ML19322B470

Safety Evaluation of B&W ECCS Evaluation Model

ML19322B470
Person / Time
Site:	Oconee
Issue date:	12/27/1974
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
References
NUDOCS 7912030308
Download: ML19322B470 (36)

Text

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7

. r SAFETY EVALUATION REPORT BY THE DIRECTORATE OF LICENSING 4

U.S. ATOMIC ENERGY CCMMISSION IN THE MATTER OF DUKE POWER COMPANY-3 OCONEE NUCLEAR STATION UNITS 1, 2, AND 3 9{,-270,-287 DOCKET NO DEC 271p4 e...

1912080 3o 8

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e TABLE OF CONTENTS Pave

1.0 INTRODUCTION

1 2.0 BABCOCK AND WILCOX ECCS EV ALUATION MODEL...............

5 30 APPLICABILITY OF GENERIC EVALUATION MCDEL..............

8 4.0 RESULTS OF LCCA CALCULATIONS...........................

9 50 CONCLUSIONS............................................

16

6.0 REFERENCES

20 APPENDIX A - OPERATING RESTRICTIONS APPENDIX B - LETTER FROM ADVISORY COMMITTEE ON REACTOR SAFEGUARDS, NOVEMBER 20, 1974 t

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LIST OF TABLES Page TABLE 1.

A COMPARISON CF CCONEE UNITS TO KEY PARAMETERS EMPLOYED IN THE GENERIC EVALUATION MODEL........

18 TABLE 2.

SUK4 ARY OF SENSITIVITY STUDIES....................

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LIST C5 FIGURES Page FIGURE A-1.

Control Rod Group Withdrawal Limits for 4-Pump Operation,-Units 2 and 3, Up to 100 Full Power Days..................................

A-3 FIGURE A-2.

Control Rod Group Withdrawal Limits for 4-Pump Operation--Units 2 and 3,100-435 Full Power Days........................................

A-4

? FIGURE A-3 Control Rod Group Withdrawal Limits for 4-Pump Operation--Units 2 and 3, After 435 Full Power Days..................................

'A-5 FIGURE A-4.

Control Rod Group Withdrawal Limits for 4-Pump Operation--Unit 1, Cycle 2, Up to 250 Full Power Days..................................

A-6 FIGURE A-5 Cperational Power Inbalance Envelope--Unit 1, Cycle 2.....................................

A-7 FIGURE A-6.

LOCA, Limited Maximum Allowable Linear Heat Rate--Units 1, 2, and 3.....................

A-8 O

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1..

1.0 INTRODUCTION

On January 4,1974, the Commission published its acceptance criteria for emergency core cooling systems for light water power reactors (39 FR 1003).

This rule included Appendix K to 10 CFR 50 which specifies analytical techniques to be employed for the evaluation of ECCS effectiveness..Cn August 5, 1974, Babcock and Wilcox officially submitted a five volume package (1,2,3,4,5) of topical reports constituting their proposed ECCS evaluation model.

The information contained in these reports had been the subject of a number of informal conferences and discussions between the staff and Babcock and.Wilcox, starting just prior to the publication of the Acceptance Criteria in January, 1974. The Regulatory staff reviewed these documents and published (6) a Status Report on October 15, 1974, which addressed each item required by Appendix K and identified areas which were acceptable to the staff and areas of concern which were to be resolved.

On November 13, 1974, the Regulatory staff published a Supplement (7) to the Status Report which addressed each of these areas of concern.

As reflected in the Supplement, for some items adequate additional information was provided to enable the staff to accept

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a the Babcock and Wilcox approach.

For certain other items, the staff I

concluded that adequate justification had not been provided and that further modification of the August 5,1974 model was required.

Babcock and Wilcox will codify their model to reflect these staff requirements and has evaluated the effect of all changes upon 4

(10) the previous calculations.

Accordingly, the Babcock and.Wilcox model with the modifications presented in Section 2.0 and 4.0 of this SER is acceptable and would confor= to Appendix K.

A report of the Advisory Committee on Reactor Safeguards, attached as Appendix B, was issued en November 20, 1974 regarding the generic review and the acceptability of the Babcock and Wilcox ECCS Evaluation Model.

On August 5,1974, Duke Power. Company (the licensee) submitted an analysis of ECCS performance for the Oconee Units 1, 2, and 3 (the plant or facility) along with proposed Technical Specification (8) changes to reflect the new ECCS evaluation model calculations.

On September 20,~1974, Duke Power'tompany also submitted an (9) analysis of ECCS performance for Oconee Unit 1, Cycle 2.

