 0 $109% of RIP * $111.EE of RTP*

Po FJ 3; b. Low Setpoint C.3 4.56 0 $2S% of RIP * $27.1% of RTP"

, $3 3. Power Range, Neutron flux, 1.6 0.5 0 $S% of RIP

• with $6.3% of RTP* with Q$ liigh Positive Rate a time constant a time constant O CD

-> 2 seconds -> 2 seconds

$$ 4. Power Range, Neutron flux, 1. 6 0.5 0 $5% of RIP

• with $6.3% of RTP" with c,g rJ liigh Negative Rate a time constant a time constant o, ->2 seconds ->2 seconds e..s o

" 5. Intermediate Range, 17.0 8.4 0 $25% of RTP* $31% of RTP*

q* Neutron flux e.

6. Source Range, Neutron flux 17.0 10 0 $105 cps $1.4x10jcps

7. Overtemperature of 7. 2 4.47 2.03 See Note 1 See Note 2

8. Overpower AI 4.3 1. 3 1. 2 See Note 3 See Note 4

9. Pressurizer Pressure-Low 4.0 2.21 1.5 >1945 psig >1938 psig***

10. Pressuriier Pressure-High 7. 5 4.96 0.5 $2385 psig $2399 psig II. Pressurizer Water Level-High S.0 2.18 1. 5 $92% of instrument $93.8% of instrument span span

12. Reactor Coolant flow-Low 2.5 1.77 0.6 >90% of loop >89.2% of loop

&cign flow"* i:!;- flow"*

mirimum measured = thimm nasured m;ne'esam meaurea

• RIP = R ED IllERMAL POW [R

• • Loop A ig;; f low = 96,980 gpa

• • • line constants utilized in the lead-lag controller for Pressurizer Pressure-tow are 2 seconds for lead and I second for lag. Chasinel calibration shall ensure that these time constants are adjusted to these values.

. 1 1

POWER OISTRIBUTION LIMITS f 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT  !

CHANNEL FACTOR 1 LIMITING CONDITION FOR OPERATION 4 3.2.3 The combination of indicated Reactor Coolant System total flow rate and R shall be maintained within the region of ;11;= t1; operation shown on Figure W 3.2-3 for four loop operation. 43, t Where:

8 a.

Fh '

g R = 1.49 [1.0 + 0.3 (1.0 - P)]

8 and  !

THERMAL POWER ,

b* P

} = RATED THERMAL POWER ,

I

• N N l

c. Fg = Measured values of Fg obtained by using the movable incore detectors to obtain a power distribution map. The measured values of F g shall ha used to calculate R since Figure 3.2-3 includes penalties for undetected feedwater venturi fouling of 4* 0.1% and for measurement uncertainties of 2.1% for flow and 4%

$ l N

d{ j APPLICABILITY: MODE 1.

for incore measurement of F g.

I v Q u ACTION: 'Med 33%oo M a. With the combinatio of Reactor Coolant System total flow rate nd R within e N ,g the region of .;;;pt ti; operation with the(flow rate less than 3^M00 gp$,

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> reduce the Power Range Neutron Flux-Hi c 3

0 h Trip Setpoint to J .

• below the nominal setpoint by the same amount (% RTP) as the power reduction 4 7

4 e 6 required by Figure 3.2-3. gl4(,. n I g u

b. With the combination of Reactor Coolant System total flow rate and R^;;t;id;

,9 the region ofgec;;-t:ble operation shown on Figure 3.2-3:

1. Within 2 h er: Tep*o af O'N#IU' *i" S ett T 6n '

o C t ACTl04 m.

i

~

RestorethecombinatiohfReactorCoolantdy*stemtotalflow ( l

.9 Q x) j W rateandRtowithinthept;;;li;it;,or 1 e c X) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER [ l c v and reduce the Power Range Neutron Flux - High Trip Setpoint  ;

~+~ f tolessthanorequalto55%ofRATEjTHERMAL?OWERwit g the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.gitWn tb ng

.4 ~2 2. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially beingi;;;;..;.th; et;ve li;it;, verify g

$%3 through incere flux mapping and Reactor Coolant System total flow cd rate comparison that the combination of R and Reactor Coolant System e6 total flow rate are restored to within the ;t;v: li;it;, or reduce l 3

THERMAL POWER to less than 5% of RATED H RMAL POWER within the i next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, g ,p gggj g g;ssekle. eferd*"

CATAWBA - UNITS 1 & 2 3/4 2-9 Amendment No.M (Unit 1) I AmendmentNo.g(Unit 2)

POWER OISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATIOP ACTION (Continued) b ,j,,c) W /or b. 7. .

