TXX-4431, Responds to 841005 Request for Addl Info Re Final Draft Tech Specs.Util Will Participate in Industry Resolution of Five Listed Concerns.Answers to Questions & marked-up Tech Specs Encl

Responds to 841005 Request for Addl Info Re Final Draft Tech Specs.Util Will Participate in Industry Resolution of Five Listed Concerns.Answers to Questions & marked-up Tech Specs Encl

ML20113D282
Person / Time
Site:	Comanche Peak
Issue date:	04/09/1985
From:	Beck J TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	Youngblood B Office of Nuclear Reactor Regulation
References
TXX-4431, NUDOCS 8504150188
Download: ML20113D282 (27)

Text

._ _

+

Log # TXX-4431 FHe #

TEXAS UTILITIES GENERATING COMPANY

]o NKYWAY TOWEN. 400 NONTH OLIVE MTHEET, L.H. Mt. IBALI.AM. TEXAM TS208 April 9, 1985 E"."..*0.." 5*

• Director of Nuclear Reactor Regulation Attention: Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission 9 on, D.C.

20555

Wast, t

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION, UNIT 1 DOCKET N0. 50-445 FINAL DRAFT TECHNICAL SPECIFICATIONS REF:

B. J. Youngblood letter to M. D. Spence dated October 5, 1984.

Dear Mr. Youngblood:

The above referenced letter requested additional information concerning Comanche Peak Unit 1 Technical Specifications. The purpose of this letter is to transmit the requested information and to discuss those five items which you identified as requiring extensive effort to properly address.

Attached for your review are responses to your questions and in some cases marked up Technical Specification pages indicating required changes. The indicated Technical Specification changes are required to provide consistency with plant configuration.

As you mentioned in your letter, the following five concerns require substantial effort for resolution: (1) the absence of tech spec requirements for the atmospheric dump valves operability, (2) the absence of an automatic safety injection signal during mode 4 (3) the absence of the capability (per tech specs) to close the steam line isolation valves in mode 4, (4) the inconsistency between the control rod withdrawal accident analysis which assumes two reactor coolant pumps circulating water for cooldown, and the tech specs which allow only one RHR pump operating for cooldown during modes 4 and 5, and (5) the relationship between the process variables values as limited by the tech specs, as measured, and as assumed in the safety analysis.

Because these same issues affect other Westinghouse plants as well as Comanche Peak, we will participate in industry resolution of these questions through the Westinghouse Owners Group or other appropriate sub-group.

Sincerely, 8504150100 850409 b kl /

$DR ADOCK 050 5

ghnW. Beck RWH/grr 0

Attachments c - S. B. Burwell I l A. Vietti d'"""'"""""""""d""""""""'"'""'

Question 1.

Table 3.2-1, DNB Parameters The staff notes the following apparent discrepancies The temperature is limited to Tavg 1 592.7 degrees-F 1.

a.

b.

However, FSAR page 15.0-10 states the temperature error to be 1 5.5 degrees-F.

It follows that the high value of Tavg for analysis purpose should be:

Tavg 1 592.7 + 5.5 = 598.2 degrees-F c.

In a TUGC0 letter dated August 31, 1983, it is stated that Westinghouse calculates a T minimum error allowance of 4.6 avg degrees-F. Therefore, in accordance with this allowance the Tavg value used for analysis should be:

Tavg 1 592.7 + 4.6 = 597.3 degrees-F d.

FSAR page 15.2-29, also, states that the feed water line break is analyzed at the nominal T plus 5.5 degrees-F, i.e. 589.2 + 5.5 avg

= 594.7 degrees-F.

e.

The Technical Specification basis page B3/4 2-6 states that the analysis value of Tavg is 595 degrees-F.

Please address the above and explain the differences.

11.

The pressure limit is 1 2230 psig.

Page 15.0-10 of the FSAR states the pressure error allowance to be 1 30 psi. Also, FSAR Table 15.0-3 states the nominal value of the pressure to be 2235 psig. Accordingly, the low value used for analysis should I

be:

2235 - 30 = 2205 psig,

This value is properly reflected in the basis page B3/4 2-6.

However, the tech spec limit of 2230 psig minus the error allowance of 30 psi equals 2200 psig.