These evaluations were based upon the Babcock and Wilcox August 5,1974 L

Evaluation Model..Section 3 0 of this SER discusses the applicability.

of the generic evaluation model to the specific Oconee Units 1, O

~ 2, and 3 plants.

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As stated in the Status Report and its Supplement, the August 5th Babcock and Wilcox Evaluation Model was not completely acceptable and specific model changes noted in the Status Report and its Supplement were required.

These changes are now being made to the generic Babcock and.Wilcox evaluation model.

Since the Oconee Units 1, 2 and 3 evaluations were based upon a model which was not acceptable, they also require some changes.

A revised set of computations for the plant (and for other facilities in a like position), using the newly revised and acceptable evaluation model, cannot be submitted for a number of months.

To determine the effect of the model changes made to the August 5,1974 Babcock and Wilcox Evaluation Model, the staff requested, and Babcock and Wilcox submitted, a series of. generic plant sensitivity studies which quantified the effect of the (10) model. changes. on the resu.lts of the previous calculaticns.

The staff followed the performance of these sensitivity studies while they were in progress and has reviewed the results.

These results are presented. in Section 4.0 along witn a discussion of.

their effects on the evaluation submitted for Oconee Units 1, (8,9) 2, and 3 From these' studies, it appears that certain operating

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restrictions are requiredito ensure that in the event'o( a

postulated loss-of-coolant accident, ECCS cooling performance will not exceed the values for calculated peak clad temperature and oxidation and hydrogen generation limits set forth in 10 CFR 50.46(b). Appendix A of this SER presents the acceptable operating-limitations. Although these restrictions were established on the basis of applicable generic sensitivity studies of the effect of moda? changes, the staff believes that in conformity with the requirements of 10 CFR 50.46, these restrictions should be verified by a re-analysis based upon the Sabcock and Wilcox Evaluation Model, as corrected. An evaluation of ECCS performance, wholly in conformity with 10 CFR 50.46 and Appendix K, and based on an approved evaluation model should be submitted for this facility, along with proposed Technical Specifications based upon such an evaluation, as promptly as it can reasonably be performed.

During the interim, before an evaluation wholly in conformity with the requirements of 10 CFR 50.46 can be submitted and evaluated, continued conformance to the requirements of the Commission's Interim Acceptance Criteria (IAC) and the restrictions contained (8,9) in-the licensee's August 5,1974 and September 20, 1974 submittals, combined with the additional limitations set forth in Appendix A hereto, will-provide' reasonable assurance that'the public health

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and safety will-not be endanEered.

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• 2.0 BABC0CK AND WILCOX ECCS EVALUATION MODEL (6)

The staff Status Report provides a complete evaluation of a

the Babcock and Wilcox ECCS Evaluation Model.

Each part-of 10 CFR 50,* Appendix K was addressed and appropriate comments regarding compliance to each aspect of the model were included.

All phases of the Babcock and Wilcox analytical methods were concluded to meet Appendix. K requirements with the exceptiens noted in Supplement (7) 1 of the Status Report.

Of the fourteen areas of concern addressed in Supplement 1 to the Status Report, five were identified as model deficiencies for Oconee Class reactors (177 fuel assembly plants with a lowered loop arrangement) requiring modification or additional data to justify conformance to Appendix K.

These areas are briefly discussed below.

Additional detail of each

• deficiency is presented in Section 4.0 of this SER and in the staff

( 6,7 )

1 Status Report.

f e A complete -listing of each computer program, in the same form as used in the evaluation codel, was furnished to the Regulatory _

staff.

These listings, combined with the Babcock ad Wilcox. impact (10)

studies, constitute the currently acceptable ECCS codel.

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2.1 Metal-Water Reaction The staff required that the Babcock and Wilcox ECCS model be revised to account for thinning of the oxide layer on the inside and outside of the fuel cladding.

In addition, an improved calculational technique for arriving at a predicted value for total core-wide metal-water reaction resulted from staff comments.

Babcock and Wilcox is modifying its ECCS model to incorporate these features.

See Section 4.0 for an assessment of impact upon the current plant operating restrictions, 2.2 Swelling and Rupture of the Claddine and Fuel Red Thermal 'Paraneters (6)

(7)

As noted in the Status Report and Supplement 1, the staff accepted the Babcock and Wilcox.modeling of swelling and rupture with three limitations.

As discussed in Section 4.0 of 'this SER, these limitations were satisfied in the Oconee evaluations.

Babcock and Wilcox has proposed to modify its model to incorporate a plastic swelling model, discussed in the Status Report Supplement, and a transient pin pressure codel, which would eliminate two of the staff limitations.

