ML19323F831

Discusses Review of WCAP-8720, Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations. Changes Described in Safety Evaluation Are Acceptable for Use W/ Westinghouse PAD-3.3 Code in Plant Safety Analyses

ML19323F831
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	03/27/1980
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
References
NUDOCS 8005290514
Download: ML19323F831 (2)

Text

e a

6f f ** *8 cg"o, d

UNITED STATES

!',(gl,c,//,f( j NUCLEAR REGULATORY COMMISSION e

p j

WASHINGTON, D C. 20555

%, % [f MAR 2 7 gggg

3. ?

e4 J

Mr. Thomas M. Anderson, Manager Nuclear Safety Department Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230

SUBJECT:

REVIEW OF WCAP-8720, " IMPROVED ANALYTICAL MODELS USED IN WESTINGHOUSE FUEL R00 DESIGN COMPUTATIONS" Enclosed is a copy of our March 13, 1980 letter to Comonwealth Edison Company that transmits license Amendment Nos. 53 and 50 for Zion Station Unit Nos.1 &

2, respectively, and our safety evaluation of these amendments. These amend-rents allow an increase in the LOCA peaking factor limit, based on removal of some of the conservatism from the PAD computer code as described in WCAP-8720.

The enclosure includes the NRC Core Performance Branch safety evaluation of the reduction of conservatism in the Westinghouse PAD computer code.

Based upon our review, we have concluded that the changet described in our enclosed safety evaluation are acceptable for use with the Westinghouse PAD-3.3 code in plant safety analyses. This acceptance is limited to the current version of the code as approved by the staff in its February 9,1979 letter to you for application in LOCA analyses.

Although our evaluation has been made as the basis for license amendments for Zion Units 1 and 2, it has been used also as the basis for similar license amendments for Turkey Point Units 3 and 4, and is also applicable to any plant LOCA analysis which uses the Westinghouse PAD code. You should incor-porate these changes and the enclosed Core Perfomance Branch safety evaluation report in the revised approved issue of WCAP-8720.

In the meantime should you incorporate these changes in future safety analyses, please reference the enclosed license amendments for the Zion Station including the Core Performance Branch safety evaluation report.

Sincerely 1

!)

e ZLG Joh F. Stolz, Chief i ht Water Reactors Branch No. 1 Division of Project Management

Enclosure:

March 13,1980 Letter from A. Schwencer (NRC) to D. Peoples (Commonwealth Edison Co.), transmitting license amendments and safety evaluation (m\\)

for Zion Station s

cc w/o enclosure: See next page 8005290514 j

l l

MAR 2 7 'lc80

)

Mr. Themas M. Anderson !

cc: Mr. W. Spezialetti l-Westinghouse Electric Corporation P. 0. Box 355 Pittsburgh, Pennsylvania 15230 Mr. A. Ball Westinghouse Electric Corporation P. O. Box 355 i

Pittsburgh, Pennsylvania 15230 I

I i.

T 4

i f

i l

1 0

i e

o i

gaatoy y

'JNITED STATES y $,,

j NUCLEAR REGULATORY COMMISSION

, t ]

-l wA:HINGTo N. D. C. :20555 e

'%,,,, /

March 13,1980 Docket Nos. 50-295 and 50-304 Mr.'D. Louis Peoples Director of Nuclear Licensing Comonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

Dear Mr. Peoples:

The Conrnission has issued the enclosed Amendment No. 53 to Facility Operating License No. DPR-39 and Amendment No. 50 to Facility Operating License No. DPR-48 for the Zion Station, Unit Nos.1 and 2, respectively.

The amendments consist of changes to the Technical Specifications in response to your application transmitted by letter dated March 22, 1979, as supplemented May 3,1979, and January 25, 1980.

These amendments modify the Technical Specifications, Appendix A to the licenses, to increase the allowable LOCA peaking factor from 1.86 to 1.93 based on an ECCS reanalysis.

Copies of the Safety Evaluation and the Notice of Issuance are also enclosed.

Sincerely,

.4MM4WD A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors I" '

Enclosures:

l.

Amendment No. 53 to OPR-39 2.

Amendment No. 50 to DPR-48 3.

Safety Evaluation 4.

Notice of Issuance Lf b[

cc: w/ enclosures See next page V

j, effr0

\\

Maf(. f vamms

(

'*r. D. Lcuis Feoples Cemnonwealth Edison Company March 13,1980 cc: Robert J. Vollen, Esquire U. 'S. Enviromnental Protection Agency 109 North Dearborn Street Federal Activities Building

~

Chicago, Illinois 60602 Region V Office ATTN:

EIS COORDINATOR Dr. Cecil Lue-Hing 230 South Dearborn Street Director of Research and Development Chicago, Illinois 60604 Metropolitan Sanitary District of Greater Chicago.

100 East Erie Street Chicago, Illinois 60611 Zion-Benton Public Library District 2600 Emmaus Avenue Zion, Illinois 60099 Mr. Phillip P. Steptoe Isham, Lincoln and Beale Counselors at Law One First National Plaza 42nd Floor Chicago,1111 noir 60603 Susan N. Sekuler, Esquire Assistant Attorney General

" - " ~ ~~ ~ -

Environmental Control Division 188 West Randolph Street, Suite 2315 Chicago, Illinois 60601 Mr. W. Bruce Dunbar Mayor of Zion Zion, Illinois 60099 Department of Public Health ATTN: Chief, Division of Nuclear Safety 535 West Jefferson Springfield, Illinois '62761 Director, Technical " Assessment Division Office of Radiation Programs (AW-459)

U. S. Environmental Protection Agency Crystal Mall !2 Arlington, Virginia 20460 e

l

0 neco 9

• 1

[.

UNITED STATES f 70 e NUCLEAR REGULATORY COMMISSION 3 iF( 3C WASHINGTC N. D. C. 20555 5[4b COMM0 WEALTH EDISON COMPANY DOCXET NO. 50-295 ZION STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.53 License No. OPR-39 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Comonwealth Edison Cogany (the licensee) dated March 22, 1979 as supplemented on May 3, 1979 and January 25, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's mies and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provistens of the Act, and the mies and regulations of the Comission; C.

There is reasonable ' assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in cogliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

W

[do yDCOG O f

2-2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-39 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 53, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION hw William P. Gammill, Acting Assistant Director for Operating Reactor Projects Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: March 13,'1980 9

jf**

4 f~

.h*'

UNITED STATES NUCLEAR REGULATORY COMMISSION 3jW,j WASHINGTON, D. C. 20555

• z e

% 'b a

%,...../

~

COMMONWEALTH EDISON COMPANY DOCKET NO. 50-304 ZION STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.50 License No. DPR-48 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated March 22, 1979 as supplemented on May 3, 1979 and January 25, 1980, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility'will operate in conformity with the application -- - - - ---

the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the he..lth and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the cosmon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

j

)

! 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-48 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 50, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment'is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

&, h, wr==::

William P. Gamill, Acting Assistant Director for Operating Reactor Projects Div'ision of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: March 13,1980 e

ATTACHMENT TO LICENSE AMENONENTS A.MENOMENT NO. 53 TO FACILITY OPERATING LICENSE NO. OPR-39 AMENOMENT NO. 50 TO FACILITY OPERATING LICENSE NO. DPR-48 DOCKET NOS. 50-295 AND 50-304 Revise Appendix A as follows:

Remove Pages Insert Pages 45 45 63a 63a D

4 eee 4

e

~

e~w e

LIMITING CONDITION WR OPERATION SURVEILLANCE REQUIRIG4ENT 3.2.2 Power Distribution Limits 4.2.2 Power Distribution A. IIot Channel Fac' tor Limits

• A.

Ilot Channel Factor Limits 1.1 At all times, except during physics 1.1 Following initial core loading tests at $$.75% rated power **, the i

and at a minimum of regular hot channel factors defined in the effective full power monthly bases must meet the following limits:

intervals thereafter, power Units 1 and 2 distribution maps, using the movable detector system, shall O(Z)f o {Z) =d*3.06 x K (Z), for P /_.5 F

93/P x K (Z), for P >.5 F

I l

specification are satisfied and F g fl.55 1+0.2(1-P) XRBP, Following initial loading and each subsequent reloading, a where:

power distribution map using the Hovable Detector System, i

POIZI

=Fg(Z) 11mitJ shall be made to confirm that n

power distribution limits are 1.93 = F constant (LOCA limiting value))

q met, in the full power con-figuration before a unit is P = fraction of rated power at which operated above 75% of rating.

the core operated'during Fg andF],g measurement; K (Z) = factor from Figure 3. 2-9 selected at the core elevation, Z, of the measured Fgr i

• The hot channel factors above a're defined for a period not to exceed the predicted minimum time to collapsa exposure levels for each fuel region as referenced in the bases.