Please explain the difference.

l 2

Response

The Comanche Peak Safety Analysis was performed assuming an initial Tavg of 594.7 degrees-F and an initial pressure of 2220 psia. These safety analysis limits are determined as follows:

Reference Tavg at

+

Allowances for T Safety Analysis

=

avg 100% Power Measurement and Control Limit and S/G Tube Fouling 594.7 degrees-F 589.2 degrees-F

+

5.5 degrees-F

=

Allowances for pressure = Safety Analysis Nominal pressure at 100% Power measurement and control Limit 2220 psia

• 30 psi 2250 psia

=

To ensure that the safety analysis limits are not exceeded, the technical specification limits must take into account the measurement uncertainty for each parameter. For the Comanche Peak plant the calculated measurement uncertainties are 2 degrees-F and 14.1 psi for temperature and pressure respectively. The technical specification limits are then obtained as follows:

Technical Safety Analysis

+

Measurement Uncertainty

=

Limit Specification Limit 592.7 degrees-F 594.7 degrees-F 2 degrees-F

=

(593 degrees-F) 2219.1 psig 2205 psig*

+

14.1 psi

=

(2220 psig) i

• 2220 psia = 2205 psig i

e Using these v.alues' allows tiie measured value to be compared directly to the technical specification limit. The attached technical specifications have been revised accordingly.

-4

acTLBiS34

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TABLE 3.2-1 DN8 PARAMETERS

(

PARAMETER LIMITS Four Loops i

in Operation Indicated Reactor Coolant System T,yg

< 593*F Indicated Pressurizer Pressure

> M psig*

2220 1

1

()

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• Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

1 i

COMANCHE PEAK - UNIT 1 3/4 2-15

~

POWER DISTRIBUTION LIMITS OCIl 91034 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

The indicated T value of 593*F and the indicated pressurizer pressure value of hl$kr psig correspond to analytical limits of 595*F and 2205 psig respectively, with l

allowance fcr measurement uncertainty.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

Measure-ment uncertainties must be accounted for during the periodic surveillance.

t 9

L COMANCHE PEAK - UNIT 1 B 3/4 2-6

Question 2.

Table 3.3-2, Reac_t_or_ Trip, S_ystem Instrumentation Response Times The overtemperature N-16 and overpower N-16 reactor trips have response times of i 7 sec each. However, FSAR Table 15.0-4 shows a delay time of 2 sec each. Please explain the difference.

Response

The 2 second delay time contained in FSAR Table 15.0-4 models only the delay which occurs between generation of the reactor trip signal and the point at which the rods are free to fall into the core. Sensor response times, some signal conditioning delays, etc. are not included in this 2 seconds. An additional 5 seconds is included in the analysis to account for the delays associated with these later factors.

The overall analytical response time is 7 seconds as the technical specifications correctly indicate. The CPSES FSAR will be clarified accordingly (see attached Section 15.0.6).

The respoase time for the overpower N-16 reactor trip was not requireJ for the safety analysis and is added to respond to the NRC staff request.

Testing will be based on an assumed response time of 5 seconds for the N-16 detector.

5-

r CPSES/FSAR be used in the CPSES cores, as previously stated.

{}

The nonnalized rod cluster control assembly negative reactivity inser-tion versus time curve for an axial power distribution skewed to the bottom (Figure 15.0-5) is used in those transient analyses for which a point kinetics core model is used. Where special analyses require use af three dimensional or axial one dimensional core models, the negative reactivity insertion resulting from the reactor trip is calculated dir-ectly by the reactor kinetics code and is not separable from the other reactivity feedback effects.

In this case, the rod cluster control assembly position versus time of Figure 15.0-3 is used as code input.

15.0.6 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES A reactor trip signal acts to open two trip breakers connected in series feeding power to the control rod drive mechanisms.

The loss of power to the mechanism coils causes the mechanisms to release the rod

()

cluster control assemblies which then fall by gravity into the core.

There are various instrumentation delays associated with each trip function, including celays in signal actuation, in opening the trip breakers, and in the release of the rods by the mechanisms.

The total delay to trip is defined as the time delay from the time that trip conditions are reached to the time the rods are free and begin to fall.

Limiting trip setpoints assuned in accident analyses and the time delay

+

==d for each trip function are given in Table 15.0-4.

Reference is made in that table to Overtemperature and Over-power N-16 trip shown in 5

Figure 15.0-1.

The difference between the limiting trip point assumed for the analysis and the nominal trip point represents an allowance for instrunentation channel error and setpoint error.

Nominal trip setpoints are specified in the plant Technical Specifications. During plant startup tests it will be demonstrated that actual instruaent time delays are equal to or w h'. & e u.or s behate.n ganarch.'en oh %d rea. dor feh SQna.l a.nA the To' int ad ch*c.h ne. rods a.ca 4ree h 4 H JANUARY 30, 1981 15.0-14

Question 3.

a.

