ML20126G017

Forwards Repts Required by 10CFR50.46, Acceptance Criteria for ECCS for Light Water Nuclear Power Reactors, Indicating Cumulative Change in PCT for Limiting Large Break LOCA ECCS Analysis Due to Changes or Errors in ECCS Evaluation Model

ML20126G017
Person / Time
Site:	San Onofre
Issue date:	12/29/1992
From:	Marsh W SOUTHERN CALIFORNIA EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9301040024
Download: ML20126G017 (3)

Text

,

4 n-TA Southem Califomia Edison Company 23 PARKER STREET

- IRVINE, CALIFORNIA 97718 December 29, 1992 WALTER C. MARSH ltLEPHONE NUCL E. AR Rf C TORY Am$

U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C.

20555 Gentlemen:

Subject:

Docket Nos. 50-361 and 50-362-Emergency Core Cooling-System Annual-10 CFR 50.46 Report San Onofre Nuclear Generating Station Units 2'and 3-4 This letter transmits the San Onofre Units 2 and 3 reports required by paragraph (a)(3)(ii) of 10 CFR-50.46, " Acceptance-Criteria for Emergency. Core-Cooling.

Systems-for Light Water Nuclear Power Reactors." These reports indicate-the cumulative change in Peak Cladding Temperature (PCT) for the limiting large break Loss of Coolant Accident (LOCA) Emergency Core Coolir.g System (ECCS) analysis due to the changes or errors discovered in the ECCS evaluation model is less than I?F.

In addition, this letter provides the cumulative changes to the plant specific assumptions for the limiting large_ break LOCA used in the model.

The revised 10 CFR 50.46 rule became-effective-on 0ctober 17, 1988, and requires reporting to the-NRC at least annually each chan acceptable Emergency Core Cooling System (ECCS) ge to or error discovered in an evaluation model or :in--t!.e -

application of such a model that affects the temperature calculation on the limiting-ECCS analysis. During a review of the-. reporting requirements of 10 CFR-50.46 -it was determined the completed reports for 1988,.'1989,'and 1990 were not i

previously forwarded to the NRC. The:10 CFR 50.46-(a)(3)(ii)Lannual ECCS reporting requirement is now tracked by the San Onofre Regulatory Commitment-o Tracking Sistem to ensure future-submittal of this information on an annual

- basis.

San Onofre I: nits 2 and 3 were initially licensed to the 1974-75 version of the Combustier Engineering-(CE) ECCS evaluation model. CE updated the 1974-75 ECCS evaluation model.and the update was approved by the NRC_in 1986. Since 1988 when the revised-10 CFR 50.46 rule became effective,-minor-changes to or errors

= discovered in the updated ECCS evaluation model have been addresse'd in CENPD-279 (Enclosure 1 for 1988), CENPD-279 Supplement-1 (Enclosure 2:for 1989), and CENPD-2791 Supplement 2 (Enclosure 3 for 1990). There were no changes or errors discovered in the acceptable-evaluation model that affected the limiting ECCS analysis during 1991.

310033

=9301040024 921229

]

y maa wwg

. Document Control Desk provides a summary of all the changes to the ECCS evaluation model since the last base PCT for Cycle 3 was submitted by Southern California Edison on September 5, 1985, ar.d approved by the NRC on May 16, 1986.

The changes to the PCT which were the result of changes to the model are shown under the 10 CFR 50.46 section of this enclosure. The sum of the absolute magnitudes of these changes is less than l'F.

To provide a complete and thorough PCT history, Enclosure 4 also includes a summary of changes to the )lant specific input assumptions for the limiting large break LOCA used in tie model. Changes to plant specific input assumptions are not changes to the application of the model.

These changes to the plant specific assumptions were determined to be within the intended range of the ECCS evaluation model and were performed under the requirements of 10 CFR 50.59.

The arithmetic sum of these changes is less than 48 F.

In addition, the results of the 10 CFR 50.59 evaluations concluded that no unreviewed safety questions exist 1

and the conclusions of the existing ECCS analysis remained valid.

Specifically, the limiting large break LOCA analysis continued to meet all specified acceptance criteria.

If you have any questions regarding these reports, please let me know.