These modifications have not yet been completed.

At present, the existing swelling and rupt' re model is acceptable if the staff 11m'itations are

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23 End-o f-Blowdown As indicated in the Status Report and Supplement 1, the staff -

accepted the modeling of end-of-blowdown with the conditions that the definition of end-of-bypass be changed and that the down-comer noding representation be changed to use a homogeneous noding.

Babcock and Wilcox is modifying its codel to incorporate these changes. Section 4.0 discusses the impact of these deficiencies on the licensee's calculations.

2.4 Containment Pressure (6)

Page 4-41 of the Status Report-states that the containment backpressure calculation performed for the Oconee Class plants is conservative and acceptable. For plants of a different type, specific input assumptions must be justified on an individual plant basis.

Although the backpressure codel is acceptable for the Oconee Class plants, the effect of the use of the conservatively assumed parameters should be assessed by comparison with actual as-built values. Accordingly, the licensee has been requested to provide i

as-built values and.to discuss the methods used to determine the passive containment heat sinks for the Oconee Units.

Also required is an identification of each sink by category. (e.;., cable tray equipment supports, floor grating, crane wall) and surface area,

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- thickness, materials of construction, thermal conductivity and volumetric heat capacity by component category.

Values of paint thickness, thermal con'ductivity and volumetric heat capacity for

-- containment' internal structures are also requested.

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2.5 Steam Interaction with Emergency Core Cooling Water in Pressurized Water Reactors Two concerns discussed by the staff in Supplement 1 to the (7)

Status Report are related to the effect of hot walls on the ECC water being injected in the downcocer and the appropriateness of the value used for vent valve resistance. Babcock and Wilcox will' modify their model to incorporate the resolution of these concerns. Section 4.0 assesses the impact of these concerns upon the plant operating restrictions.

30 APPLICABILITY OF_GENEFIC EVALUATION MODEL (1)

As noted in BAW-10091 and in the staff's Status (6)

Report, the development of the generic Babcock and Wilecx i

Evaluation Model involved the utilization of a plant design appropriate to all Oconee Class reactors.

The series of sensiti-vity studies described in BAW-10091, Section 5 0 were therefore directly applicable to Oconee Units 1, 2, and 3. _ Also worthy of note are the actual key parameters utilized in the generic model calculations. Babcock and Wilcox stated that they bounded the variations in key parameters within the Oconee Class plants by choosing values in. their generic calculations which conservatively include any plant-to-plant variations. ' Table 1 provides a list of such key parameters employed in the generic evaluation and e

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1 compares each parameter to the actual values for Cconee Units 1, 2, and 3 This list shows that the generic calculation sufficiently incorporated the differences in these key parameters found in this plant.

4.0 RESULTS CF LOCA CALCULATICNS From a break spectrum analysis, the worst break examined by 2

Babcock and Wilcox using the August 5,1974 model was an 8.55 ft (1) double-ended rupture at the reactor coolant pump discharge.

This (8.9) generic analysis was the basis for the licensee's submittals.

o This calculation resulted in a peak clad temperature of-2062 F, 3 38% local metal-water reaction, and 0.14% whole core metal-water reaction.

These values are within the criteria of 10 CFR.50.46 o

(2200 F, 17%, and 1%, respectively).

4 All of the model deficiencies noted in Section 2.0 of this l

SER were examiIne'd by Babcock and Wilcox with regard t'o (10) an, impact assessment on current' operating reactors.

The j

-following secti ns address each or the relevant model deficiencies and their effects on the August 5,1974 LOCA analysis.

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4.1 Metal-Water Reaction As indicated in Section 2.1 of this SER, the staff has requested that the Babcock and Wilcox ECCS Evaluation Model be revised to account for thinning of the oxide layer on the inside and outside of the fuel cladding.

The generic model LOCA limit (1) calculations assumed initial values of 0.0001 inches oxide layer thickness and 1800 psia internal pin pressure.

An oxide (7) thickness sensitivity study conducted by Babcock and Wilcox-yielded the conclusion that the value of internal pin pressure combined with the value of the oxide thickness used by Babcock and Wilcox in their generic calculations conservatively predicted the highest peak cladding temperature for fuel cycle 1 operations.

The initial oxide thickness of 0.0001 inches and a pin pressure of 1800 psia are also appropriate for some reactors (Oco' nee Unit 1) through fuel cycle 2' operations.

The Babcock and Wilcox study thinned the oxide layers consistent with the degree of pin-swelling predicted.