1 o*During physics tests which may exceed these hot channel factor limits, the reactor may be in this condition for a period of time not to exceed eight hours continuously.

. Ameadment No. 53. Unit'l Amendment No. 50, Unit 2 1

Figure 3 2-9 Hot Channel Factor Normalized Operating Envelope for Unita 1 and 2 i

F Constant (LOCA Limiting Value) = 1.93 q

a n

f f 6.

6

.,..i.

! t 6.

i 6

i 4

6...

.e 6.....

.i.

l

. 1 6

6.. 6 i... 4 4

l

l

l

-(6d,4 c000-)

1.0

~.,

- i J.1.32,0,932L.

i i..,

i

, i t

i,

. 4

. il I i.

e t

6 4

.I 6.. e i s I

i., i l 6 1.

6.

.i,

i,,

i

, 6.

\\

t g

\\

.. i

.i.

. 6 i 6 4.

t i i

! i.

4 i.,

i,

,, t

.. i

. i.

, t 1

\\

i

. i.

1 #

6,.

t, t

.g 6 6 a 6 6

4 4 l. il.

\\

\\

g

,.. 6 4 l i e i i i.

I i

.. 6 !

e..

. 6..

\\

i e,

O 3 ' ' I i * ' '

i ' 'i

+

i * ' i i

\\

N

' I ' '.

i i t i

2 - - -

• i '.

e.'

I l af. b.U.IO1 4 4 4 i

+.

6 l

,. 6

. 6.

4

. t i 6 t,

t, 6

, i.

• 4 i

I e

6

,a t !

t a

6,. i 6 i 6 i 6 6,,

t.

. 6 i i i i

+

+

t..

g

. i.

i,.

4.

6...

.. i 6

,. 6 i

4

.. i i

. i, i 6. i,

i e8 t

y

.g' O

.b

4. _ '

. i.,.

"F 6 i 6.

i 6 i I 6 i..

6.

..[m{;j 6

6 i

4 g

i, t i j t i

• .. I

! t..

,. i

. I 4

, i i,

e

. 6.

4 s

. 4.

. l.,

4

,6 6

, i t.

. 6,

, t 6..

6 i eN i

I t

s i g

t...

. t

_. i.

t e

4 i

we

! 6. i

. i

.. 3 I

t 6, '

i i.

. I I

6 i ! i i i 4

i l

I t. t t t t i i e i

! t i t t

i

.. 6

.. t g-M

. 6..

i..

i

,, t 4

. I I !,

, t i 4 4 4 t

, i f

g i

• g "r.

6 i

6., t

... i i.

6 4 i 6

I i t

i 6

1 I t i i

6 4

6 6 i

. I 4

. 8 I

i I i 1

I f

. 4 4 6 t

4 i

6 4 i e i 4 i i 6

1 i

t i

1..

t i

}

j 6 i i. 6 g

a I

)

6 6

I I

i.

i i

i i

. e i s..

! 6. 4 t

e t

• 8 l

t 4 i i,

l

. }

• I

,e i

i 4

. i i

6 i

• g i i.

1 t 6 i,..

f

. 3..

i.

i.

l.

6 i.

. i

(,

t i

. 4 1

e i t i t,.

~

I,.

.. I I

. t t t 6 6. t 4.

i t 6

. t..

i

~

I.

t, i

.. i

.. i.

4. t 6.. t +

... i e e, i..

..,e 6

6 i,

f...

.... 6..

g

. t W4 t 4 i

i t.

4 i i I

I 4, T,. ;

. i i

. i t

.. i.

i,. i _.

! i, 4

. i i i i i,.

i.

i.

, 6 -

. i i i

. 1 i i i..

i

. i 0

6.

o 2

4

.6 8

lo 12 Core Height (Feet) l l

l

-63a..

Amendment No. 53, Unit I l

Amendment No. 50, Unit 2 i

l

\\

m-

,m_S

.r

...g.

m.

c.,

/pa %q$0 UNITED STATES

[ ', y,,.

i NUCLEAR REGULATORY COMMISSION W ASHINGTo N, D. C. 20555 gg "

l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENCMENT NO. 53 TO FACILITY OPERATING LICENSE NO. DPR-39 AND AMENDMENT NO. 50 TO FACILITY OPERATING LICENSE NO. DPR-48 COMM0fMEALTH EDISON COMPANY ZION STATION, UNITS 1 AND 2 DOCKET NOS. 50-295 AND 50-304 Introduction By letters dated March 22, 1979, as supplemented May 3,1979 and January 25, 1980. The Comonwealth Edison Cocpany (the licensee) requested an amendment to Facility Operating License Nos. DPR-39 and DPR-48 for the Zion Station Unit Nos. I and 2.

The application was in support of a request to modify the Technical Specifications, Appendix A to the licenses, to increase the allowable LOCA peaking f actor from 1.86 to 1.93 based on an ECCS reanalysis. The letter contains a LOCA analysis and proposed Technical Specification changes in connection with the operation of Units 1 and 2 with I percent of steam generator tubes plugged and a peaking factor Fg of 1.93 The changes to the Technical Specifications requested by the licensee are the following:

(a) Change of FQ to 1.93 for plant operation with 1 percent of steam generator tubes plugged.

(b) Chnge of the Hot Channel Factor Nomalized Operating Envelope for Units 1 and 2 (Fig. 3.2-9).

Since the limiting value of FQ is below the level at which the excore detectors could provide reliable readings and because the "18 case FAC analyses" perfo."ned for both units indicated that the maximum predicted Fq exceeded the' LOCA determined limits, the licensee is required either to operate the plant with the augmented powar distribution surveillance or at the suitably reduced power levels.

Eva lua tio n The licensee has provided an evaluation of the performance of Emergency Core Cooling System (ECCS) for Units 1 and 2 corresponding to the hot channel peaking factor value of Fgal.93 and assuming a steam generator tube plugging level of 1 percent, a plant specific initial pellet temperature and a removal of 65?F fuel temperature conservatism in the PAD 3.3 fuel performance evaluation code.

360Yorb008

. In performing analyses of the Loss of Coolant Accident. the Westinghouse method starts with a calculation of the volumetric average fuel temperature. For conservatism, an. additional temperature increase is added to the calculated value. This increase consists of two components, one of which is a 65'F margin for modeling uncertainties. This is the claimed difference between the i

predicted value of the volumetric average temperature and the measured value.

The licensee has requested removal of this value from the LOCA calculations on the grounds that adequate conservatism remains.

The licensee provided comparisons of the PAD 3.3 code with appropriate data on fuel pellet temperatures as measured with thermocouples located in the fuel.

This data included fuel rods having dimensions, fuel densities, powers and burnups in the range of interest. These comparisons demonstrated that even without the 65'F modeling uncertainty, the PAD 3.3 code was sufficiently conservative.

In addition the staff performed an independent calculation which utilized studies on fuel temperature uncertainties by Battelle Pacific Northwest Laboratories (Reference 2), and EG&G Idaho Inc. (Reference 3).

Based on these estimates of fuel temperature uncertainty and our calculations, we conclude that the PAD 3.3 computer code meets the criterion of bounding a large portion of the expected volumetric average fuel temperatures when using nominal input conditions without the 65'F.

We also find that there is sufficient remaining margin of conservatism to bound the expected uncertainty in other state of the art fuel performance computer codes.

It is therefore acceptable to delete the 65'F frem the PAD 3.3 computer code.

The detailed description of our evaluation is included in Annex 1.

In addition to the request to drop the 65'F model uncertainty, the licensee has requested the use of as-fabricated fuel parameters applicable to Zion fuel rather than more bounding values usually used by Westinghouse in LOCA analyses. The staff has reviewed the statistical methods and assumptions which the licensee will use for determining the dimensions to be used in the LOCA analyses and finds these acceptable. The licensee has proposed taking ' credit for a 20*F decrease in the volumetric average temperature, due to the difference in assumptions about the as fabricated fuel parameters.

This is a conservative estimate of the expected change.

The LOCA analysis was performed using the February 1978 version of the Westinghouse ! valuation Model (Reference 2) which was reviewed and approved by us.