Very truly yours, hN )fflU J

f I

Enclosures cc:

J. B. Martin, Regional Administrator, NRC Region V C. W. Caldwell, NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3 H. B. Fields, NRC Project Manager, San Onofre Units 2 and -3 i

I l

l l

__4..

.a

-m;

-222-~aa 11 _

u.h,,-M uk,4,4 44 s ab a

J_

w_e.

7_.m

,e i

4 e

.4 4

r 5

l ENCLOSURE 1 1

1988 REPORTING PERIOD a

k 1

9 f

i 9

w e.,

5 1

i i

I i

CENPD - 279 i

i i

1 1

t ANNUAL REPORT ON l

C-E ECCS CODES AND METHODS FOR 10CFR50.46 l

i t

i l

TRANSIENT METHODS AND LOCA NUCLEAR FUEL ENGINEERING j.

APRIL 1989 i

l EOMBUSTION ENGINEERING

n,,,,,,.

i I

Abstract This report describes changes and errors in the Combustion Engineering codes ar.d analysis methodology for ECCS analysis from the last approved version through the end of 1988 per the requirements of 10CTR50.46. For this reporting period only one computer code had reportable changes or errors. Tha corrections and changes reduced the peak cladding temperature by less than 1*F.

i i

i

_=

1

'l Table of Centents 4

a J

1.0 Introduction 1

2.0 codes for ECCS Evaluation 2

e a

3.0 Error Corrections and Model Changes in Computer Codes 2

i 3.1 STRIKIN.II 3

4.0 Conclusions 4

5,0 References 4

a e

e 4

a e

11 i

i

1 I

t i

1.0 Introduction Thf6 report addresses the NRC requirement to report changes or errors in licensed codes for ECCS analysis. The revision to the ECCS Acceptance I1) spells out reporting requirements and actions required when Criteria errors are corrected or changes are made in an evaluation model or in the application of a model for operating licensee or construction permittee of a nuclear power plant.

The action requirements in I $0.46(a)(3) are:

1. Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For this purpose, a significant change or error is one which results in a calculated peak fuel cladding temperature (PCT) different by more than $0'F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes'of the respective temperature changes is ' greater than 50'F.

2. For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall-report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in i 50.4. This report is to be filed within one year of discovery of the error and must be reported each year thereafter until a revised evaluation model or a. revised evaluation correcting minor errors is approved by the NRC staff.

3. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may_ be needed to show compliance with I 50.46 1

4 requirements. This schedule may be developed using an integrated scheduling system previously approved for the facility by the NRC,

" For those facilities not using an NRC approved integrated scheduling system, a schedule will be established by the NRC staff within 60 days of receipt of the proposed schedule.

4. Any change or error correction that results in a calculated ECCS performance that does not conform to the criteria set forth in paragraph (b) of I 50.7,6 is a reportable event as described in $$

50.55(e), 50.72 and 50.73. The affected applicant or ~1 censes shall propose i.'amediate steps to demonstrate compliance or bring plant design or operation into compliance with I 50.46 requirements.

This report documents all the changes made to the presently licensed C E IDCA analysis models and methodology which have not been reviewed by the NRC staff. This is specifically to satisfy the requirements described in the second item above.

2.0 Codes for ECCS Evaluation C E uses several digital computer codes for ECCS analysis that are described in topical reports, are licensed by the NRC and are covered by the provisions of CFR 50.46.

Those for large break 14CA calculations are: CEFIASH 4A, COMPERC II, PARCH, STRIKIN-II, and COMZIRC. CEFt ASH 4AS is used in conjunction with COMPERC.II, STRIKIN II, and PARCH for small break 14CA calculations.

3,0 Error Corrections and Model Changes in Computer Codes l

This sect-ion discusses all error corrrections or model changes to the licensed codes which may affect calculated PCT. Only the STRIKIN II computer code has been changed since the last approved submittal to the NRC. No changes to analysis procedures have been made since the last approved submittal to the NRC.

l 2

7 3.1. STRIKIN.II A. Code Description I

STRIKIN.II is a FORTRAN digital computer prograr which is used by Combustion Engineering, Inc. to calculate the core hot spot transient peak clad temperatute (PCT) and peak local clad oxidation percentage for a large break 14CA. It is also used to provide initial fuel temperatures for the small break !.0CA peak cladding temperature calculation. A detailed code description is presented in References 2 through 5.