The applicability of these assumptions to'Oconee Unit 1, cycle 2, results from significantly lower pin pressures calculated to~ occur..

The staff also noted in the Status Report that furthec justificatien.was _ required to support the Babcoc'k and Wilecx

' calculational technique for predicting total ~ccre-wide metal-water reaction.

In the Supplement, the staff reported that e

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Babcock and Wilcox had chosen to modify their model in a manner which the staff found would be adequate.

These modifications are now being made to the Babcock and Wilcox model.

To determine whether this modification would affect the calculations submitted by the licensee, the staff considered sensitivity studies performed using staff models previously developea for confirmation of analyses submitted under the IAC.

Although these models do not fully incorporate all required evaluation features, they are adequate to demonstrate that the results will fall well within the hydrogen generation criteria of 10 CFR 50.46(b)(3).

Therefore, this modification has no impact on the licensee's calculations.

4.2 Swelling and Ruoture of the Cladding and Fuel Rod Thernal Parameters (6)

As noted in the Status Report, the staff accepted the generic calculation if three limitations were observed:

a)

The internal pin pressure selected for the initial condition value must exceed the maximum predicted during nor=al operation for the design being analyzed.

b)

If the rod with the highest - peak clad temperature ruptures, then the time of rupture is restricted to' a time period prior to the end of blowdown.

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s c) 70% circumferential swelling for certain rupture tempera-tures must be employed.

It is permissible to increase o

linearly from 1700 F (about 40% circumferential swelling) o to the 70% plateau at 2000 F.

'The Oconee analyses satisfy each of these-limitations (maximum pin pressure was assumed, ruptures occurred price to o

end of blowdown, and rupture temperatures were less than 1700 F).

Accordingly, there is no impact en the licensee's calculation.

43 End-o f-Biowdown (1)

Since the generic calculation showed that end-of-bypass always occurred prior to, or at the same time as, end-of-blowdown, the model change regarding the definition of end-of-bypass has no effect on peak clad temperatures for this plant.

With regard to the staff concern that the downcomer model did not appear to be properly represented, Babcock and Wilcox has now changed the downcomer noding to a homogeneous noding representation as required in the Status Report Supplement.

In ' connection with this change, a number of other areas previously modeled on a heterogeneous basis have also been changed to homogeneous noding.

This is acceptable.

These modifications will require related changes to the generic model sensitivity

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' studies. These are being performed by Babcock and Wilcox.

However, in assessing the impact of this required change upon the calculations made using the August 5th model, Babcock and Wilcox found that two counteracting phenomena occur to result in an overall decrease in peak clad temperature at the 6-foot elevation of about 0

80 F.

Although less water remains in the vessel at the end-of-bypass (leading to a longer adiabatic heatup), reduced water head in the downcomer allows a significantly higher negative flow

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through the core for a longer period of time.

As previously indicated, the overall effect is to decrease the peak clad temperature, especially at the higher core elevations.

4.4 Containment Pressure For the reasons stated in Section 2.4 of this SER, ' staff concerns in the area of the containment backpressure calculation have no-effect en the licensee's calculations.

45 Steam Interaction with Emergency Core Cooline Water in Pressurized Water Reactors

.(7)

As noted in Supplement 1 to the Status Report, the staff required' that Babcock and Wilcox correct the-vent valve resistance (K) for two-phase flow by applying a factor of 1.5 to the single phase value. With respect to the vent valve flow resistsnce J

-14 factor used by Babcock and,Wilcox (K = 3 9), the staff requir_ed correction of this factor for two-phase flow.

As indicated in the Supplement, a correction factor of C = 1.5 based upon appropriate experimental data for gate valves was proper along with a further correction to account for the pressure dependence of C. In the Reynolds number range of interest during reflood (starting with a reference K of 3 3 based on single-phase data),

a multiplier of 0.85 is acceptable to correct for pressure effects.

Therefore, the required vent valve K-factor to be used in reflood calculations is:

K = 3 3 x 1.5 x 0.85 = 4.2 Babcock and Wilcox will modify its model to use this value. Various sensitivity studies were performed by Babcock and Wilcox to assess the i= pact of this change of assumed vent valve K.

The results of these studies showed that an increase in vent valve (1) resistance from the value of 3 9 used in the generic calculations o

to 4.2 showed about a 20 F increase in peak clad temperature.

With regard to the effect of hot walls on the ECC water being injected in the downcomer, the staff has provided Babcock and Wil-(6) cox a description of an acceptable hot wall ti.me delay model.

During the hot wall delay period, ECC water,. which is delayed in passing through the downcomer, accumulates in available storage volumes in the following manner:

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1) Lower downcomer - regicn between the bottom of the downcomer and tne lower lip cf'the cold leg.

