ML20084S577

Safety Evaluation Supporting Amend 11 to License NPF-15

ML20084S577
Person / Time
Site:	San Onofre
Issue date:	05/18/1984
From:	Office of Nuclear Reactor Regulation
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ML20084S574	List: ML20084S572 ML20084S577
References
TAC-51625, NUDOCS 8406050061
Download: ML20084S577 (3)

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SAFETY EVALUATION AMENDMENT NO. 11 TO NPF-15 SAN ON0FRE NUCLEAR GENERATING STATION, UNIT 3 DOCKET N0. 50-362 Introduction and Summary By letters dated January 25, July 14, and September 23, 1983, the licensees (Southern California Edison Company, San Diego Gas and Electric Com City of Anaheim, California, and the City of Riverside, California)pany, the requested that an amendment be issued to Facility Operating License NPF-11 for operation of the San Onofre Nuclear Generating Station, Unit 3.

The chan a reduction of the Departure from Nucleate Boiling Ratio (DNBR)ge results in rod bow penalty factors of values supported by the NRC-approved Combustion Engineering Topical Report CENPD-225P, " Fuel and Poison Rod Bowing."

Evaluation Specifically the amendments requested by the licensees make the following changes:

(1) Note 5 in Table 2.2-1 of Technical Specifications 2.2-1 is changed by the deletion of a description of the specific methodology used to calculate the minimum Depature from Nucleate Boiling Ratio (DNBR) trip setpoint from the safety system settings.

(2) Section B 2.2.1 of the Technical Specifications is modified by.he addition of a description of the specific methodology used to calculate the minimum DNBR trip setpoint from the safety system settings. The methodology differs from that deleted from Note 5 of Table 2.2-1 in that it includes metholology for incorporation of rod bow penalty factors into the Core Operating Limit Supervisory System (COLSS) and Core Protection Calculator (CPC) calculations of DNBR, (3) The ACTION statement of Technical Specification 3/4.2.4 is changed by requiring the plant operators to " restore" the DNBR to within acceptable limits if it goes outside such limits. The present wording requires the operators to " reduce" the DNBR to within acceptable limits if it goes outside such limits.

8406050061 84051e PDR ADOCK 05000362 P

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. !(4) Technical Specifications 4.2.4.4 and B 3/4.2.4 are changed to incorporate revised, burnup-deperdent DNBR rod bow penalty factors. The revised factors are based on Combustion Engineering Topical Report CENPD-225P.

Items.(1)and(3)aboveareeditorialinnature.

Items (1) and part of (2)

..merely change the location of the description of the methodology used to 1 calculate the min wJm DNBR from Note 5 of Table 2.2-1 to Section B.2.2.1.

Item (3) changes the word " reduce" to " restore", because in some cases, the DNBR may have to be increased, rather than reduced, to restore it to within acceptable limits. Thus, changes (1) 6nd (3) involve no significant hazards consideration.

Items (2) and (4) above incorporete the revised, burnup-dependent rod bow

" penalty factors. As is discussed in CENPD-225P, the revised penalty factors are based on experimental data taken from 16 x 16 fuel element assemblies, such as those used in San Onofre 2.

The previous rod bow penalty factors were based on a conservative extrapolation of data from 14 x 14 fuel element assemblies. Thus, the revised penalty factors are based on data which is more directly applicable to San Onofre 2.

Both sets of penalty factors were selected to provide a 95% probability with 95% confidence that DNB will not occur on a fuel rod having the minimum DNBR during steady-state operation and anticipated operational occurrences. The revised rod bow penalty factors are in agreement with those given in the Combustion Engineering Topical Report, CENPD-225P and its supplements, " Fuel and Poison Rod Bowing." This report has been reviewed and approved by the NRC staff in a letter from C. O. Thomas (NRC) to A. E. Scherer (CE) dated February 15,.1983.

Thus, while the use of the revised rod bow penalty factors may, under some operating conditions, reduce a safety margin, the results of the change are clearly within all acceptable safety criteria.

In particular, the' revised penalty factors provided the same 95/95 percent probability and confidence that DNBR will not be exceeded. Therefore, this amendment is essentially the same as item (vi) of the examples of actions involving no significant hazards consideration given in 48 FR 14870 and is acceptable.

Contact With State Official By copy of a letter dated December 15, 1983 to the licensees, the NRC staff advised the Chief of the Radiological Health Branch, State Department of Health Services, State of California, of its proposed determination of no j

significant hazards consideration. No comments were received.

Environmental Consideration We have determined that these amendments do not authorize a change in effluent L

types of total amount nor an increase in power level and will not result in any significant environmental inpact. Having made this determination, we hava y

further concluded that these amendments involve action which is insignificant

C

. from the standpoint of environmental impact and pursuant 10 CFR Section 51.5(d)

(4), that an environmental impact appraisal need not be prepared in connection with the issuance of these amendments.