B. Error in STRIKIN II DNB Model Coding An error in the approved version of STRIKIN II which may potentially-affect calculated PCT has been identified and corrected, Reference (6). A revised version has been prepared for licensing calculations.

Due to a coding error STRIKIN II formerly limited the ' fluid quality to a positive value for the MacBeth correlation. The revised version allows use of a negative fluid quality as appropriate. The impact of of. this correction for a large break thCA is a 0.19*F decrease in the peak cladding temperature. The impact on a small break iDCA could be prediction of DNB slightly earlier than it would actually occur if MacBeth is used with a negative quality. Correction of the error actually produces no change in cladding temperature at the beginning of steam cooling, therefore no change in PCT.

C.

Changes in STRIKIN II Code l

An option has been added to STRIKIN-II to limit the cladding strain rate for the pre rupture strain model to a realistic strain rate

=:

I

-instead of introducing the pre rupture strain in a single time step.

j Without a strain rate limit, STRIKIN II changes the fuel. cladding gap width too quickly when the pre rupture strain model is invoked.

3 l

L..

.==

--- - - - - - - --- '"" ~ '~ '~

Ihis challenges the gap conductivity and temperature model and can j

cause the code to abort. Use of this option produced no effect on PCT.

4.0 concidstons The error correction and the change to STRIKIN.II have the potential to affect the PCT. However, the actual effect of the two changes is to reduce PCT by lesi than 1 F.

This is a very small change in PCT for the limiting transient. There were no significant changes in the sense of CTR 50.46.

This summarizes the error corrections and changes to the C.E 14CA codes and models from the isst accepted versions through December 1988.

5.0 References

1. " Emergency Core Cooling System; Revisions to Acceptance Criteria,"

10CFR50 Federal Register Voi, 53,'No. 180 September 16, 1988 2.-CENPD.135P, 'STRIKIN.II A Cylindrical Coometry Fuel Rod Heat Transfer Program," August, 1974

3. CENPD 135P, Supplement 2. 'STRIKIN II, A Cylindrical Coometry Fuel Rod Heat Transfer Program (Modifications)." Tebruary,1975

4. CENPD 135, Supplement 4.P. 'STRIKIN-II A Cylindrical Coometry Fuel Rod Meat Transfer Program." August,_1976

5. CENPD 135, Supplement 5 P, 'STRIX:N.II, A Cylindrical Geometry Fuel-Rod Heat Transfer Prograa," April, 1977_

6. "STRIKIN II 87316 from 85074,

• CD-TMI. 064, M. Michonski, January '20, 1988.

4

ea-a

+

4%

4

--M

_ W4 =

+

--.-.4 4.5i 6-

_+}.--

aSu_.

4m,,

. as 4 3

b

.2

3. u 12 44_

4s A4 2

aa h

4 g9 9

i se e

i I

ll ENCLOSURE 2 1989 REPORTING PERIOD 4

1 1

4 4

4

.'7 e

r

l_ _.. _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

i j

CENPD - 279 i

SUPPLEMENT 1

k i

i i

i ANNUAL REPORT ON l

C-E ECCS CODES AND METHODS l

FOR 10CFR50.46 i

i l

i 1

l l

l l

TRANSIENT METHODS AND LOCA i

NUCLEAR FUEL ENGINEERING FEBRUARY 1990 i

iar~n -)....-

e

. - - - -.. ~...... -- -. - - -- -..- - - - -

l 4

Abstract

.+

i f

This report describes changes.and errors in the Combustion Engineering codes

{

and analysis methodology for ECCS analysis in 1989 per the requirements of 10CFR50.46. For this reporting period only one computer code had reportable j

i changes or errors. The corrections and changes did not affect the peak-l l

cladding temperature. The cummulative temperature change for large break LOCA.

0 is 'a reduction of less than 1 F, No changes or errors that affect the peak cladding temperature for small break LOCA have occured. Per the criteria of i

10CFR50.46, no action beyond this annual report is required.