It was perfomed for a spectrum of three double ended cold leg

~

guillotine breaks (DECLG) with discharge coefficients of C =0.6, 0.8 and 0

1.0.

The input parameters assumed in the analysis are listed below:

i l

3 Core Power: 102 percent of 3250 MWt (rated power)

Peak Lineer Power: 102 percent of 13.086 kw/ft Peaking Factor: 1.93 Accumulator Water Volume: 818.65 cu ft/each The results of the analysis indicate a peak cladding temperature of 2157'F.

a maximum local Zr-water reaction of 6.71 percent and a total Zr-water reaction of less 0.3 percent, all these values occurring at the critical break size of C =D.8.

D The licensee has performed the "18 case FAC analyses" for Cycle 4 in Units 1 and 2 (Reference 1) because tne limiting peaking factor in the LOCA analysis was below the value for which the excore detectors could give reliable results. The results of these analyses have indicated that for both units.

the predicted maximum peaking factor exceeds the limiting value of Fg. The Itcensee is therefore required either to limit power to the rated power multiplied by the ratio of 1.93 divided by the predicted peaking factor or to operate the plant at higher power levels with augnented axial power distribution surveillance in order to ascertain that the peaking factor would not exceed the limiting value cf 1.93.

This requirements could be lifted any time during plant operation if the licensee demonstrates by the "18 case FAC analysis" that the maximum predicted FQ is within the LOCA determined l imi t.

Conclusions Based on the review of the submitted documents, we conclude that the results of the LOCA analysis performed with FQ=1.93 are conservative relative to the 10 CFR 50.46 criteria. We consider the resultant changes to the Technical Specifications acceptable for operating Units 1 and 2 with a maximum 1 percent of steam generator tubes plugged.

References 1.

Letter from Cordell Reed (Commonwealth Edison) to H. R. Denton (NRC),

dated March 22, 1979.

2.

M.

. Cunningham, " Stored Energy Calculation: The State of the Art,"

Battelle Pacific Northwest Laboratories, PNL-2581, May 1978.

3.

D. R. Coleman, E. T. Laats and N. R. Scofield, "FRAP-53: A Computer Code for the Steady-State Analysis of Fuel Rods, Volume 2, Model Verification Report, EG&G Idaho, Inc. Report TFBP-TR-228, Aucus t,1977.

. Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determinaticn, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environnental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of these amendments.

Conclusion We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in co:pliance with the Commission's regulations and the issuance of these amendments will not be inimical to the coeren defense and security or to the health and safety of the public.

Date: March 13,1980 l

SAFETY EV ALU AT 10 tl 0F THE REQUEST FOR

~

REMOVAL OF C0?iSERVATISM If1 THE WESTIflGHOUSE PAD COMPUTER CODE ps REGy,My C

4o O

a 4(.;sa e

S V*%i /j v. '1 n,.s MARCH 1980 CORE PERFORMANCE BRANCH UNITED STATES NUCLEAR REGULATORY COMMISS10N W ASHING"CN. D. C. 20E45 egooVOg 2//

O e

TAELE 0~ COMTENTS par,E 1.

Introduction.........................

1 2.

Request for Reduction in Conservatism.............

2 3.

Conservatism in the Westinghouse PAD Computer Code......

3 3.1 Design vs. As-Fabricated Conditions...........

5 3.2 Fuel Model Conservatism..... '............

6 3.3 Audit of the Revised Analysis..............

10 4

Staff Criterion for Margin of Conservatism..........

15 4.1 Basis for Marcin of Conservatism.............

15 4.2 The Staff Criterion...................

18 5.

Application of the Criterion. '................

19 5.1 The Westinghouse PAD Computer Code 22 5.2 The NRC Fuel Performance Codes..............

23 5.2.1 FRAP-S3......................

23 5.2.2 FRAPCON-1.....................

24 5.2.3 GAPCON......................

25 5.3 The State-of-the-Art...................

26 6.

Conclusions..........................

27 7.

References..........................

28 O

N

=

a w

!'"C'v EY_'.'!?" 0F THE R E0'.'EST FO'. P.E"0"?.L OF CONSERVATISM IN THE WESTINGHOUSE PAD COMPUTER CODE 1.

Introduction The thermal conditions within the fuel of a light. water reactor during its normal lifetime must be described in the safety analysis for each reactor. The fuel temperatures are'used as initial conditions in describing the response of the reactor to a number of hypothetical transients and accidents, such as the loss-of-coolant accident (LOCA).

Connonwea1th Edison Company (Com-Ed), the owner of Zion Station 1 and 2, has requested a license amendment for these two plants to increase the allowable LOCA peaking factor limit from 1.86 to 1.93. This change is based on reanalysis of the loss-of-coolant accident wherein reduced initial fuel temperatures are assumed. The reduced fuel temperature are a result of the removal of some of the margin of conservatism in the fuel performance code used in the analysis. Both Connonwealth Edison and th,e code developer, Westinghouse Electric Corporation, believe that the remaining conservative feature's of the code are adequate for the safety analysis.

A review of the proposed revisions to the fuel code, PAD-3.3, and our evaluation of these changes, are presented in the following sections.

The discussion will consist of a technical review of the submittal, com-parision of the Westinghouse code with a traditional staff audit code, and the development and application of a new criterion for margin of conservatism in codes of this type. All of these methods lead us to conclude that the remaining conservative features of the code are adequate for safety analysis.

. 2.

Key suf i for K6 duction 1h L.OnServatism

~

On March 22, 1979, Commenwealth Edisen Company requested (Ref.1) a license amendment to increase the allowable LOCA peaking factor limit.

This request was based on revised analysis of the emergency-core-cooling-system (ECCS) to meet 10 CFR 50.46 (Ref. 2) requirements. The LOCA peaking factor, also known as the limiting heat flux hot channel factor or the limiting F, is defined as the maximum local heat flux on the g

surface of a fuel rod divided by the core average fuel rod heat flux.

The maximum allowable local heat flux is calculated in the plant safety analysis, usually for the loss-of-coolant accident. For most reactors, the LOCA peaking facter limits the operational flexibility or power maneuvering capability, but not the total power generating capability, of the plant. In the case of Zion Station Unit 1, there is some evidence (Refs. 3-6) that the LOCA peaking factor may also limit the power production capability of the plant.

The potential peaking factor limitation caused Com-Ed to reanalyze the loss-of-coolant accident for the Zion facility. An increase in the allowa,ble LOCA peaking factor from 1.86, to 1.93 was projected based on reduced fuel temperatures. The reduced temperatures were calculated with a modified version of Westinghouse PAD-3.3 code (Ref. 7).

We reviewed the Com-Ed submittal and requested (Refs. 8-9) additional information with regard to the proposed changes. Com-Ed responded to these req 0ests with additional information (Refs.10-11), which has also been reviewed. The details of the Com-Ed proposal are discussed in the next section, i

1 O

y...

. 3.

Coro avctisi.. in the Westinghouse FAO Computer Code The Nestinghouse PAD computer code iteratively calculates the interrelated effects of temperature, pressure, cladding elastic and plastic deformation, fission gas release, and fuel densification and swelling as a function of time and power density. The most recent version of the code, PAD-3.3, was described by Westinghouse in a Licensing Topical Report (Ref.

7). This report was previously reviewed and approved by the NRC staff (Ref. 12 ).

As part of the emergency-core-cooling-system evaluation requirements (Ref.13) for plant safety analysis, the PAD code utilizes a number of conservatisms in the prediction of fuel temperatures. These include conservative inputs to the code, conservatisms within the code itself, and conservative margin applied to the code output. Some of these con-servatisms, such as the 102% of maximum allowable power that is input to the code, are specifically required by the regulations. Other conservatisms, such as the conservative margin applied to the code output, are not specifically required by law. These additional conservatisms were submitted by Westinghouse as part of earlier safety analysis reports or were required by the NRC staff during the review process.

It is the second category, those conservatisms which are not specifically required by law, which is the subject of this report.