3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMA POWER above the reduced THERMAL POWER limit required by ACTION4 :. 2. =d/r L, above; subsequent POWER OPERA-TION may proceed provided that the combination of R and indicated Reactor Coolant System total flow rate are demonstrated, through incore flux mapping and Reactor Coo ant System total flow rate cortparison, to be within the regie'nt;f unpt;bk-operation shown on Figure 3.2-3 prior to exceeding t Qo11owing THERMAL POWER levels:

re.sTricted oc f,N'"'n'4/,

a) A nominal 50% of RATED THERMAL POWER, f

b) A nominal 75% of RATED THERMAL POWER, and p c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% [

of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated Reactor Coolant System total flow rate determined by process computer readings orjigital voltmeter measurement and R shall be determined to be within the regictsjf- d

==pt;bh- operation of Figure 3.2-3: re.stdc.ied oc qperw6ssNs

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

resiric4e4 er permissible 4.2.3.3 TheindicatedReacgCoolantSystemtotalflowrateshallbeverified to be within the regigf junAk-operation of Figure 3.2-3 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the most recently obtained value of R, obtained per Specifica-tion 4.2.3.2, is assumed to exist.

4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.

4.2.3.5 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

CATAWBA - UNITS 1 & 2 3/4 2-10 Amendment No. (Unit 1)

Amendment No (Unit 2)

PENAL. TIES OF 0.1% FOR UNDETECTED FEEDWATER VENTURI FOUUNG AND MEASUREMENT UNCERTAINTIES OF 2.1% FOR F1.OW AND 4.0% FOR INCORE MEASUREMENT OF FL ARE I INCt UDED IN THIS FIGURE. 1 l

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4 R . gn.49 n . 0.2n.Pn FIGURE 3'2-3 REACTOR COOLANT SYSTEM TOTAL FLOW RATE VERSUS R - COUR LOOPS IN OPERATION CATAWEA - UNIT 3 1 & 2 _

3/4 2-11 - . _ - _ __

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ATTACID4ENT 2 Discussion and No Significant Hazards Analysis .

l 4

4 l

. _ _ - . _ - - _ _ _ . _ _ _ , , ~ , . _ _ . , . _ _ _ _ _ , - _ _ _ . .

r i

DISCUSSION AND NO SIGNIFICANT RAZARDS ANALYSIS I,

The proposed change to Catawba's Technical Specification 3/4.2.3 and Figure 3.2-3 ,

4 would reduce the required Reactor Coolant System (RCS) total flow from 396,100 l gpm to 387,600 gpm.

Catawba Unit I recently returned to power operation following its end-of-cycle 2 refueling outage. Flow measurements were taken as required by Technical Specification 4.2.3.2 and 4.2.3.3 both prior to the outage and upon startup '

following refueling. Average RCS flows were consistent at approximately 100.3%

of the required minimum total RCS flow. I'ollowing refueling on January 13, 1988, a precision calorimetric test was conducted as required by Technical Specification surveillance requirement 4.2.3.5. This test resulted in the lowering of the RCS elbow tap flow coefficients which are used to convert elbow tap pressure drops to RCS flow rates. Upon insertion of the new constants into the operator aid computer, indicated RCS flow decreased to between 99.9% and 100.1% of the required flow. Since RrS flow has not been consistently above 100% i of the required RCS flow, power has been limited to 98% of the licensed power level of the unit in accordance with Technical Specification Figure 3.2-3. The fact that the RCS flow rate had remained constant throughout the past cycle and  !

had returned to the same value (100.3%) following startup indicates that there is j not degradation in actual RCS flow rate but that there is an amount of  ;

uncertainty attributable to the RCS flow measurement.