1 i

}

i l

j-i-

i d

i-4 t

4

.I 1

h l-1 i

x u+

y

-,,,,w~e,,we.--

n...,..,.-

-e-r~,r

,-%,----,,,-,v.w

---v.r#,m

,4..-o.,

,+%.-

y...,y

. - +.

y me.-+-

.+,.,4

Ju-4-,

M--

-.4~-_Aa

---a a

m

-w--

1w -

a,-4 a

Table of Contents 1.0 Introduction 1

2.0 Codes for ECCS Evaluttion 2

3.0 Error Corrections and Model Changes in Computer Codes 2

3.1 COMPERC l!

3 1

I 4.0 Conclusions 5

5.0 References 5

J t

a e

11

4 1.0 Introduction 4

This report addresses the NRC requirement to report changes or errors in licensed codes for ECCS analysis. The revision to the ECCS Acceptance CriteriaII) spells out reporting requirements and actions required when errors are corrected or changes are made in an evaluation model or in the application of a model for an operating licensee or construction j

permittee of a nuclear power plant, i

The action requirements in i 50.40(a)(3) are:

d

1. Each applicant for or holder of an operating Itcense or construction permit shall estimate the effect of any change to or error in an

~

acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For this purpose, a significant change or error is one which results in a calculated 0

peak fuel cladding temperature (PCT) different by more than 50 F from the temperature calculated for the limiting transient using the j

last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50'F.

2. For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or-licensen shall report the j

nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least-annually as specified in i 50.4. This report is to be filed within one year of discovery of.the error and must be reported each year thereafter untti-a revised evaluation model or a revised evaluation correcting-minor errors is approved by the NRC staff.

3. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or_taking other action as may be needed to show compliance with 9 50.46 1

1 requirements. This schedule may be developed using an integrated

• scheduling system previously approved for the facility by the NRC.

for those facilities not using an NRC approved integrated scheduling system, a schedule will be established by the NRC staff within 60 days of receipt of the proposed schedule.

.l

4. Any change or error correction that results in a calculated ECCS j

performance that does not conform to the criteria set forth in paragraph (b) of 9 50.46 is a reportable event as-described in il 50.55(e), 50.72 and 50.73. The affected applicant or licensee shall propose imediate steps to demonstrate compliance or bring plant i

design or operation into compliance with 9 50.46 requirements.

l This report documents all the changes made to the presently licensed C E LOCA analysis models and methodology which have not been reviewed by' the NRC staff. This is specifically to satisfy the requirements described in the second item above, i

2.0 Codes for ECCS Evaluation C E uses several digital computer codes for ECCS analysis that-are described in topical reports, are licensed by the NRC and are covered by the provisions of CFR 50.46.

Those for large break LOCA calculations are: CEFLASH 4A, COMPERC-II PARCH, STRIKIN !!, and COMZlRC. CEFLASH 4AS I

is used in conjunction with COMPERC !!, STRIK!N !!, and PARCH for small i

break LOCA calculations.

3.0 Error Corrections and Model Changes in Computer Codes ThissecElondiscussesallerrorcorrrectionsormodelchangesto-the licensed codes which' may affect the calculated PCT. Only the COMPERC !!

for a large break has been changed in 1989. No. changes to analysis procedures have'been made since the-1ast approved submittal to the NRC.

2

-w t

es v

y m-y

-+ee yw7

+

1 3.1. COMPERC !!

b 5

A. Code Description i

COMPERC !! is a FORTRAN digital computer program which is used by Combustion Engineering, Inc. to calculate the core refill and reflood transient portion of a PWR loss of coolant accident (LOCA).

A detailed code description is presented in References 2 through 4 j

B. Model Change in COMPERC !! for S! Spillage i

The model for the spillage calculation in COMPERC !! has been changed to reflect a more realistic physical representation. The change is as described below:

)

Present Model (Page 10 of Reference 2) 4 A s Z MAXI If I. MAX < Z A

A A,F (Z, MAXI

• I. MAX) 2 ge#A B

A A

(1)

Wspill "

K spill where W,pgjj : Rate of water spillage out of the break, K,pgjj : 1.oss coefficient for the spillage of water out of the break, A

Flow area in the core, g,p Z

Height of the water in the downcomer, g

Z,ggy : Distance between bottom of core and bottom of.

g Inlet pipe, Z. MAXI: Distance between bottom of core and top of inlet A

pipe, pf

Density of water in the downcomer/ lower plenum, g

Conversion constant.

e 3

L Modification Equation (1) was modified as

' /2 1

2 2

A Z, MAX) 2 9e#A B.F (Z

spill,

A A

(2) g g

spill The difference between Equations (1) and (2) is the first term of the numerator on the right hand side of the equations. This change uses the real head term instead of the fixed head term while the mixture level is in the span of the cold leg.