The derivation and application of conservative margin applied to the PAD code output have been described previously (Ref.14). The margin is due to unce,rtainties in the fuel temperature predictions due to manufacturing variations. The parameters considered include:

a) cladding inside diameter b) pellet outside diameter

~

~

c) pellet density d) pellet sintering temperature s

e e

j

_4 Varie:icr: in the first two p:r: meters affect the calculated fuel-to-cladding gap size. Variations in the last two parameters affect the densification behavior of the fuel.

j For safety analyses, nominal design values of the above parameters are used as input to the FAD code and allowance for manufacturing variations are then added. The allowance for each of these four parameters is 1

determined by using a bounding value for each quantity. The allowance is simply the difference between the nominal'and the bounded input code prediction. As an example, the PAD code will predict higher temperatures if an upper bound cladding inside diamete'r is used as input rather than the nominal design value of this parameter. The difference between the two predictions, in degrees Fahrenheit, is the allowance for manufacturing variatiors in cladding inside diameter. The bounding value for each input parameter is derived on a normally applied 95% prob' ability basis at a 95% confidence level. Each allowance is calculated at the time in life when fuel temperatures are maximum and at a power level of 15 kW/f t, the approximate LOCA limit.

The allowance calculated for each of the four input parameters is statistically combined with the others to form the total fabrication uncertainty. To the total fabrication uncertainty, a second, so-called model uncertainty, of 65'F is added. It is the Westinghouse position that this additional 65'F margin was added to ensure that the best-1 estimate model predictions would bound most of the experimentally measured fuel temperature data. The best estimate model is the same PAD-3.3 code using nominal input values 2nd no explicit internal or external code conserva tisms. At the time the 65'F margin was accepted, it was the staff's opinion that this margin was used to account for uncertainties not explicitly considered elsewhere. Both Comonwealth Edison and Westinghouse believe that the 65 F margin is already considered else-where,in the analysis and that the remaining conservative features of the code are adequate.

Inc'"

• r....r ;t t, 'rg tL: C5F model uncertcinty, Ca n:r Edison has requested that the use of as-fabricated, rather than as-desianed, fuel conditions be allcwed in the safety analysis. Generally.

this would permit the use of nominal values of cladding inside diameter, pellet outside diameter, pellet density, pellet sintering temperature and their respective uncertainty allowances based only on the fuel supplied to each Zion Unit, rather than the entire Westinghouse product line. In practice, the dimensional parameters (i.e., cladding I.D. and fuel 0.D.) and their uncertainties do not change significantly. As a result, the use of as-fabricated fuel conditions affects only the pellet density, pellet sintering temperature and their respective un-certainties.

A complicating feature of the request is a change in the current product line a..alysis by Westinghouse.

or the current Westinghouse fuel design,

~

the fuel is sintered in such a tranner that the statistical lower bound of the actual sintering temperature is always above the sintering temperiture used as input to the PAD code for safety analyses. This means the code input value is 1cwer than virtually all of the sintering temperatures used ip manufacturing the fuel. As a result, the code predicts more densif'ication than is expected, but the' allowance for uncertainty in sintering temperature becomes zero.

The Com-Ed request for the use of as-fabricated values would cause the analysis to revert back to its criginal fonn. Namely, the use of a nominal sintering temperature ard a non-ze's allowance for uncertainty in this temperature. Because ttese values would be based only on fuel in each Zion Unit, the result 15. a higher sintering temperature, an almost unchanged total fabrication uncertainty, and a reduction in average fuel temperature predictions of approximately 20'F.

3.1 Design vs. As-Fabricated Conditions The request for the use of as-fabricated, rather than as-designed fuel parameters is fundamentally sound. The LOCA' analysis should,

~

~

. ir. ;.... :1, apply to a s;. :ific plant -atner tnan a generic design.

'Je therefore agree with the proposed use of plant-specific input conditions.

We have, however, examined the proposed change to determine whether the approach is indeed applicable to Zion and is statistically valid. Comonwealth Edison has stated that they will use generic fuel parameters (i.e., sintering temperature and pel'.et density) which are bounding to the actual values' determined for each Zion reload core but not for the entire Westinghouse product line.

j These parameters will be selected ori the basis of previous fuel reload data and anticipated future reload data. However, the fuel for each future reload will be measured to ensure that it meets the acceptance criteria for all fuel batches used in that reload. The statistics for the reload region are based on the complete set of data for all batches. At each Zion reload review, Commonwealth Edison will verify that the specific fuel parameters are bounded by the Zicn generic fuel parameters.

Westinghouse, the fuel supplier for Zion Units 1 and 2, has also described (Ref.11) the statistical methods on which the change will be based. For Zion Unit 1, Region 7, the data include over 28,000 density measurements and the sintering temperature for each sintering host used in manufacturing the fuel for the region (3.3 million pellets). The large number of observations used in the process is well in excess of the levels required for proper statis-tical analysis. We conclude that the request to use as-fabricated, rather than as-designed, fuel parameters is acceptable.

3.2 Fuel Model Conservatism As discussed previously, safetf analyses with the PAD code currently apnlies a conservative margin to the fuel temperature predictions of the code. *This margin is composed of a component due to

. f atm:! ion uncertain'.ies (claccing ano peilet diameters, pellet density and sinterirq temperature) and the 65'F component termed tle m:.dct ancertair '.y.

Westinghouse has stated that the final component was added to ensure that the best estimate code prediction; would bound most of the experimentally measured fuel temperature data. 13ecause the evaluation model or conservative version of PAD, rather than the best estimate version, is used in safety analyses, j

it is not necessary (from the regulatory standpoint) that the best estimate-version bound any data. It is'our position, rather, that whichever version of PAD is used in plant safety analyses should conservatively predict fuel temperatures. This requires an explicit or implicit consideration of uncertainties, including uncertainties in the models used.

Both Comanwealth Edison and Westinghouse believe that the 65'F l

margin is already considered elsewhere in the analysis. They further state that the remaining conservative features of the code are adequate for ECCS analysis. The bases for this statement are:

13 The "best estimate" version of PAD bounds the majority of the experimental fuel temperature data considered.

2.

The evaluation model or safety analysis version of PAD always predicts fuel temperatures greater than or equal to those I

predicted by the 'best estimate" version of the code.

3.

'The limiting time in life for ECCS analyses is such that the conservative version of PAD always predicts fuel temperatures greater than those predicted by the best estimate version.

4.

The overall conservatism in the calculation of fuel temperatum, the conservative application of those temperatures in the LOCA analysis, and the conservatism associated with the overall LOCA/ECC5 evaluation warrant the elimination, of the 65'F fuel temperature model uncertainty.

O

. 'l: i :11 discuss 6cch of these bases individu6lly.

1.

Westinghouse has submitted the results of a number of com-parisons between the best-estimate predictions of the PAD-3.3 code and experimentally determined fuel centerline temperature data. The data were taken from the Halden Heavy Boiling Water Reactor (AE-318, HPR-80 and IFA-226) and the Materials Testing Reactor (NAPD-228). Westinghouse selected these data because they represent helium-filled cylin'drical fuel rods near beginning-of-life with densities and gap sizes typical for the standard product line. We agree that these data are representative of the current Westinghouse product. However, it is clear that the "best-estimate" version of PAD-3.3 is not a best-estimate code at all, but a conservative one. In other words, even with nominal input values and the removal of explicitly con-servative models within the code, PAD-3.3 tends to overpredict the experimental data.

2.

When the PAD code is used in safety analysis, certain evaluation model options are activated.. These include the' fuel densifi-c3 tion, gap conductance (gap closure) and cladding creep model s.

Because all of these processes are time-dependent, the difference between the consbryative and best-estimate rptions is zero at time zero. At all non-zero exposures, the j

evaluation model options do result in higher fuel temperature predictions as stated by Westinghouse.

3.

Appendix X of 10 CFR 50 states that "the steady-state temperature distribution and stored energy in the fuel before the hypothetical accident shall be calculated for the burn-up that yields the highest calculated cladding temperature (or, optionally, the t

i

~

\\

\\

l l

{

i f

. 1 caicuiates s:creo energy.)"

(Ref.13)

In the Westing-

..22.

house safety analysis, the limiting burnup occurs shortly

~

ef ter beginning-of-life, at the point of maximum fuel densifi-cation. If the best-estimate version of PAD is used, however, the burnup at which maximum burnup occurs is different than that calculated with the evaluation model options. For'the best-estimate code, this burnup is at, rather than shortly af ter, beginning-of-life. We believe that the margin of conservatism between the best-estimate and evaluation model versions of PAD is misleading when measured at any specific burnup. We conclude that the difference between the maximum temperatures predicted by each version of the code is a more appropriate basis of comparison. The margin calculated in this fashion is significant, but less than that assumed by Westinghouse.

4 The overall conservatism in the calculation of fuel temperatures and the overall conservatism in the LOCA analysis have not been rigorously demonstrated by Commonwealth Edison or Westinghouse.