In the absence of a Technical Specification section that specifies the operability requirements for the plant's atmospheric relief valves (ADVs) discuss the mitigation features for a postulated SGTR event at CPSES (with the assumption that the ADVs are inoperable).

b.

Table 3.3-3 page 3/4 3-16, item 1, Saf_ety_,Inj,ection Provide assurance that an automatic SI signal on containment high pressure in mode 4 is not necessary to mitigate a LOCA or a SLB in the absence of the cold leg accumulators and MSIVs operable.

c.

Table 3.3-3 page 3/4 3-39, item 4, Steam Line Isolation Provide assurance that manual capability of MSIV isolation in mode 4 to mitigate a SGTR is not necessary.

Please provide a plan for resolving the above concerns and justification for interim operation until the resolution has been implemented.

Response

The response to this question will be provided through participation in a Westinghouse Owner's Group or appropriate sub-group action on generic technical specification issues. An SGTR subcommittee to the Westinghouse Owners Group has already been formed to address the SGTR issue.

i 1

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Question 4.

The staff notes the lack of tech specs preventing control rod withdrawal accidents in modes 4 and 5 when only one RHR train may be providing the cooling. The FSAR analysis of rod withdrawal events assumes 2 RCPs circulating primary water for heat removal. Please address this issue. Please provide a plan for resolution and justification for interim nperation until such resolution is implemented.

Response

The response to this question will be provided through participation in a Westinghouse Owner's Group or appropriate sub-group action on generic technical specification issues. --

Question S.

Several Technical Specification Sections The different sections in the Technical Specifications state limits on minimum and/or maximum values of process variables, e.g., temperature, pressure, flow rates, levels, and volumes.

The staff is concerned that these process variable limits are not, in all cases, reflected in the safety analyses.

The staff requires that the applicant: (1) provide justification for not assuming in the safety analyses steady state conditions that are consistent with the limits specified in the tech specs after adding a conservative uncertainty margin, and (2) provide a discussion in the basis for choosing the uncertainty margin. The applicant should make a l

distinction between the value of the parameter as measured, as limited by the tech specs, and as assumed in the safety analyses.

Please provide a plan for resolution and justification for interim operation until such resolution is implemented.

l

Response

I i

The response tc this question will be provided through participation in a Westinghouse Owner's Group or appropriate sub-group action on generic technical specification issues.

1 i

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I

-8 I

Question 6.

Surveillance Requirement 4.1.2.2.d, F,eactivity Control Systems Provide the basis for the 30 gpm minimum boron injection flow rate to the RCS.

Response

30 gpm is the design basis for the Boron Addition System to obtain 1.0%

Delta-K/K shutdown margin at 200 degrees-F within the allotted time specified in the Action Statement.

9

Question 7.

Table 3.4-1 RCS Pressure Isolation Valves State why the high pressure SI flow line check valves are not included in the list of reactor coolant pressure isolation check valves requiring periodic leak checks.

Response

The questioned valves (ISI-8900A, B, C and D and 1-8815) have been added to the attached Technical Specification Table 3.4-1. -

CIN'Al DRAFT TABLE 3.4-1 h

d

_n REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 8948 A, B, C, D Accumulator Tank Discharge 8956 A, 8, C, D Accumulator Tank Discharge

--51-8905 A, B, C, D SI Hot Leg Injection l

8949 A, B, C, D SI Hot Leg Injection 8818 A, B, C, D RHR Cold Leg Injection

--Sf-8819 A, B, C, D SI Cold Leg Infection l

8701 A, B RHR Suction Isolation 8702 A, B RHR Suction Isolation

--WF8705 A, B RHR Suction Isolation Relief l

8841 A, 8 RHR Hot Leg Injection 88)5 SI CoM g Tsolo.%n 890o A, B, C., D SI CoM L Isoldion 3

(

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COMANCHE PEAK - UNIT 1 3/4 4-21

Question 8.

Surveillance Requirement 4.5.2.f, ECCS Explain the differences as listed below for the minimum developed pump differential pressures between the tech spec values and the values assumed in the safety analysis as shown in FSAR Figures 6.3-3,

-4, and -5:

T.S. Value FSAR Value Centrifugal Charging Pump 2350 psid 2390 psid Safety Injection Pump 1435 psid 1485 psid RHR Pump 170 psid 186 psid

Response

The FSAR figures have been changed and now agree with the technical specification values and reflect the current ECCS analysis.

1 e, _

m.. - _ _ _ _ _ -., _.

Question 9.

Surveillance Requirement 4.5.2.h, ECCS (Flow Balance Test) a.

Demonstrate that the minimum SI flow rates specified in Sections 1.a, 2.a and 3.a for each ECCS pump satisfy the minimum SI flow rates for all discharge pressures as shown in FSAR Figure 15.6-47.

b.