C. Reasons for the Modification As indicated in the previous section, this change reflects a better physical representation than the model described in Reference 1.

However, a more important reason for this modification is to remove low amplitude flow esci11ations introduced by the discontinuity of the fixed head in the old model.

D. Impact of the Spillage Model Change on PCT The change in downcomer spillage head term has the possibility to affect PCT through two effects

-- reflood rate and-two-phase level.

Comparison of the reflood rates for cases without and with the change in head ters shows that.the small' oscillations in the reflood rate are removed. However, the reflood rate selected for subsequent use is not changed; therefore, there is no change in PCT from this effect. The change in the head term for downcomer spillage also eliminater oscillations in the two phase _ level-but does not change the base two phase level. Elimination of the-oscillations reduces the uncertainty in the two phase level selected for.the next step in the analysis. However, due to the small sensitivity of the C E i

methodology to two phase level changes, the change in the.two phase 4

level due to the change in the head term for spillage has no effect

- on PCT.

i 4.0 Conclusions The change to COMPERC II has the potential to affect the PCT by changing the reflood rate or the two phase level. However, an evaluation of the reflood rates and effect of the two phase level for cases before and after the change in head term for the downcomer spillage shows that there is no change in PCT.

The cummulative change in PCT for large break LOCA including that from 0

the previous annual report, Reference 5, is a reduction of less than 1 F.

There have been no changes in the small break LOCA results to date.

Therefore, there was no significant change in the sense of CFR 50.46 in 1989 and no action beyond the submission of this report is needed.

5.0 References

1. " Emergency Core. Cooling System; Revisions to Acceptance Criteria,"

10CFR50, Federal Register, Vol. 53, No.180, September 16, 1988.

2. CENPD 134P, "COMPERC II, A Program for Emergency Refill Reflood of the Core," August, 1974.

3. CENPD-134P, Supplement 1, "COMPERC II, A Program for Emergency Refill Reflood of the Core (Modifications)," February,1975.

4. CENPtF134, Supplement 2, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core," June,1985.

5. CENPD 279, Annual Report on C-E ECCS Codes and-Methods for 10CFR50.46, April, 1989.

5

~~

'4-o y

4 d

a 60 J

4 d

i 4

ENCLOSURE 3 1990 REPORTING PERIOD 4

f I

e 1

1 CENPD - 279 j

SUPPLEMENT 2

I l

l i

ANNUAL REPORT l

ON i

C-E ECCS CODES AND METHODS i

l FOR 10CFR50.46

\\

l LOCA ANALYSIS AND METH005 NUCLEAR FUEL ENGINEERING.

APRIL, 1991 qmscrLW6to 7pp.

i n

i Abstract 3

This report describes changes and errors in the ABB Combustion Engineering 4

4 codes and analysis methodology for ECCS analysis in 1990 per the requirements of 10CFR50.46. For this reporting period only one computer code had reportable i

changes or errors. The corrections and changes did not affect the peak cladding temperature. The cumulative temperature change for large break LOCA 4

l-is a reduction of less than l'F. No changes or errors that affect the peak cladding temperature for small break LOCA have occured. Per the criteria of l

10CFR50.46, no action beyond this annual report is required, i

i 5

i 4

il i

m y.=,

I Table of Contents 1.0 Introduction 1

2.0 Codes for ECCS Evaluation 2

3.0 Error Corrections and Model Changes in Computer Codes 2

3.1 BORON 3

4.0 Conclusions 4

5.0 References 5

i l

I 11

1.0 Introduction This report addresses the NRC requirement to report changes or errors in licensed codes for ECCS analysis. The revision to the ECCS Acceptance criteria (I) spells out reporting requirements and actions requi ed when errors are corrected or changes are made in an evaluation model or in the application of a model for an operating licensee or construction permittee of a nuclear power plant.