Only individual details of the analysis, such as the impact of LOCA peaking factor uncertainties on fuel temperatures, have been described. The impact of the proposed modification on the overall analyses has not been addressed. As a result, we are unable to consider the overall conservatism in the LOCA analysis as a basis for the removal of the 65*F model uncertainty in PAD-3.3. This conclusion is discussed further in Section 4.1 of this evaluation.

We conclude that Westinghouse has shown the evaluation model or safety analysis version of PAD-3.3 to be conservative in calculating steady-state fuel temperatures for LOCA analysis. This alone is not sufficient to demonstrate that-the degree of conservatism, when used in LOCA analysis, is sufficient to warrant elimination of the 65*F model uncertainty. We will discuss this further in Section 4

,,,e w

-w

- - +--

. 3.3 /c.f.t cf tr.c Reviced An:ly:is

~

The historical method of regulatory review of fuel codes may be divided into three areas:

(1) establishing the technical validity of the methods and supporting data described by the applicant, (2) verifying the existence of conservatism in the analyses, and (3) determining the degree of conservatism relative to traditionally accepted audit codes. The technical validity of the methods used in the Westinghouse PAb-3.3 code was established in Sections 3.1 and 3.2 of this report as well as earlier staff evaluations (e.g., Ref.12). The fact that the code is indeed conservative without the 65'F model uncertainty was also established in Section 3.2 of this report. The remaining item, a comparison of the PAD-3.3 predictions with a traditionally accepted audit code, is presented in this section c' the.4 port.

The GAPC0f!-THER!%L-2 code (Refs.15 and ?6) is one of a series of fuel thermal performance codes developed by Battelle Pacific Northwest 1.aboratories for the Core Performance Branch of the Nuclear Regulatory Comission. Since 1975, it has been used by the staff to aadit vendor fuel code submittals, including the Westinghouse PAD-3.3 code. GAPCON-2 predicts fuel temperatures, fuel densification and swelling, fission gap release and other fuel conditions as a function of time and power in a fashion much like that of PAD-3.3.

GAPCON-2 also has a number of conservative model options similar to PAD.

fn order to audit the proposed modifications to the PAD-3.3 code, a current version of the GAPCON-2 code was used to calculate (Ref.17) volume average fuel temperatures as a function of burnup for the Westinghouse 15x15 fuel design used in Zion Unit 1.

The results of these calculations are shown in Figure 1.

The two lines represent the best-estim:te and conservative preoictions for GAPCON-2.

Figure 2 of ifef.11 is similar, showing (1) the best estimate PAD-3.3 prediction, (2) the conservative PAD-3.3 prediction (3) the conservative i

c

9 WESTINGHOUSE 15X15 FUEL DESIGN N

I m

NRC EVALUATION MODEL J-)

E-4c:8 S: gr N

e E-m 8_

?

oM 4

o:

N

<l NRC DEST-ISTIMATE MODEL ANb "8

ea 0

1000 2000 3?'Ja ROD AVERACE BURNUP, MWD /MTU Figure 1

. ~---

. u.

4-.

3 7

fcr f 5-ir:tfen uncertair.tia,

........4-and (a) the conservative PAD-3.3 prediction plus the margin for fabrication uncertainties and the 65*F model uncertainty. The fourth Westinghouse curve (the highest) is representative of the current LOCA analysis. The third Westinghouse curve is the version requested by Comnonwealth Edison. All calculations (both GAPCON and PAD) were made at a local linear power of 15 kW/ft, approxi-nately the LOCA limit.

A number of conclusions can be drawn in examining these two figures.

First, the current (highest) Westinghouse curve is higher than all of the other curves, including the conservative GAPCON-2 prediction traditionally accepted for audit of LOCA analyses.

It is not until much higher burnups are reached that the Westinghouse prediction is exceeded by the conservative version of GAPCON-2. The change at higher exposures is due to the effect of cladding creepdown, which is considered in PAD-3.3, but not in GAPCON-2. The second ob-servation to be made from these figures is that all of the PAD predictions are higher than the best-estimate version of GAPCON-2.

This confirms our earlier conclusion that even the "best estimate" v'ersion of PAD-3.3 is not best-est'imate at all, but conservative.

There is additional evidence, not presented here, that even the best-estimate version of GAPCON-2 is, conservative with respect to the data.

In order to obtain a more representative comparison between these two figures, a second set of results were gene *ated in which cladding creepdown (which is already considered in PAD) was included in the GAPCON-2 predictions. The creepdown values used by GAPCON-2 were generated with the Zircaloy creep model from a second code called BUCKLE (Ref.18). The results are shown in Figure 2.

The revised results are very similar to these shown previously except both GAPCON calculations exhibit a significant decrease in fuel tem-peratures as ; function of burnup. The conservative versions of both GAPCON-2 and P.AD predict rising fuel ter.1pera'tures from e

m w p

-e-,

- - sp

WESTINGHOUSE 15X15 FUEL DESIGN I

k EdaD NRC EVALUATION MODEL WITil CREEPDOWN m 8-Ng N

a

~

E

~

N E-*

m S _.

9 oN 4

g N

,4 J

NRC DEST-ESTIMATE MODEL WITl! CREEPDOWN N

'D

~

-4_

l l

O 1000 2000 30f'O ROD AVERAGE BURNUP, MWD /MTU Figure 2 e

I

. eu.,..

.c...:,+c t,

10w burnup peak then falling terreratures.

It should be noted that the conservative GAPCON-2 and PAD predicted r.;,.i..;...; cccai at dif ferent burneps, but the maximum value calculatec' by the proposed version of PAD is not unreasonable compared to the conservative version of GAPCON.

It may also be noted the best-estimate vercions of both GAPCON-2 and PAD-3.3 predict high tem-peratures at beginning-of-life and monotonically decreasing tem-peratures thereafter. For the burnup range considered, the best-estimate version of GAPCON was continually and substantially over-predicted by the best-estimate version of PAD.

From our audit calculations of the PAD-3.3 code, we observe 'a similar, but not idsntical, behavior between this code and GAPCON-2. We also note that the proposed modifications to PAD result in peak volume average 'uel temperatures in reasonable agreement with that tradi-tionally accepted for LOCA analyses with a conservative version of the NRC audit code, GAPCON-2.

.l 7

1 1

l l

. 4

"' - e - i-

<;. ", - -' Ten:cre-ti:r The Appendix K requirements for fuel thermal performance codes state that:

"The steady-state te.mperature distribution and stored energy' in the fuel before the hypothetical accident shall be cal-culated for the burn-up that yields the highest calculated cladding temperature (or, optionally, the highest calculated stored energy.) To accomplish this, the thermal conductivity of the U0 shall be evaluated as a fun'ction of burn-up and 7

temperature, taking into consideration differences in initial density, and the thermal conductance of the gap between the UO, and the cladding shall be evaluated as a function of the buPn-up, taking into consideration fuel densification and expansion, the composition and pressure of the gases within the fuel rod, the initial cold gap dimension with its toler-ances, and cladding creep." (Ref.13)

There is no explicit requirement within this section of the Code for conservatism in the fuel performance codes.

4.1 Basis for Margin of Conservatism 8

Althouah there is no explicit requirement for conservatism in the calculation of the initial stored energy of the fuel, the Commission has expressed an opinion on the subject.

"The assumption of 102% of maximum power, highest allowed peaking factor, and highest estimated thermal resistance between the UO and the cladding provides a calculated 7

stored heat that is possible but unlikely to occur at the time of a hypothetical accident. While not necessarily a margin over the extreme condition, it represents at least an assumption that an. accident happens at a time which is not typical." (Ref.19)

This opinion establishes the requirement for conservatism in the calculation of initial fuel temperatures for Appendix K calculations.

However, the degree of conservatism for this heat source was never established. It is possible that even a best-estimate fuel code a

. predici conservative fuel temperatures in the LOCA analyses wsu.o because of conservatisms imbedded elsewhere in the calculation.

For the number of vendor fuel codes (including the Westinghouse PAD-3.3 code) that lave previously been approved by the staff, this is probably the case. They also exhibit various degrees of con-servatism by themselves, depending on the vendor and the type of calculation performed.

Sim lar behavior is 'xhibited by the fuel i

performance codes utilized by the NRC staff.

'hese are discussed in Sections 3.3 and 5.2 of this report.'

It may be noted, however, that a staff opinion has been developed for another heat source, the energy due to the decay of fission products.