A maximum of 1/3 of this volume will become available linearly over the hot wall delay period.

2) Upper downcocer - region of downcomer above the lower lip of cold leg pipe.

If the lower downcomer volume cannot handle all accumulator ECC water, some water will spill out the break.

A storage volume is available in the upper downcomer which is determined by the elevation head above the bottom of the cold leg.

The same elevat' ion head should be used to determine the break flow rate.

3)

Cold les piping from the reactor coolant pump discharge to the vessel nozzle. A storage' volume consistent with the upper downcomer water level is available.

Once the hot wall delay time has elasped and flow through the down. comer begins, a 'further period of time is required for the ECC water to flow from the available storage volumes to the lower plenum.

To reflect this period, a downcomer transport (free fall) delay time is calculated which is added to the hot wall delay time to yield the total ti=e required for ECC water to fall from the inlet elevation to the bottom of the downcomer (lower plenum).

Once the hot wall delay time is ended and free fall starts, no further spillage of ECC water out the break would occur.

Babcock

1 and Wilcox has indicated that sufficient stora5e capacity exists to account for the volume of water which could be -accumulated during the hot wall delay time. Therefore, there is no net change in the generic calculation due to hot wall effects.

4.6 Summary or Results review of preceding Sections 4.1 through 4.5 shows that

'A the two model deficiencies which have an impact on the previous generic calculations are region noding (Section 4 3) and vent valve K-factor (Section 4.5). Table 2 shows a summary of the results of sensitivity studies by Babccck and Wilcox on peak clad temperature, local metal-water reaction, and whole core metal-water reaction.

These calculations indicate that, while the model corrections could cause an increase in peak clad temperature, this increase would not be large enotSh to exceed the criteria of 10 CFR 50.46, provided that the LOCA limit curves submitted in the licensee's proposed Technical Specifications are observed in facility operation.

These curves are set forth in Appendix A along with additional restric-tions for Unit 1 required to assure conformity with the appropriate limitations.

5.0 CONCLUSION

S Based on the analysis set forth in this Safety Evaluation, the limitati31s contained 'in the licensee's submittals, particularly the LOCA limit curve set forth in Appendix A, along with the addi-ditional restrictions on Unit 1 operation,;will assure conformance with the peak clad temperature limit, and ' maximum oxidation and

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These hydrogen generation criteria of 10 CFR 50.46(b).

restrictions should be verified by a re-analysis based on the Babcock and Wilcox Evaluation Model, modified as described in this Safety Evaluation Report.

In addition, the Oconee Units satisfy the two remaining (6) and

criteria, i.e., maintes. nce of a coolable geometry (11) long-term cooling.

The heat removal system for long term cooling of the plant as described in the FSAR is satisfactory for these requirements.

An evaluation of ECCS performance wholly in conformance with 10 CFR 50.46 and Appendix K, based on an approved evaluation todel* should be submitted for this facility as soon as practicable, but within six months or before any refueling is authorized.

During the interim, until each evaluation is submitted and evaluated by the staff, operation should conform to the requirements of the Interim Acceptance Criteria

, as well as to the (8,9) requirements of the licensee's submittals and the require-ments of Appendix A.

'The Babcock and Wilcox ECCS ' Evaluation bbdel, which is wholly in conformance with Appendix K of 10 CFR 50.46, is described in a (10)

' letter from Babcock and Wilcox dated December 18, 1974.

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TABLE 1 A COMPARISON OF GCONEE UNITS TO KEY PARAMETERS EMPLOYED IN THE GENERIC EVALUATION MODEL PARAMETER GENERIC MODEL UNIT 1 UNIT 2 UNIT 3 Rated Core Power, Mwt 2,772 2,568 2,568 2,568 Reactor Vessel Flow, (2)

(1) lbm/sec 38,306 39,740 40,137 40,834 Reactor Coolant System Pressure.at Core Outlet, psig 2,182 2,185 2,185 2,185 Core Inlet Fluid (2) o Temperat ure, F 556 556

$56.5 556.5 Volume Average Fuel Temperature at 18 Kw/ft with a Sink Temperature o

o of 580 F, F

3,105 3,050 3,105 2,980 ECCS Delay Time, seconds-35 25 25 25 Reactor Building Free 3

6 6

6 6

Volume ft 2.205x10 1.91x10 1 91x10 1.91x10 1.

Flows are total systems flows because core flow is not measured.

2.

These are estimates since full power has not yet been achieved.

Other vessel flows and fluid temperatures are ceasured values.