Conclusion Based upon our evaluation of the proposed changes to the Sen Onofre, Unit 3 Technical Specifications, we have concluded that: there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and such activities will be conducted in compliance with the Comission's regulations and the issuance of this amend-ment will not be inimical to the comon defense and security or to the health and safety of the public. We, therefore, conclude that the proposed changes are acceptable.

Dated: MAY 18 1984 i

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e MAY 181934 ATTACHMENT TO LICENSE AMENDMENT NO. 11 FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Also to be rep 1' aced are the following overleaf pages to the amended pages.

Amendment Pages Overleaf Pages 2-4 2-3 8 2-6 B 2-5 B 2-7 B 2-8 3/4 2-6 3/4 2-5 B 3/4 2-4 B 3/4 2-3

p 1l TABLE 2.2-1 m,

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS z@

ALLOWABLE VALUES A.

FUNCTIONAL UNIT TRIP SETPOINT b

1.

Manual Reactor Trip Not Applicable Not Applicable i

2.

Linear Power Level - High -

Four Reactor Coolant Pumps 5 110.0% of RATED THERMAL POWER 1 111.3% of RATED THERMAL POWER

,e l

Operating 3.

Logarithmic Power Level - High (1) 1 0.89% of RATED THERMAL POWER i 0.96% of RATED THERMAL POWER 1

4.

Pressurizer Pressure - High 1 2382 psia 5 2389 psia 5.

Pressurizer Pressure - Low (2)

> 1806 psia

> 1763 psia T

6.

Containment Pressure - High 1 2.95 psig i 3.14 psig w

7.

Steam Generator Pressure - Low (3) > 729 psia

> 711 psia 8.

Steam Generator Level - Low

> 25% (4)

> 24.23% (4) 9.

Local Power Density - High (5) i 19.95 kw/ft i 19.95 kw/ft-j

10. DNBR - Low

> 1.20 (5)

> 1.20 (5) 1 11.

Reactor Coolant flow - Low

]

a) DN Rate

< 0.22 psid/sec (6)(8)

< 0.231 psid/sec (6)(8) i 13.2 psid (6)(8)

I 12.1 psid (6)(8) b) Floor j

c) Step 36.82psid(6)(8) 57.231psid(6)(8) 12.

Steam Generator Level - High 1 90% (4) 1 90.74% (4) 13.

Seismic - High 1 0.48/0.60 (7) 1 0.48/0.60 (7) 14.

Loss of Load Turbine stop valve closed Turbine stop valve closed i

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS z

TABLE NOTATION h

(1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically 2

removed when THERMAL POWER is less than or equal to 10 *% of RATED THERMAL POWER.

(2) Value may be decreased manually, to a minimum value of 300 psia, as pressurizer pressure is reduced, 5

provided the margin between the p*essurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is

~

increased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

(3) 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 equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(4) % of the distance between steam generator upper and low level instrument nozzles.

(5) As stored within the Core Protection Calculator (CPC).

Calculation of the trip setpoint includes measurement, calculational and processor uncertainties, and dynamic allowances.

Trip may be y

manually bypassed below 10 *% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10 *% of RATED THERMAL POWER.

The approved DNBR limit accounting for use of HID-2 grid is 1.2.0.

A DNBR t.-ip setpoint of 1.19 is allowed provided that the difference is compensated by an increase in the addressable constants BERRI for CPC and EPOL2 for COLSS.

(6) DN RATE is the maximum decrease rate of the trip setpoint.

FLOOR is the minimum value of the trip setpoint.

STEP is the amount by which the trip setpoint is below the input signal i

unless limited by DN Rate or Floor.

(7) Acceleration, horizontal / vertical, g.

${

(8) Setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.

5m 5

5 4

1

( -_. _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

.(

BASES Local Power Density-High (Continued 1 The local power density (LPD), the trip variable, calculated by Lthe CPC incorporates uncertainties and dynamic compensation rcutines.

These uncer-tainties and dynamic compensation" routines ~ensupe that"a' reactor trip ~ occurs when the actual core peak LPD 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 LPD Safety Limit. CPC uncertainties related to peak LPD are the same types used for DNBR calculation.

Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.

DNBR-Low The ONBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of anticipated operational occurrences.

The DNBR - Low trip incorporates a low pressurizer pressure floor of 1825 psia.

At this pressure a DNBR - Low trip will automatically occur.

The DNBR is calculated in the CPC utilizing i

the following information:

Nuclear flux power and axial power distribution from the excore a.

neutron flux monitoring system; b.

Reactor Coolant System pressure from pressurizer pressure measurement; Differential temperature (Delta T) power from reactor coolant c.

temperature and coolant flow measurements; d.