Explain why the maxiinum flow rates specified in Sections 1.b, l

2.b and 3.b for the ECCS pumps are to be obtained with two pumps rather tnan a single pump, especially since the flow balance tests discussed in (a) above are performed with one pump. The Westinghouse Standard Technical Specifications specify that this check is to be performed with one pump.

Response

Specification 4.5.2.h has been revised to specify pump runout for one pump.

This is consistent with the Westinghouse Standard Technical Specifications.

Please see the attached revised technical specifications.

The minimum flow rates specified in 4.5.2.h and the minimum discharge pressures specified in 4.5.2.f represent two points on the head vs. flow curve for the respective ECCS pumps. These criteria if met ensure that the ECCS pumps are capable of performing adequately to ensure that the assumptions used as input to the ECCS analysis are valid. This method of ensuring ECCS pump capability is consistent with the Westinghouse Standard Technical Specifications. Additional periodic testing to check other points on the pump curve is not considered to be of significant benefit to the overall understanding of pump capability and therefore is not warranted.

EMERGENCY CORE COOLING SYSTEMS bEM%.UnhTlf"lM rv

(

SURVEILLANCE REOUIREMENTS (Continued) h.

By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the sub-system flow characteristics and verifying that:

1)

For centrifugal charging pump lines:

a)

With a single pump running, the sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 333 gpm, and b)

With bet.one yomy : running, the total pump flow rate is less p '2.. - p than or equal to 476 gpm.

555 2)

For Safety Injection pump lines:

a)

With a single pump running, the sum of the cold leg injec-tion line flow rates, excluding the highest flow rate, is greater than or equal to 437 gpm, and one yum With 5:tb p pp: running, the total pump flow rate is less b) than or equal to-0Er gpm.

{

660 3)

For'RHR pump lines:

a)

With a single pump running, the sum of the cold leg infec-tion line flow rates is greater than or equal to 4652 gpm, and one pom With bet.. p=y,p; running, the total pump flow rate is less b) than or equal to 10,027 gpm.

5,500 1.

By verifying that the refueling cavity to containment sump drain valves, SF-027 and SF-028, are open prior to establishing CONTAINMENT INTEGRITY.

k COMANCHE PEAK - UNIT 1 3/4 5-6

Question 10.

Section 3.5.3, ECCS - Tavg < 350 degrees-F This section specifies that a maximum of one centrifugal charging pump (CCP) and one SI pump be operable at or below 308.7 degrees-F.

The FSAR indicates that the CPSES LTOPS is designed for the flow from two CCPs.

Demonstrate that the LTOP can handle the combined flow from one CCP and one SI pump with a system T at the lowest cold avg leg temperature permissible by the tech specs prior to removal of the reactor pressure vessel head.

Response

The most recent cold overpressure analyses performed for the Comanche Peak Plant assumed the operation of only one centrifugal charging pump.

Therefore both SI pumps and all but one charging pump must be inoperable below 308.7 degrees-F. The attached technical specifications have been revised accordingly.

i

[.

REACTIVITY CONTROL SYSTEMS

(

CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One centrifugal charging pump in the boron injection flow path re-quired by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no centrifugal charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS

(

4.1.2.3.1_The above required centrifugal charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure of greater than or equal to 2350 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All n.tri t;:1 charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.

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i COMANCHE PEAK - UNIT 1 3/4 1-9

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RE* ACTIVITY CONTROL SYSTEMS

(

CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two# centrifugal charging pumps shall be OPERABLE.

APPLICABILITY:

KODES 1, 2, 3, and 4.

ACTION:

With only one centrifugal charging pump OPERABLE, restore at least two cen-trifugal charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borhted to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at let.st two centrifugal charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIREMENTS

{

4.1.2.4.1 At least two centrifugal charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure of greater than or equal to 2350 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.4.2 All ::r.trif ;;;l charging pumps, except the above allowed OPERABLE pump, shall be demonstrated inoperable at least once per 31 days whenever the

[

temperature of one or more of the RCS cold legs is less than or equal to 308.7 F by verifying that the motor circuit breakers are secured in the open position.

A maximum of one :=tr fu;;21 charging pump shall be OPERABLE whenever the i

temperature of one or more of the RCS cold legs is less than or equal to 308.7 F.

C COMANCHE PEAK - UNIT 1 3/4 1-10

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T

< 350 F (j

avg LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump,#

b.

One OPERABLE RHR heat exchanger, c.

One OPERABLE RHR pump, and d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY:

MODE 4.