The apparent total decrease in RCS flow (from initial startup to the current RCS flow measurements) is about 1.5% of the total RCS flow. The uncertainty associated with RCS precision heat balance (as stated on Figure 3.2-3) is 2.2%.

The apparent decrease in flow is well within the flow uncertainty and is likely i due to this uncertainty as opposed to actual flow blockage. This is further supported by the fact that no steam generator tubes were plugged and no pump work )

was performed that could have affected the flow output of the pumps during the i

, last refueling outage.

An investigation into the indicated decrease in RCS flow rate is being pursued by station and Westinghouse personnel. One of the areas being investigated is the possibility that changes in RCS thermal streaming is causing a change in indicated hot and cold leg RTD temperatures. The precision heat balance calorimetrics test is extremely sensitive to any uncertainty in this parameter.

All applicable FSAR accidents and transients that have been analyzed used an assumed flow which is equal to or conservative to the proposed Technical Specification flow of 387,600 gpm.

l Certain Catawba FSAR Chapter 15 transients, those using the Improved Thermal j Design Procedure (ITDP), are analyzed with a nominal flow rate of 387,600 gpm: l

! I i

. DISCUSSION AND NO SIGNIFICANT HAZARD.,,, LYFIS (Cantinued)

Transient Flow Used In Analysis Analysis Reference 15.1.2a 387,600 1 15.1.2b 387,600 1 15.1.3a 387,600 1 15.1.3b 387,600 1 15.1.3c 387,600 1 15.1.3d 387,600 1 15.2.3a 387,600 1 15.2.3b 387,600 1 15.2.3c 387,600 1 15.2.3d 387,600 1 15.3.1 387,600 1 15.3.2 387,600 1 15.4.2a 387,600 1 15.4.2b 387,600 1 15.4.3a 387,600 1 15.4.3b 387,600 1 15.4.3c 387,600 1 15.4.3d 387,600 1 15.4.4 265,750 2 15.5.1 387,600 1 15.6.1 387,600 1 l Reference 1: 1986 Update of Catawba FSAR, Table 15.0.3-4 Reference 2: In discussions with Westinghouse it was determined that the value given in Reference 1 is incorrect. The correct value i- 265,750 gpm. This correction will be made in the 1987 FSAR Update. Note that the value is less than 387,600 gpm since the initial condition is with only three reactor coolant pumps running.

The appropriate flow rate assumption for the Catawba FSAR Chapter 15 transients not using ITDp is the proposed Technical Specification minirum measured flow, J 387,600 gpm, adjusted down by the flow uncertainty, 2.2%, to give 379,073 gpm. i All of these transients are currently analyzed with flow rates less than this  !

adjusted value and are therefore conservative:

Transient Flow Used In Analysis Analysis Reference l 15.1.4 373,200 1  ;

15.1.Sa 373,200 1 1 15.1.5b 373,200 1 15.2.6 373,200 1 l

15.2.7 373,200 1 15.2.8a 373,200 1 15.2.8b 373,200 1 15.3.3 373,200 1 15.4.1 171,672*

15.4.6 15.4.7a 15.4.7b 15.4.7c

i .

. DISCUSSION AND NO SIGNIFICANT HAZARDS ANALYSIS (Continued)

Transient Flow Used In Analysis Analysis Reference  ;

15.4.7d 1 15.4.8a 373,200 1 15.4.8b 171,672* 1 l 15.4.8c 373,200 1 l 15.4.8d 171,672* 1  :

15.6.3 373,200 1 l

~

15.6.5a 377,000 2 15.6.5b 377,000 2 15.6.5c 377,000 2 l 15.6.5d 377,000 2  ;

15.6.5e 377,000 2  !

15.6.5f 377,000 2  ;

15.6.59 377,000 2 Reference 1: 1986 Update of Catawba FSAR, Table 15.0.3-4  !