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TABLE 2.

SUMMARY

OF SENSITIVITY STUDIES 2

(8.55 ft Double-Ended Rupture)

Axial ePeak Clad-

• Local M-W

• Whole-Core M-W Kw/ft Position, ft Temperature. F Reaction, %

Reaction, 5 16.0 2

2167 3 77

<0.~ 5 17.5 4

2112 3 01, 18.0 6

2122 3 53

<0.5

<0.5 17 1-8 2059 2.21 16.0 10 1877 1.68

<0.5 J

• CRITERIA O

i Peak clad temperature............ 2200 F-Local Metal-Water Reaction........

17%

Whole-Core Metal-Water Reaction...

1%

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6.0 REFERENCES

1. BAW-10091, "B&W's ECCS Evaluation Model Report with Specific Application to 177 FA Class Plants with Lowered Loop Arrangement," August 1974

2. BAW-10092, " CRAFT 2-Fortran Program for Digital Simulation of a Multinode Reactor Plant During Loss of Coolant," July 1974.

3. BAW-10093, "REFLOOD - Description of Model for Multinode Core Reflood Analysis," July 1974

4. BAW-10094, " Revisions to THETA 1-B, A Computer Code for Nuclear Reactor Core Thermal Analysis," IN-1445, July 1974.

5. BAW-10095, " CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident," July 1974.

6. " Status Report by the Directorate of Licensing in the Matter I

of Babcot., and Wilcox ECCS Evaluation-Model Conformance to

]

10 CFR 50, Appendix K," October 1974.

7. " Supplement 1 to the Status Report by the Directorate of Licensing in the Matter of Babcock and Wilcox ECCS 4

Evaluation Model Conformance to 10 CFR 50, Appendix K,"

November 13, 1974.

8. Letter from A.C. Thies to Mr. L. Manning Muntzing dated

- August - 5, 1974, i

9. Letter from A.C. Thies to Angelo Giambusso dated.

f September 20, 1974..

10. Letter, frcm James ~ F. Mallay to T.M. Novak ' dated p

December 18,11'974.

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11. Letter from _ James F. Malley to T.M. Novak dated 1

LNovember ~25, 1974.

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APPENDIX A OPERATING RESTR'ICTIONS The Regulatory staff has reviewed the methods used by Babcock and Wilcox to derive the LOCA-related operating limits for its plants.

The review considered the basic calculation method, the range.of operating conditions calculated, the types of uncertainties and their magnitude, and the instrumentation provided to monitor plant operation.

Based on this review, we conclude that sufficient monitoring instrumenta-tion is present to provide assurance that the plant may be operated

-within LOCA-related operating restrictions. We further conclude that operation of Oconee Units 2 and 3 within the restrictions shown en Figures A-1 through A-3,which were a part of the August 5, 1974 proposed Technical Specifications from the licensee, will assure that the heat genc c_.-... limits of Figure A-6 will not be exceeded.

For Unit 1, Figure A-4 already incorporates both criceria. Fce Ocenee Unit 1, we further conclude that the heat generation limits of Figure A-6 will not be exceeded if Unit 1 is operated within the Technical Specifications for cycle 2, provided that the following additional operating restrictions pursuant to the authority contained in 10 CFR 50.46 are imposed:

1.

The power level cutoff indicated in Figure 3 5-2-1A1 of the licensee's September 20, 1974 submitted shall be reduced from i

94 percent of rated power to 92 percent of rated. power.

The power level cutoff is defined as the maximum power at which the

. reactor can' operate without regard to the reactivity held by xenon.

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2.

Power level shall not be greater than 92 percent (power level cutoff) unless one of the following requirements is met:

I Quadrant tilt is less than or equal to 2.5 percent and the a.

xenon reactivity is within 10 percent of the' value for operation at steady-state rated power.

b.

Quadrant tilt is greater than 2.5 percent and the xenon

' reactivity is within 5 percent of the value for operation at steady-state rated power.

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Operation shall be within the control rod withdrawal limits as shown in Figure A-4.

l 4.

Operation shall be-within the power imbalance envelope as shown in Figure A-5 I

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I, Rot, nr.oi x i s int-Pr. ACfNT AGI Stat or Tut ws THDR AWAL Of Tut CPI 4AilflG GROePS.

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THE Ar>DI T ION AL HESTRICTIONS ON 4: THORAWAL (H ASHE D ART AS ) Aki EODifi LD ArirR 100 rULL POWER DAYS OF oprnArion.