"A best judgment eval iation of these factors leads to the conclusion that a suitable probability level is 95%.... A change to 99% or 99.9% would increase these margins but not substantially (i.e., not produce a fundamental change in the nature of the margins).

This level is viewed as the intent of the Appendix K rule development." (Ref. 20) and further that "An additional factor to be considered is the inter-action with criteria of other energy sources such as stored energy. Logically they too should be develorad with the same uncertainty probability levels as used for decay heat." (Ref. 20)

There are other examples of the application of a 95% probability level in the calculation of heat sources and other portions of the LOCA analysis. The choice of probability level appears to be more traditional than analytical.

O s

e

.. u re is no rigsc a (~;, licit bccis for a 95; icvci, however it appears to be a conservative level and is corpatible with other acceptance levels. The uncer-tainty and small varintion analysis is inbedded in a large LOCA matrix where, for example, F nominal seldom n

if ever occurs and other elements of required LOCA energy sources are undoubtedly conservative by some large but as yet undecided upon amount. Studies of power distributions for Westinghouse reactors as a function of reactor operation modes have indicated that F limit levels are at least 155., (and usually over g2%) greater than steady state operation nominal peaking factors, and F extremes of allowable load followtransientsareheached,ifatall,wellless than 5% of the time during a cycle for any presently envisioned operation." (Ref. 21) l These two staff opinions suggest the acceptability, without a rigorously derived basis, of the probability and confidence levels i

proposed. Such levels are usually submitted in safety analysis reports and judged acceptable by the staff. We are aware of no submittals in which a basis for the 95% probability level has been established by the industry. It is recognized, however, that the establishment of such a basis should involve a statistical analysis of the entire LOCA problem. Such a study does not, as yet, exist.

We are also aware of a step in the direction of determining the overall Appendix K conservatism. Westinghouse has proposed (Ref.

22) a statistical conbination of the uncertainties in the LOCA heat sources. The proposal was not accepted by the staff (Refs. 23 and 24). An appeal by Westinghouse (Ref. 25) resulted in a second rejection by the staff (Ref. 26) on the basis that modification to explicitly-required conservatisms (such as is the case for decay heat} in Appendix K analyses should be implemented by a change in the regulation itself. We conclude that a reduction in the margin of conservatism in Appendix K stored energy analysis cannot, at this time, be based on conservatisms in other portions of the LOCA analysis. Because there is no explicit requirement for margin of conservatism in the stored energy analysis itself, we also conclude e

. ti.at ti,e conservative margir. of this heat sou ce may be esta'iished s

through the review, rather than rulemaking, process.

Indeed, this has been the practice withir. the NRC in the past.

4.2 The Staff Criterion In order to develop a more uniform review of stored energy codes, wil intend to use the following criterion for these models:

Assuming best-estimate input conditions, an acceptable fuel performance model shall yield a required out-put parameter such that the predicted value bounds a large proportion of the experimental values for this parameter.

~

This means that an acceptable fuel performance code, given best-estimate input values, will, at a high probability level, correctly predict the peak fuel temperatures, fuel stored energy, fuel-cladding gap conductance or other parameter required as input to subsequent LOCA analysis codes. We believe an appropriately high p,robability level is 0.95 or 0.95/,0.95 where a confidence level is required.

[

i

. 5.

W icatico of the Criterion The principal required output parameter from the Westinghouse PAD-3.3 code is volume-average fuel temperature. Higher volume-average fuel temperatures are conservative for LOCA analysis. Therefore, to. meet the proposed acceptance criterion, the PAD-3.3 code should show the ability to overpredict volume-everage fuel temperatures 95% of the time at a 95%

confidence level based on experimental data.

This criterion should be established with experimental data prototypic of the Westinghouse product line and, where possible, taken near LOCA conditions.

As discussed in Section 3.2 of this report, Westinghouse believes the l

experimental fuel temperature data shown in Ref.11 to be representative of their standard product line. These data were taken at linear power levels of up to 15 kW/ft, which is approximately the LOCA limit. A limitation of these data, however, is that they are based on fuel centerline thermocouple measurements. Therefore, the data are an indication of fuel centerline tec.perature rather than volume average fuel temperature.

We are' not aware of any experimental data which directly measure in-reactor volume average fuel temperatures. It is possible, however, to relate fuel centerline and volume average temperatures analytically.

This is shown in Figure 3, where best-est'imate BOL fuel centerline and volume average fuel temperature predictions from GAPCON-2 are shown as a function of linear power. This figure indicates that the fuel centerline temperature rises much more rapidly than volume average temperature as a function of power. This is an expected result because fuel surface temperatures remain relatively close to the coolant temperature whereas the fuel centerline temperature rises. The volume average temperature may be approximated by the average of the fuel surface and centerline temperatures. Figure 4 shows the same prediction replotted with fuel volume average temceratures expressed as a function of fuel centerline temperature.

We will use this figure to relate uncertainty in centerline temperature to uncertainty in volume average temperature.

h h

e 4 __

\\

-a Z

e

~

e m

e b

'b M

d 5,

N Q

g s

_c N>

a s

w M

a D

5

'l w

j N

c

~

O M

A m

M o CC E

.e w

m-CO Z

p OX a

O o<

z;

~*o A

CO N

M O

1 i

e i

000t 000S 0002 000T 0

a 'aunlyHadisal 'Iand 'Ioa

%pma 4 NlC4Cm Pq2ANc<

Zs O

I t:

a s0 98 a8o y

c 0

50, 0

W FU1 E

0 S

E 0, L

T 0

I C

N E1 G

5 N 0, H

T0 O

E U

RL2 S

F 0

E I

0, ig N0 u

1 r

E e

5 4

T2 X

5 1

E 0, 5

M0 P

F E

e R3 U

0 E

A 0, 0

L TU D

R3 E

,E 5,

S 00 I

M/

F G

A N

4 R

0, G

0 0

I N

4 5

1'

!t 0

7, -

. 'ts~

1:; bc netd t!. t tne P 3-3.3 predictions of the experir.antal da t; a... a......ir.:.1 i r.p a t vci ut s. This is ir. Lespir.g with the criterion.

While w rc:ogni:c (Ref. 27) the conservatism which may exist in the input values to PAD-3.3 when used in LOCA analyses, such uncertainties are difficult to quantify from experimental data. As discussed in Section 4 of this report, we also believe it to be inappropriate to base the reduction of conservatism in one segment of the LOCA ana'ysis on possibly excessive conservatism in another. Suct a change is more appropriate to the rulemaking process in which the conservative margin in the overall problem may be examined. For'the purpose of this review, we shall assume that the PAD-3.3 code is provided nomical input conditions as part of the LOCA analysis.

5.1 The Westinghouse PAD Computer Code A statistical analysis was performed on the measured and predicted fuel centerline temperature data shown in Figure 2 of Reference 11.

We determined the mean and standard deviation ci the difference (e.g. measured temperature minus predicted temperature) assuming the distribution to be normal. As the proposed criterion requires the PAD-3.3 code to conservatively predict volume average fuel temperature with a 95'.' probability at a 95*.' confidence level, tne marain to be added to the best-estimate prediction should be 1.64 times the standard deviation plus the best-estimate code bias (data mean), if any.

To relate this uncertainty in fuel centerlir.e temperature to volume average fuel temperature, we refer to Figure 3.

At a LOCA limit of 15 kW/ft, we find best-estimate fuel volume average and fuel center-line temperature of 2200*F and 3600*F respectively. Adding the margin due to uncertainty in fuel centerline temperature to the best-estimate centerline temperature yields a conservative prediction of centerline temperature. From Figure 4, we find the correspondino best-estimate, conservative and equivalent margin values for volume average tempe'rature. Using this process, we conclude that the PAD-3.3 code meets the proposed criterion based on the experimental data comparison supplied by Westinghouse.

. 5.2 Thi '7 Fuci Ferfor arce Codes In order to check the validity of our conclusions regarding the ovarall uncertainty in the PAD-3.3 code, we reviewed a number of

~

other fuel perfomance codes and their predictive uncertainties.

We have attempted to show that the predictive uncertainty in the PA3-3.3 not only meets the proposed criterion, but is also repre-sentative of similar fuel performance codes.

5.2.1 FPAP-53 The FRAP-S3 code (Ref. 28) was developed by Idaho National Engineering Laboratory for the themal and mechanical analysis of light water reactor fuel rods. The code considers the effects of fuel and cladding deformation, temperature dis-tribution, internal gas' pressure, and material properties like PAD and GAPCON. FRAP-S3 was developed by Idaho National Engineering Laboratory for NRC's Office of Reactor Safety Research.