ACTION:

With no ECCS subsystem OPERABLE because of the inoperability of either a.

{

the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T,yg less than 350*F by use of alternate heat removal methods.

c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the. tctal accumulated actuation cycles to date.

The current value or the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

1. A maximum of one centrifugal charging pump cn3 cnc Safety Injectica p=p shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 308.7 F.

L COMANCHE PEAK - UNIT 1 3/4 5-7

ADDITIONAL INFORMATION As requested the RCS Total Flow vs. R curve for Comanche Peak Unit 1 has been revised to reflect a flow measurement uncertainty of 1.8%.

The attached Figure 3.2-3 of the Comanche Peak Technical Specifications assumes the latest rod bow topical position. The technical specification text has also been revised to reflect a 1.8% RCS flow measurement uncertainty. This value is based on the results of the CPSES revised Improved Thermal Design Procedure (ITDP) report of December 11, 1984..,.,. _ -

_,. _ ~.

FINAL DRAFT POWER DISTRIBUTION LIMITS

(/

3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3 2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation.

Where:

N Fg R = 1.49 [1.0 + 0.2 (1.0 - P)]

a.

b*

P -

THERMAL POWER

, and RATED THERMAL POWER Fh=MeasuredvaluesofF" obtained by using the movable incore c.

detectors to obtain a power distribution map.

The measured valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 includes measurement uncertainties of-h5% for flow and 4%

forincoremeasurementofFh.

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APPLICABILITY:

MODE 1.

ACTION:

With the combination of RCS total flow rate and R outside the region of acceptable operation shown on Figure 3.2-3:

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

I 1.

Restore the combination of RCS total flow rate and R to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

l i

l THIS PAGE OPEN PEN 51N3 RECEIPT OF l

INFORMAilJN iROM rdE AppucAN7 l l l

l COMANCHE PEAK - UNIT 1 3/4 2-8

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_..- FIGURE :3.2-3:

1

[ RCS ' TOTAL :FLOWRATE-VERSUS R - FOUR. LOOPS IN OP. ERA. TION i

m..-p r

i COMANCHE PEAK - UNIT 1 3/4 2-9

~.

e POWER DISTRIBUTION LIMITS i

S O

s s n,

{3 BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) 3.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and 4.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is N intained within the limits.

N F

will be maintained within its limits provided Conditions a. through g

d. above are maintained.

As acted ca Figure; 2.2-O and 0.2-4, "0S fi;w r:te endr$g may t; S : = c m : #,:t ca: r.c h U. :., : le :::: w a %

rata i: ::: ptatic if th: ::::ur:d F i: cl:: 1:w) t; ca;ur; that th: ::lcu lotod ONO" ill net be telc.; the desiga ON"" ;;lue.

TherelaxationofFhas a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts

(

forFhlessthanorequalto1.49.

This value is used in the various accident analysis where F influences parameters other than DNBR, e.g., peak clad tem-H perature, and thus is the maximum "as found" value allowed.

Fuel rod bowing reduces the value of DNB ratio.

Credit is available to offset this reduction in the generic margin.

The generic design margins, totaling 9.1% DNBR, completely offset any rod bow penalties.

This margin includes i

the following:

1 1)

Design limit DNBR of 1.30 vs. 1.28, 2)

Grid spacing (K ) of 0.046 vs. 0.059, 3)

ThermalDiffusi8nCoefficient'of0.038vs.0.059, 4)

DNBR multiplier of 0.86 vs. 0.88, and 5)

Pitch reduction.

The applicable values of rod bow penalties are referenced in the FSAR.

When an F measurement is taken, an allowance for both experimental error q

and manufacturing tolerance must be made.

An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.

l l

L COMANCHE PEAK - UNIT 1 3 3/4 2-4 L

FINAL DRAF8 POWERDISTRIBbTIONLIMITS

(

BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

The Radial Peaking Factor, Fxy(Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit.

The RTPq F

limit for RATED THERMAL POWER (F

) as provided in the Radial Peaking xy Factor Limit Report per Specification 6.9.1.9 was determined from expected power control manuevers over the full range of burnup conditions in the core.

When RCS flow rate and Fh are measured, no additional allowances are necessarypriortocom{a[onwiththelimitsofFigures3.2-3and3.2-4.

Measurement errors of.

for RCS total flow rate and 4% for F have been AH allowed for in determination of the design DNBR value.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3.-

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts.

A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such action action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing q

the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to. confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

!l EHIS PAGE OPEal PenlN3 R INFORMATlJN FROM T.M APPUCANT 1

COMANCHE PEAK - UNIT 1 B 3/4 2-5

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