100 125 187 217

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Res t r ic t eri

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A E 80 I-370 222 Region v{

Power Level Cutolf E

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OPERATING O

A 10 g

REGION C

20 f

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50 100 150 200 250 300 Rod index, f,Witndrawal 0

25 50 75 100 I

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Group 6 -

0 25 50 75 100 t

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Group 5 FIGURE A-1 CONTROL R00 GROUP'WITH0RAWAL LIMITS FOR 4 PUMP OPERATION - UNITS 2, 3 b

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1 P00 lNDEx 15 Te4C PERrEN1 AGI SUV Of Teif As Ti4M ANAL OF a

Isti uf'l DATING GROiJPS.

2.

TisE A(.DI T ION AL RESTRICTIONS ON WI THDR A AAL IHASHED ARIASI ARE IN Ef f ECT AF TER 10] Ft:LL POWER DANS Of ODERATION, Rr.STRICTIONS ON WITH11RAAAL IHASHED ARI. AS S arf r'JHTHf R MODI FI ED AFTEle 435 FULL POAER DAYS CI' OPERATlON.

100 125 182 253

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90 _

90 182 253 4

80 '

Re s t r i c t eil Power Levei g

Region

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3 60 g

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50 100 150 200 250 300

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Rod index, *, Witndrawal 0

.25 50

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,100 Group 7 1

0 25 50 75 100-t i

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I Group 6 0-25 50 75 100 FIGbREA-2

. Group 5 i

C011 TROL R00 GROUP'WITH0RAWAL LIMIT:

FOR 4' PUNP ' 0P E R Ai_i ON - UNITS 2, 3.

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1. ROD INDEX l S 1HE PERCENTAGE SU'.1 OF THE WI THORAWAL OF THE OPERATING GROUPS.

2. THE ADDITIONAL RESTRICTIONS ON WITHDRAWAL (H ASH ED AREAS) ARE IN EFFECT AFTER 435 FULL POWER DAYS OF OPERATION.

291.4 125 M

100 90 82.5 -

244.

3 f

80 RESTRICTED POWER --

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Grcup 5 FIGURE A-3 CONTROL R00-GROUP WITH0RAWAL LIMITS FOR 4 PUNP OPERATION UNITS 2,.3 A-5

P. _ Rod inder'is the percentage su:n of the eithdraral of the operating groups.

2 The witndranal limits are modified after 250 2 5 full power days of cperation.

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s 9

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APPENDIX B ADVISOnY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. d.C. 20548 Novembe; 20, 1974 Honorable Dixy Lee Ray Chairman U. S. Atomic Energy Commission Washington, D. C. 20545 Subj ect:

REPORT ON EVAi.UAN ON MODELS FOR COMMISSION CRITERIA FOR EMERGENCY CORE COOLING SYSTEMS FOR LIGHT-WATER-COOLED NUCLEAR POWER REACTORS

Dear Dr. Ray:

At its 175th meeting, November 14-16, 1974, the Advisory Committee on Reactor Safeguards completed a review of Evsluation Models which have been submitted in accordance with the Commission criteria set forth in 10 CFR 50.46.

The following subcommittee meetings with reactor vendors were held in Washington, D. C.:

March 23, 1974, Babcock and Wilcox; April 25, 1974, General Electric Company; April 26, 1974, Westinghouse Electric Corporation; and May 18, 1974, Combustion Engineering, Inc.. Subcommittee meetings were held with the Regulatory Staff and their consultants in Washington, D. C.,

on August 6, 1974, September 28, 1974 and October 26, 1974. The Committee also had the benefit of the documents listed below.

Previous reports to the Cormiission on interim acceptance criteria were made on January 7,1972, and on the proposed changes on September 10, 1973. The Committee has also addressed the safety research pregrams and the latest report is'on November 20, 1974.

1 The ACRS believes that the four light-water reactor vendors have developed ~

Evaluation Models which, with the additional modifications required by.the Regulatory _ Staff, will confom to Appendix K to Part 50.

Approved Evaluation Models will aid in conducting the licensing reviews, but a variety of specifics must be evaluated on a case-by-case basis.

Items-such as the particular features of a containment, sequencing of operations, singic failure n alysis and special. features of the reactor design, are noted in the Staff's review of the vendor models. Additional items involving peaking factors and treatment of the uncertainties in the power distributions and monitoring of the power levels remain to be incorporated, case-by-case,

'in the Technical Specifications with appropriate conservatism.

AEV138330 3!". 30 33ldd0 OE : F Nd 02' AOL N.