The action requirements in 6 50.46(a)(3) are:

1. Each applicant for or holder of an operating license or construction permit shall estimate the effect of any change to or error in an acceptable evaluation model or in the application of such a model to determine if the change or error is significant. For this purpose, a significant change or error is one which results in a calculated 0

peak fuel cladding temperature (PCT) different by more than 50 F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50'F.

2. For each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the applicant or licensee shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Comission at least annually as specified in i 50.4.

3. If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 9 50.46 requirements. This schedule may be developed using an integrated scheduling system previously approved for the facility by the NRC.

For those facilities not using an NRC approved integrated scheduling 1

system, a schedule will be established by the NRC staff within 60

- days of receipt of the proposed schedule.

4. Any change or error correction that results in a calculated ECCS performance that does not conform to the criteria set forth in l

paragraph (b) of 6 50.46 is a reportable event as described in il 50.5S(e), 50.72 ana 50.73. The affected applicant or licensee shall propose immediate steps to demonstrate compliance or bring plant design or operation into compliance with 9 50.46 requirements.

This report documents all the changes, made in the year covered by this report, to the presently licensed ABB C E LOCA analysis models and methodology which have not been reviewed by the NRC staff. This document is provided to satisfy the reporting requirements of the second item above.

2.0 Codes for ECCS Evaluation ABB C E uses several digital computer codes for ECCS analysis that are described in topical reports, are licensed by the NRC, and are covered by the provisions of 10CFR50.46. Those for large break LOCA calculations are CEFLASH 4A, COMPERC-!!, PARCH, STRIKIN !!, and CONZIRC. CEFLASH 4AS is used in conjunction with COMPERC-!!, STRIKIN-!!, and PARCH for small break LOCA calculations. The codes for post-LOCA long term cooling analysis are BORON, CEPAC, NATFLOW, and CELDA.

3.0 Error Correcticns and Model Changes in Computer Codes This section discusses all error corrrections or model changes to the itcensed codes which may affect the calculated PCT. Only the BORON code for long term cooling analysis of large break LOCAs has been changed in 1990. This change was made to correct an error. No changes to analysis procedures have been made since.the last approved submittal to the NRC.

2

_...___._ _ _ _. _ - _ =.

I 3.11 80RON A. Code Description BORON is a FORTRAN digital computer program which is used by Combustion Engineering, Inc. to calculate the boric acid concentration in the reactor core after a LOCA.

This information is used to determine the point in time, if ever, at which boric acid concentration will reach the solubility limit before core flushing is initiated by starting combined hotside/coldside safety injection.

A detailed code description is presented in Reference 2.

B. Coding Error in BORON The coding that calculates the total boric acid content of the system was found.to have the variable VRCS mistyped as VCRS, where VRCS is the mass of solution in the reactor coolant system (RCS).

Correct coding is used if there is a boric acid storage tank (BAST) in the system, if it empties after the refueling water tank (RWT),

and-if the BAST can't supply as much water as is being boiled off in the core.

Nomally the BAST can't supply as much water as'is being boiled off in the core under these conditions.

Othemise, the incorrect coding is used.

The correct and erroneous equations are described below.

Cgrrect Eauation (Page C-9 of Reference 2)

BT0TL1=VRCS*BRCS/100.0+0.9*VRt.T*BRWT/100.0+VSIT*BSIT/100.0 (1)-

BTOTAL - BTOTl + BBAST*VBAST/100.0 (2)

where ~BTOTAL-: Total boric. acid mass in system (Ibs),

_BBAST : Boric acid concentration in BAST (w/o),

BRCS

Boric-acid. concentration in RCS (w/o),-

3 i

~.

l

BRWT

BoricacidconcentrationinRWT(w/o).

BSIT

Boric acid concentration in safety injection tank (SIT) (w/o).

VBAST

Mass of solution in BAST (1bs),

VRCS

Mass of solution in RCS (1bs),

VRWT

Mass of solution in RWT (1bs),

VSIT

Mass of solution in SIT (Ibe),

Incorrect Equation BTOTL1 - VCRS*BRCS/100.0+0.9*VRWT*BRWT/100.0+VSIT*BSIT/100.0 (3)

The difference between Equations (1) and (3) is the first term on the right hand side of the equations. The error was introduced in 1983 and was detected and corrected in 1990.