Reference 2: June 12, 1987 UHI Removal submittal (However, Table 15.0.3-4 was  !

inadvertently omitted in the FSAR markups for this submittal and 1

l will be changed in the 1987 Update) l

• The transients are analyzed from hot zero power with two reactor l i coolant pumps running j

• • This transient is not sensitive to small changes in RCS flow  ;

The thermal hydraulic design analyses for the latest reload cores, Catawba 1 -

Cycle 3 and Catawba 2 Cycle 2. used the minimum measured flow of 387,600 gpm. It  !

can be seen from this and from the preceding discussion of FSAR Chapter 15 analyses, that all applicable steady-state and transient core thermal-hydraulic .

analyses have been performed with flows equal to, or conservative with respect l to, the proposed Technical Specification minimum mersured flow.  ;

i  !

RCS average temperaturo will remain unchanged with the change in minimum measured i flow. This means that RCS initial fluid and metal stored energy will remain l essentially unchanged. Further, a constant RCS average temperature implies that i the driving temperature difference for primary-to-secondary heat transfer will  ;

remain essentially unchanged. These two parameters, initial energy content and i rate of energy transfer across the steam generator tubes, are the means by which  ;

mass and energy releases influence containment response for the transients analyzed in Section 6.2.1 of the FSAR. Because the change in RCS flow is being made with a negligible change in RCS average temperature, the mass and energy releases calculated in Sections 6.2.1.3 through 6.2.1.5 of the FSAR will not be affected.

From the above discussions it can be seen that the attached proposed Technical 4 Specification will not adversely impact the accident analyses documented in Sections 6.2.1 and 15 of the FSAR nor the steady-state thermal-hydraulic reload j design analyses discussed in Section 4.4 of the FSAR.

1 J

4 i

5 _

s DISCUSSION AND NO SIGNIFICANT HAZARDS ANALYSIS (Continued)

Several wording changes have been proposed to Specification 3/4.2.3. These changes reflect the changes made in the wording on Figure 3.2-3. Figure 3.2-3 was previously revised (amendments 34 and 25) to allow ple.nt operation with flow less than 100%. Five specific regions below 100% power were added to the Figure and plant operation was allowed in these regions as long as the new ACTION a. was complied with. The proposed wording changes to Table 2.2-1, Specification 3/4.2.3 and on Figure 3.2-3 will clarify wording and ACTIONS required when the plant is operating within one of the particular regions. These changes are administrative in nature and are intended to add clarity to the specification for the operators.

10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if operation in accordance with the 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.

The preposed amendment does not involve an increase in the probability or consequences of any previously evaluated accident. The probability and consequences of all applicable accidents have already been revised using the RCS flow which is being proposed. The results of the analyses using the new flow assumptions have been found to be acceptable. Therefore, the current analyses will not be affected by this proposed change.

The proposed amendment will not create the possibility of a new or different kind of accident frem any accident previousiv evaluated. The amendment will not affect the design of the station. The operation of the station will only be affected by allowing a lower total RCS flow rate at 100% power. This lower flow rate has been accounted for in all applicable accident analyses. The results of the analyses using the new flow assumptions have been found to be acceptable. No new modes of operatien will be introduced that have not been analyzed.

The proposed amendment will not involve a significant reduction in a margin of safety. All applicable safety ana' :ses have been performed using the proposed flow rate or a flow rate which is conservative to the proposed flow rate. All accidents analysis results remain within acceptable limits and therefore the proposed change will not significantly impact the margin of safety.

The proposed wording changes to Table 2.2-1 and Specification 3/4.2.3 are administrative in nature and therefore do not involve significant hazards l considerations. The wording change on Table 2.2-1 is clarification to show that i the required loop flow trip setpoints are based on the measured RCS loop flow. l This wording is consistent with the safety analysis and will avoid confusion by specifying the proper flow which is to be used to determine the RCS low flow Reactor trip setpoints.

The Commission has provided examples of proposed Technical Specification changes l which would not involve significant hazards considerations (48 FR 14870). This I change is similar to example (vi). Example (vi) reads:

L DISCUSSION AND NO SIGNIFICANT HAZARDS ANALYSIS (Continued)

"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 calculational model or design method."

This proposed change is the result of a change in the RCS flow assumption which is currently in use in the accident analyses.

For the reasons stated above, Duke power concludes that the proposed amendment does not involve significant hazards considerations.