Radial peaking factors from the position measurement for the CEAs; Reactor coolant mass flow rate from reactor coolant pump speed; e.

f.

Core inlet temperature from reactor coolant cold leg temperature measurements.

The DNBR, the trip variable calculated by the CPC incorporates various uncertainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits.

These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core ONBR is sufficiently greater than 1.20 such that the decrease in actual core SAN ON0fRE-UNIT 3 6 2-5

~

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l

8ASES DN8R-Low (Continued)

DN8R after the trip will not result in a violation of the DNBR Safety Limit.

CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertaintics, and computer equipment processing uncer-tainties.

Dynamic compensation is provided in the CPC calculations for the effects of coolant transport deiays, 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 wi.11 result in a CPC initiated trip.

a.

RCS Cold Leg Temperature-Low

> 495"F b.

RCS Cold Leg Temperature-High I 580*F c.

Axial Shape Index-Positive

< +0.5 d.

Axial Shape Index-Negative

> -0.5 e.

Pressurizer Pressure-Low I 1825 psia f.

Pressurizer Pressure-High i 2375 psia g.

Integrated Radial Peaking Factor-Low

> 1.28 h.

Integrated Radial Peaking Factor-High I 4.28 1.

Quality Margin-Low "

<0 The DNBR Trip sett, int in CPC and COLSS is 1.19.

The values of the penalty factors BERR1 (CPC) and EPOL2 (COLSS) may be adjusted to implement requirements for tripping at other values of DNBR.

The following formula is used to adjust the CPC addressable constant BERR1:

= BW1old [1 + ADNBR(%)4 BERR1

.0D d

)

new where:

BERR1 new required value of BERR1,

=

new BERR1

=

old present implemented value of BERR1, ADNBR(%)

percent increase in DNBR trip setpoint requirement,

=

d(% POL)/d(% DNBR)

= The absolute value of the most adverse deriva(ive of percent POL with respect to percent DNBR as reported in CEN-184(S)-P.

Similarly, for the COLSS addressable constant EP0L2:

d (1+ADNBR(%)*ld 08 EPOL2 l*0.01)*(1 + EPOL20ld)-1.0

=

new where:

EPOL2

=

new required value of EPOL2, new EPOL2 present implemented value of EPOL2,

=

old and the other terms are as previously defined.

SAN ONOFRE-UNIT 3 8 2-6 AMEN 0 MENT NO.11

SAFETY LIMITS AND LIMITING SAFFTY SYSTEM SETTINGS BASES DN8R-Low (Continud)

This illustrates the methodology used for conversion of any DNBR penalty into a format that is useable and addressable in both CPC and COLSS.

The addressable constants BERR1 and EPOL2 are also used to accommodate the DN8R rod bow penalties listed in Technical Specification 4.2.4.4 Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against a reactor coolant pump sheared shaft event and a two pump opposite loop flow coastdown event.

A trip is initiated when the pressure differential across the primary side of eitner steam generator goes below a variable setpoint.

This variable setpoint stays a set amount below the pressure differential unless limited by a set maximum decrease rate or a set minimum value.

The specified,setpoint ensures that a reactor trip occurs to prevent violation of local power density or DNBR safety limits under the stated conditions.

Seismic - High The Seismic - High trip is provided to trip the reactor in the event of an earthquake which exceeds 60% of the Safe Shutdown Earthquake level.

This trip's setpoint does not correspond to a safety limit and no credit was taken in the accident analyses for operation of this trip.

Loss of Load The Loss of Load trip is provided to trip the reactor when the turbine is tripped above a predetermined power level.

This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.

Its functional capability at the specified trip setting enhances the overall reliability of the Reactor Protection System.

Steam Generator Level-High 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 excessive moisture carry over.

This trip's setpoint does not correspond to a Safety Limit and no credit was taken in the accident analyses for operation of this trip.

Its functional capability at the specified trip setting enhances the overall reliability of the Reactor Protection System.

SAN ONOFRE-UNIT 3 8 2-7 AMENDMENT NO. 11

~

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.2 CPC ADDRESSA8LE CONSTANTS The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications such as calorimetric measurements for power level and RCS flow rate and incore detector signals for axial flux shape, radial peaking factors and CEA deviation penalties.

Other CPC addressable constants allow penalization of the calculated DN8R and LPD values based on measurement uncertainties or inoperable equipment.

Admints-trative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inad-vertent misloading of addressable constants into the CPC's is unlikely.

1 SAN ONOFRE-UNIT 3 8 2-8 AMENDMENT NO. 11 l

POWER DISTRIBUTION LIMITS 3/4.2.4 DN8R MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DN8R margin shall be maintained by operating within the region of acceptable operation of Figure 3.2-1 or 3.2-2, as applicable.

APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.