C. Impact of RCS Mass Error on PCT The coding error in BORON caused the code to use undefined input for the RCS solution mass for some transients under the conditions described above. This error could affect the PCT by over-predicting the time available before core flushing must be initiated to prevent precipitation of boric acid and possible blockage of the coolant channels in the fuel. However, an evaluation of the effect of this error for those analyses that were performed with the code version with the erroneous coding shows that precipitation does not occur before flushing is started. Since cooling water circulation is not impeded by boric acid precipitation throughout the post-LOCA period, the coding error has no_ effect on PCT for a large break LOCA.

4.0 Conclusions The error in BORON had the potential to affect the PCT by over predicting the time until core flushing must be initiated. However, it is the case that precipitation will not occur before flushing is initiated by 4

=

l startingcombinedhotside/coldsidesafetyinjection. Consequently,there is.,no change in PCT for a large break LOCA due to the code error.

The cumulative change in PCT for large break LOCA including that from the previous annual reports, References 3 and 4, is a reduction of less than 0

3 1 F. There have been no changes in the small break LOCA results to date.

Therefore, there was no significant change in the sense of 10CFR50.46 in

]

1990 and no action beyond the submission of this report is needed.

l 1

5.0 References

1. " Emergency Core Cooling Systemt Revisions to Acceptance Criteria,"

10CFR50, Federal Register, Vol. 53, No.180, September 16, 1988, i-t

2. CENPD 254 P A, " Post LOCA Long Term Cooling Evaluation Model,' June, 1980.

3. CENPD-279, " Annual Report on C E ECCS Codes and Methods for i

10CFR50.46," April, 1989.

~

4. CENPD-279, Supplement 1, " Annual Report on C-E ECCS Codes and Methods for 10CFR50.46," February, 1990.

4 l

l I

I 5

q e

e g

w ENCLOSURE 4

SUMMARY

OF CHANGES TO Tile ECCS EVALVATION MODEL SINCE THE LAST NRC APPROVED-BASE PEAK CLAD TEMPERATURE I

I

i i

i 1

LOSS OF COOLANT ACCIDENT (LOCA) MARGIN

SUMMARY

SAN ON0FRE NUCLEAR GENERATING STATION l

UNils 2 AND 3 A.

Limiting Large Break LOCA Analysis of Record PCT =

2116*F B.

10 CFR 50.46 Changes / Errors Discovered APC1 EFFECT i

l Prior 10 CFR 50.46 Errors / Changes for 1988 APCT= 0.19'F 1

l Prior 10 CFR 50.46 Errors / Changes for 1989 aPCT- 0.00*F Prior 10 CFR 50.46 Errors / Changes for 1990 APCT- 0.00*F f

Prior 10 CFR 50.46 Errors / Changes for 1991 NONE i

l Cumulative Change in PCT 6 PCT = -0.19'F C.

Limiting Large Break LOCA 10 CFR 50.59 Changes j

12/87 Cycle specific changes,' Cycle 4 APCT-

+45'F.

I 12/87 Evaluation of PCT impact from 0.8 j

mil reduction in fuel pellet dia.

6 PCT-

+14*F i

04/88 Evaluation 'of debris-resistant i

fuel aPCT-

<+5'F 08/88 Evaluation of flow measurement i

uncertainties on Technical-i Specification limits.

APCT=

<+20*F 11/89 Cycle specific changes, Cycle 5 APCT-46*F 01/91 Revised physics model resulting i

in " flatter" power distributions.

APCT-

+6*F 11/91 Cycle specific changes,' Cycle 6-APCT-

-l*F

).

j 11/91 Evaluation of Erbium test fuel j

assembly APCT-

+5*F i

Cumulative Change in PCT aPCT=

-+48'F 1

D.

Current Limiting Large Break'LOCA PCT.

Net PCT =

2164*F (Due to the above 10 CFR 50.46 changes /

===----

errors and 10 CFR.50.59 changes)

T

'Last Base Peak Clad-Temperature (PCT); submitted by SCE on 09/05/85 and approved by the NRC on 05/16/86-- Cycle 3 reload.

L' Cycle specific core physics and fuel thermal characteristics.-

N L:

_-