It is a representative example of a state-of-the-art fuel perforr.ance cod'e.

The FRAP-53 verification repo'rt (Ref. 29), presents predicted versus measured fuel centerline temperatures based on themo-couple measurements from approximately 100 rods, representing over 800 data points. All fuel rods used by Westinghouse, except WAPD-228 rods 22-3 and 22-4 were included in this study.

The standard error between measured and predicted fuel center-line temperature was stated to be 356'F and 457'F for un-pressurized and pressurized rods respectively. Assuming the standard deviation is independent of fuel centerline temperature (as was assumed in the study), this would result in a maximum u.1 certainty in fuel centerline temperature of 1,64 x 457'F = 750*F.

This is a.95/.95 statistical tolerance interval.

Tc elats this uncertainty in fuel centerline temperature to volume average fuel temperature, we again re'fer to Figure 3.

Adding _750*F margin to the best estimate centerline temperature

. ic!d: e ccr.:ervative centerline temperature prediction of

,-.C F.

From Fioure 4, we find that this 750 F marcin on centerline temperature is approximately equal to a margin cf

~

27F F on volume average fuel temperature. We conclude that FRAP-53 volume average fuel temperature predictions have a maximum uncertainty of 273*F at a.95/.95 tolerance level.

There is evidence (Ref. 30) that these values would drop considerably if the data base were restricted to helium pres-surized rods nc3r beginning-of-life with typical densities and gas sizes as proposed by Westinghcuse. We have not considered that possibility in this evaluati;n.

5.2.2 FRAPCON-1 The FRAPCON-1 code (Ref. 31) is a more recent version of the FRAP-53 code discussed previously. This computer program is the most recent of the fuel performance codes developed for the NRC. As is the case for its predecessor, FRAPCON-1 is intended to calculate the effects of power and burnup on fuel behavior under normal operating conditions.

The FRAPCON-1 verification report (Ref. 32) presents the results of predicted versus measured fuel centerline temceratures for approximately the same number (93 rods /740 data points) of fuel centerline thermocouple measurements as FRAP-S3. The standard deviation between measured and predicted values is 306*F for unpressurized rods and 529'F for the pressurized rod data. The latter value is larger than that calculated for FRAP-53 but no explanation for the regression in predictive Jbility is presented in the report.

The FRAPCON-1 assessment report again assumes that the standard deviation is constant for the range of centerline temperatures considered. Using the same process described for FRAP-S3, we calculated maximum FRAPCON-1 predictive uncertainties of 868'F for fuel centerline te:aperature and 319'F for volume average fuel temperature at a 95/95 tolerance limit.

-ci-l r -

The,APC0" series of comouter codes, which are also utilized r

by the NRC staff, have not been subjected to the same verifi-I cation process used for FRAP-53 and FRAPCON-1. All of the GAPCOM series codes have been verified with experimental data but the measured and predicted values have not been statis-tically analyzed. However, the developers of GAPCON, Battelle Pacific Northwest Laboratories, have attempted to establish the predictive uncertainty in these codes from first principles (Refs. 33 and 34).

A recent investigation by Cunningham et al. (Ref 35) deter-nined the effect of input and model uncertainties on fuel temperature and stored energy calculations. The study identified analytical models necessary for calculating stored energy and then utilized both the method of linear propagation and Monte Carlo technique to determine prediction uncertainties. Re-sults were generated for a typical BWR fuel rod, but the study is also applicable to PWR fuel designs. The authors estimate the maximum uncertainties for fuel centerline temperature at a linear power of 500 W/cm (15.2 kW/ft) to be 15.5% for the lionte Carlo technique and 18.2% for the linear propagation method.

These figures are given at a 3o (99.9%) confidence level, but will be assumed.95/.95 tolerance intervals in tr.is report.

Using these figures, we concluded that the expected uncertainty in predicting the fuel centerline temperature of a PWR rod operating at 15 kW/ft to be 553 F and 655*F by two first-principles methods. These values correspond to 206*F and 239*F uncertainties on the volume average fuel temperatures.

t a

. 5.3

.; '.*.::t of the Art We have examined a san.ple of state-of-the-art fuel performance codes for an indication of the expected uncertainty ir predicting volume average fuel temperatures at approximately 15 kW/ft. This sample included two similar data-prediction studies (Refs. 29 and

32) and two similar first principles methods (Ref. 35). The maximum uncertainties assumed in this sample are summari:ed below:

Maximum uncertainty in volume average Study fuel temperature at 15 kW/ft Data-prediction (Ref. 29) 273*F Data-prediction (Ref. 32) 319'F Linear propagation (Ref. 35) 206*F Monte Carlo (Ref. 35) 239*F Average 259"F To determine the state-of-the-art uncertainty in volume averace fuel temperature, we have taken the average of these values. The value obtained is 259'F above the data mean. The Westinghouse margin for volume average fuel temperature is also the difference between the evaluation model prediction and the data mean. This value is the maximum volume average temperature predicted by the evaluation model version of the code minus the maximum volume average tem-perature predicted by the best estimate version plus the bias in the best estimate code, if any. We have determined this value and conclude that the PAD-3.3 meets the proposed criterion without the use of the 65'F model uncertainty and that the remaining margin of conservatisn is similar to the expected uncertainty in other state-of-the-art fuel codes.

j i

1

-;7-E.

.a Ue have ev= mined the proposed revisions to the Westinghouse fuel code, PAD-3.3, as described by the code developer and Commonwealth Edison Company. These changes consist of the use of as-fabricated, rather than as-designed, values of fuel density, fuel sintering temperature a'nd their asseciated tolerances, and the deletion of the 65'F nodel uncertainty from the Westinghouse fuel thermal performance analysis. Based uoon our technical review of the submittal, comparison of the Westinghouse code predictions with a traditional staff audit code, and the development and application of a new criterion for margin of conservatism in codes of this type, we conclude that these changes are acceptable. This acceptance is limited to the current ion of the Westinghouse PAD-3.3 code as approved by the staff (Ref.14) for application in LOCA analyses, but acceptance is not limited to Zion.

M

e

-2 'a -

7.

Reic sn;us 1.

C. Reed (Com-Ed) letter to H. R. Denton (NRC) on " Zion Station Units 1 and 2 Proposed Change to Facility Operating License Nos.

DPR-39 and DPR-48," dated March 22, 1979.

2.

Title 10, Code of Federal Reculations, Part 50, Section 46,.

" Acceptance Criieria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," January 1979.

3.

C. Reed (Com-Ed) letter to H. R. Denton-(NRC) on " Zion Station Unit i Base Load Operation Under Revised ECCS Evaluation" dated January 24, 1979.

4.

C. Reed (Com-Ed) letter to H. R. Denton (NRC) on " Zion Station Unit 1 Additional Information for Base Load Operation Under Revised ECCS Evaluation" dated Janaury 25, 1979.

5.,

C. Reed (Com-Ed) letter to H. R. Denton (NRC) on " Zion Station Units 1 and 2 Revised ECCS Evaluation" dated January 29, 1979.

6.

C. Reed (Cca-Ed) letter to D. G. Eisenhut (NRC) on " Zion Station Units 1 and 2 LOCA Reanalysis Review" dated July 16, 1979.

7.

J. V. Miller, Ed., " Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations," Westinghouse Electric Corporation Report WCAP-8720, October 1975 (proprietary) and WCAP-8785, October 1976 (non-proprietary).

8.

P. S. Check (NRC) memorandum to A. Schwencer (NRC) on " Questions Concerning Removal of Conservatism from the Westinghouse PAD Code" dated April 18, 1979.

9.

P. S. Check (NRC) memorandum to A. Schwencer (NRC) on " Zion Request for Technical Specification Change to F Based on Conservatism in 0

the PAD Computer Code" dated August 30, 1979. Enclosure reprinted as Appendix A to Ref 11.

10.

W. F. Naughton (Com-Ed) letter to H. R. Denton (NRC) on " Zion Station Units 1 and 2 Additional Information for Increase in F Peaking Factor" dated May 3,1979, q

11.

W. F. Naughton (Com-Ed) letter to H. R. Denton (NRC) on " Zion Station Units 1 and 2 Additional Information for Increase in F Peaking Factor, NRC Docket Nos. 50-295 and 50 ~04" dated n

JXnuary 25,1980 (proprietary).'