ACTION:

With operation outside of the region of acceptable operation, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; or (2) when the COLSS is not being used, any OPERABLE Low ONBR channel exceeding the DNBR limit, within 15 minutes initiate corrective action to restore the ONBR to within the limits and either:

a.

Restore the ONBR to within its limits within one hour, or b.

Be in at least HOT STANDBY within thn next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 7

4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory Syrtem (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that the DNBR, as indicated on all OPERABLE ONBR channels, is within the limit shown on Figure 3.2-2.

4.2.4.3 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 operating limit based on DNRR.

1 SAN ON0FRE-UNIT 3 3/4 2-5

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.4.4 The DN8R penalty factors included in the COLSS and CPC DN8R calculttions shall be verified at least once per 31 EFPDs to be greater than or equal to the values listed below.

This verification will be made on the basis of the BERR1 addressable constant for the CPC and the EPOL2 addressable constant for the COLSS.

GWO Burnup RTO ON8R Penalty (%)

0-10 0.5 10-20

1. 0 20-30 2.0 30-40 3.5 40-50 5.5 I

AMENDMENT NO. U SAN ONOFRE-UNIT 3 1/4 2-6

POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TILT - T[(Cont [inued)

T is the peak fractional tilt amplitude at the core periphery q

g is the radial normalizing factor 0 is the azimuthal core location O is the azimuthal core location of maximum tilt g

is the ratio of the power at a core location in the presence Pggjg/Puntilt of a tilt to the power at that location with no tflt.

3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SilAPE INDEX represents a 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, of wnich the loss of flow transient is..the mest limiting.

Opera-tion 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.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNOR 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 Ilmits.

The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operaJng limit corresponding to the allowable minimum DNBR.

Reactor operation at or below this calculated power level assures that the limits of Figure 3.2-1 are not violated.

The COLSS calculation of core power operating limit based on the minimum DNOR li'mit includes appropriate penalty factors which provide, with a 95/95 probability /

confidence level, that the core power limit calculated by COLSS (based on the minumum DNBR limit) is conservative with respect to the actual core power limit.

These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, engineering design factors, state parameter measurement, software algorithm modelling, computer processing, rod bow and core power measurement.

Parameters required to maintain the margin to DNB and total core power are also monitored by the CPCs.

Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3,2-2 can be maintained by utilizing a prodetormined DNBR as a function of AXIAL SilAPE INOCX and by monitoring the CPC trip channels.

The above listed uncertainty penalty factors plus those associatent with startup tect acceptance criteria are also included in the CPC's which d'sume a minimum c are power of 20% of RATLD filLRMAL POWLR.

ino 20% Rated Therma Power threshold is due to the neutron flux detector system being inaccurate below 20% core power.

Core noite Icvel at low power is too large to obtain usable detector readings.

SAN ON0fRL-UNii 3 0 3/4 2 3 1

POWER DISTRIBUTION LIMITS BASES The DN8R penalty factors IIsted in Section 4.2.4.4 are penalties used to accommodate the effects of rod bow.

The amount of rod bow in each assembly is depende..t upon the average burnup experienced by that assembly.

Fuel assem-blies that incur higher average burnup will experience a greater magnitude of rod bow.

Conversely, lower burnup assemblies will expertence less rod bow.

The 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 integrated planar-radial power peak.

A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.

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 LOCA safety analyses.

3/4.2.6 REACTOR COOLANT COLO LEG TEMPERATilRE This specification is provided to ensure that the actual valua of reactor coolant cold Icg temperature is maintained within the range of values used in the safety analyses, i

2. 4. 2,7 AXIAt SHAPE INDEX This specification is provided to ensure that the actual value of AXIAL SHAPC !!:0CX is maintained within the range of values used in the safety analyses.

3/4.2,8 PRE 55tlRITER PRESSU_RE This speciffcation is prnvided to ensure that the actual value of pressurizer prossure is maintained within the range of values used in the safety analyses.

AMUN

$AN ON0rRE UNIT 3 0 3/4 2 4

Document Namet SONG AMENDMENT W %I.

t.,

Requestor's ID:

PEGGY Author's Name:

HRood/yt-Document Carments:

M Issuance of Amendment No. W to Facility Operating License e

May 18, 1984 ISSUANCE OF AMENDMENT NO.11 TO FACILITY OPERATING LICENSE NPF-15 SAN ONOFRE NUCLEAR GENERATINr STATION, UNIT 3 DISTRIBUTION Docket File 50-362'"

-NPC POR Local POR PRC System NSIC LB C Readin J. Lee (5) g H. Rood T. Novak it. Saltzman, SAB L. Chandler, OELD C. Hiles H. Denton J. Rutberg A. Toalston W. Miller, LFMB N. Grace E..fordan L. Harmon D. Brinkman,(4)

SSPB T. Barnhart