12.

J. F. Stolz (NRC) letter to T. ii. Anderson (W) on " Safety Evalu-ation of WCAPt872'" dated February 9,1979.

9 e

e w

3

+&r>

I

. A 4+

o.

,, kff Y ;;l 4

5. +Q pV' I& jhhfk R/ l/

%l-l}

's 4 V

IMAGE EVALUATION NN TEST TARGET (MT-3) 1.0 Ta m jf $ IIM4 b

-9 "llll 2.0 llu===

l,l 3

i l.25 1.4 1.6 s-

=

MICROCOPY RESOLUTION TEST CHART

+ tsp 4%

Af! >///h,

bQffb 5,,,,, 7.

>gy /

<y y,..

?

-r.

o e..

.w

d er 13. Titie 10, Code c_f Federal Reculatiens, Part 50, Appendix K, "ECCS Evaluation l'odels" January 1979.

14. R Salvatori (W) letter NS-RS-133 to D. B. Vassallo (AEC) with Suppleeental Information on Fuel Densification dated February 5, 197 '.. Also Appendix C.1 of " Fuel Densification Experimental Results and Model for Reactor Application," J. M. Hellman, Ed.,

Westinghouse Electric Corporation Report WCAP-8218-P-A, fiarch 1975 (proprietary) and WCAP-8219-A, !! arch 1975 (non-proprietary).

15.

C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko, and L. J.

Parchen, "GAPCON-THERitAL-2: A Computer

• Program for Calculating the Thermal Behavior of an Oxide Fuel Rod," Battelle Pacific Northwest Laboratory Report BNWL-1898, Novem!ser 1975.

16.

C. E. Beyer, C. R. Hann, D. D. Lanning, F. E. Panisko, and L. J.

Parchen, " User's Guide for GAPCON-THERMAL-2: A Computer Program for Calculating the Thermal Behavior of an Oxide Fuel Rod,"

Battelle Pacific Northwest Laboratories Report BNWL-1897, November 1975.

17.

K. Kniel (NRC) memorandum to S. Fabic (NRC) on "GAPCON Input Parameters for the WRAP Program" dated Septembe. ?7,1979.

18.

P. J. Pankaskie, "An Analytical Computer Code for Calculating Creep Buckling of an Initially Oval Tube," Battelle Pacific fWthwest Laboratories Report BNWL-1784, ! day 1974.

19. Opinion of the Comnission, "In the Matter of Rulemaking Hearing, Acceptance Criteria for Emergency Core Cooling Systems for Light Water-Cooled-Nuclear Powr Reactors," USAEC Docket No. Rf1-50-1, December 28, 1973.

20.

H. J. Richings (NRC) memorandum to P. S. Check (NRC) on " Potential i

Chances in CPB Position on Decay Heat for LOCA" dated July 14, 1976.

21.

H. Richings (NRC) memorandum to P. S. Check (NRC) on " Policy Questions Arising)from W Maxi Convoluttoa" dated April 6, 1978 (Proprietary.

22.

C. C. Little, S. D. I:opelic and H. Chelemer, " Consideration of Uncertainties in the Specification of Core Nt Channel Factor Limits," Westinghouse Electric Corporatior Report WCAP-9180, September 1977.

23.

D. F. Poss, Jr. (NRC) memorandum to R. J. Mattson (NRC) on "A recoosed Change to LOCA Peaking Factor Uncertainty Require-ments" dated July 24, 1978.

. 24 R. J. "attson (NRC) letter to T. Anderson (W} dated August 9, 197S.

25.

T. M. Anderson (W) letter NS-TMA-1929 to R. J. Mattson (NRC) dated September 5,1978.

26.

R. J. Mattson (f!RC) letter to T. Anderson (W) dated October 26, 1973.

27.

H. Richings (NRC) memorandum to P. S. Check (NRC) on "Some Notes on PWR (W) Power Distribution Probabilities for LOCA Probabilistic Analyses' dated July 5,1977.

28.

J. A.

Dearien,

G. A. Berna, M. P. Bohn, J. D. Kerrigan and D. R.

Coleman, "FRAP-53: A Computer Code for the Steady-State Analysis of Oxide Fuel Rods, Volume 1, FRAP-S3 Analytical Models and Input flanual," EG&G Idaho, Inc. Report TFBP-TR-164, March 1978.

29.

D. R. Coleman, E. T. Laats and N. R. Scofield, "FRAP-53: A Computer Code for the Steady-State Analysis of Fuel Rods, Vclume 2, Model Verification Report, EG&G Idaho, Inc. Report TFBP-TR-228, August,1977.

30.

J. D. Kerri.gan and D. R. Coleman, " Standard Design Analysis - A Statistical Determination of Corewide Initial Accident Conditions -

Fuel Stored Energy Results," Table X, EGAG Idaho, Inc. Report CAAP-TR-034, December 1978.

31.

G'. A. Berna, M. P. Bohn, D. R. Coleman and D. D. Lanning, "FRAPCON-1:

A Computer Code for the Steady-State Analysis of 0xide Fuel Rods,"

EGSG Idaho, Inc. Report CDAP-TR-78-032, August 1978.

32.

E. T. Laats, G. B. Peeler and N. S. Scofield, " Independent Assessment of the Steady State Fuel Rod Analysis Code FRAPCON-1," EG&G Idaho, Inc. Report CAAP-TR-050, May 1979.

33.

D. D. Lanning, C. R. Hann and E. S. Gilbert, " Statistical Analysis of Gap Conductance Data for Reactor Fuel Rods Containing U0 Pellets,"

Battelle Pacific Northwest Laboratories Report BNWL-1832, Abaust 1974.

34 C, R. Hann, D. D. Lanning, R. K. Marshall, A. R. Olsen and R. E.

Williford, "A Method for Determining the Uncertainty of Gap Con-ductance Deduced from Measured Fuel Centerline Temperatures,"

Battelle Pacific Northwest Laboratories Report BNWL-2091, February 1977.

35.

M. F. Cunninnham, D. D. Lanning, A. R. Olsen, R. E. Williford and C. R. Hann, " Stored Energy Calculation: The State of the Art,"

Battelle Paci,fic Northwest Laboratories Report PNL-2581, May 1978.

e

v 7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION 00CXET N05. 50-295 AND 50-304 COMMONWEALTH EDISON COM'ANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Comission) has issued Amendment No. 53 to Facility Operating License No. DPR-39, and Amendment No. 50 to Facility Operating License No. DPR-48 issued to the Comonwealth Edison Company (the licensee), which revised Technical Specifications for operation of Zion Station, Units 1 and 2 (the facilities) located in Zion, Illinois. The amendments are effective as of the date of issuance.

The amendments, modify the Technical Specifications, Appendix A to the licenses, to increase the allowable LOCA peaking factor from 1.86 to 1.93 based on an ECCS reanalysis.

The application for the amendments conplies with the standards ant equirements of the Atomic Energy Act of 1954, as amended (the Act), and the Conmission's nsles and regulations. The Comission has made appropriate findings as required by the Act and the Cortnission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments. Notice of Proposed Issuance of Amendment to Facility Operating License in connection with this action was published in the FEDERAL REGISTER' on May 7, 1979 (44 FR 26816).

No request for a hearing or petition for leave to intervene was filed fellowing notice of the proposed action.

l 300 9'D 8'06 / 3

7590-01 1 '

The Comission has determined that the issuance of these amendments will not result in any signf ticant environmental impact and that pursuant to 10 CFR 51.5(d)(4) an environmental impact statement or negative declaration and environmental i pact appraisal need not be prepared in connection with issuance of these amendments.

For further details with respect to this action, see (1) the application for amendments dated March 22, 1979, as supplemented on May 3,1979 and January 25,1980,(2) Amendment Nos. 53 and 50 to License Nos. DPR-38 and DPR-49, and (3) the Comission's related Safety Evaluation. All of these items are available for public inspection at the Comission's Public Document Room,1717 H Street, N.W., Washington, D.C. and at the Zion-Benton Public Library District, 2600 Emmaus Avenue, Zion, Illinois 60099. A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Comission, Washington, D.C.

20555, Attention: Director, Division of Operating Reactors.

Dated at Bethesda, Madland, this 13th day of March,1980.

FOR THE NUCLEAR REGULATORY COMMISSION l#$1 A. Schwencer, Chief 1

Operating Reactors Branch #1 Division of Operating Reactors

_,y

-,