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. November 20, 1974 Honorable Dixy Lcc Ray The generic review of the vendor models proposed for Appendix K, like the reviews of the Interim Acceptance Criteria models, has contributed to improved understanding of the modeling techniques, including the applicability and limitations on current knowledge of thenral and hydraulic phenomena, and the need for more definitive safety research programs'and code developments. The implementation of safety research programs, noted in the Committee's (November 20,1974) report, and their results should have impact on the future evaluation methods and ECC systems.

The ACRS remains mindful that the Eva'luation Models, in themselves are not the desired end products, but that effective, reliable emergency core cooling systems are the objective. The Committee acknowledges the contri-bution to reduced peak clad tcapcratures resulting from recent core design changes but reaffirms its position stated in the September 10, 1973 report that improved ECCS reliability and capability should continue to be sought and, to the extent practical, employed.

Sincerely yours, n/

h W. R. Stratton Chairman References Attached.

b 8

I o

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- 1 a

T

Honorable Dixy Lee. Ray November 20, 1974 References t

1) WCAP-8170 (P) dated June 1974, "Caleulational Model for Core Reflooding After a Loss-of-Coolant Accident"

2) WCAP-8200 (Rev. 2 (P)) dated June 1974, "WFLASH - A Fortran - IV Computer Program for Simulation of Transients in a Multi-Loop PWR"

3) WCAP-8301'(P) dated June 1974, "LOCTA - IV Program Loss-of-Coolant Transient Analysis"

4) WCAP-8302 (P) dated June 1974, " SATAN IV Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant"

5) WCAP-6327 (P) dated July 1974, " Containment Pressure Analysis"

6) WCAP-8339 (NP) dated June 1974, " Westinghouse Emergency Core Cooling System Evaluation Model, Summary"

7) WCAP-8340 dated July 1974, " Westinghouse ECCS - Plant Sensitivity Studies"

8) WCAP-3341 (P) dated July 1974, " Westinghouse Emergency Core Cooling System Evaluation Model Sensitivity Studies"

9) WCAP-8354 (P) dated July 1974, "Long Tem Ice Condenser Containment Code -

IATIC Code"

10) BAW-10091, "B&W's ECCS Evaluation Model Report with Specific Application to 177 FA Class Plants with Lowcred Loop Arrangement," August 1974

11) BAW-10092, CRAPr2 Program for Digital Simulation of a Multinode Reactor Plant Durir

. wr-Coolant," July 1974

12) BAW-10093, "REFLOOD - Description of Model for Multinode Core Reflood Analysis," July 1974

13) BAW-10094, " Revisions to THETA 1-B, A Computer Code for Nuclear Reactor Core Thermal Analysis," IN-1445, July 1974

14) BAW-10095, " CONTEMPT - Computer Program for Predicting Containment Pressure-Temperature Response to a Loss-of-Coolant Accident," July 1974

15) CENPD-132, " Calculative Methods for the CE Large Break LOCA Evaluation Model," (P) August 1974

16) CENPD-134 (P) April 1974, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core"

17) CENPD-135 (P) April 1974, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program"

18) CENPD-136 (P) July 1974, "High Temperature Properties of Zircaloy and UO2 for Use in LOCA Evaluation Models"

19) CENPD-137 (P) August 1974, " Calculative Methods for the CE Small Break LOCA Evaluation Model" 20)- CENPD-138 (P) April 1974, " PARCH'- A Fortran IV Digital Computer Program-to Evaluate Pool-Boiling Axial Rod and Coolant Heatup"

21) CENPD-139, "CE Fuel Evaluation Model (FATES Thermal Performance and Densification Model)"

22) NEDO-20566 (DRAFr) (P)

" General Electric Analytical Model~for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K"

23) Status Report'By The Directorate of Licensing In The Matter Of Babcock and-Wilcox ECCS Evaluation Model Conformance To 10 CFR 50, Appendix'K

(undated)

4

~

Honorable Dixy Lee Ray November 20, 1974

24) Supplement 1 To The Status Report By The Directorate Of Licensing In The Matter Of Babcock And Wilcox ECCS' Evaluation Model Conformance To 10 CFR 50, Appendix K, November 13, 1976

25) Status Report By The Directorate Of Licensing In The Matter of Combustion Engineering, Inc. ECCS Evaluation Model Conformance To 10 CFR 50, Appendix K (undated)

26) Supplement To The Status Report By The Directorate Of Licensing In The Matter Of Combustion Engincering, Inc. ECCS Evaluation Model Conformance To 10 CFR 50, Appendix K, November 13, 1974 27)

Status Report By The Directorate Of Licensing In The Matter Of General Electric ECCS Evaluatien Model Conformance To 10 CFR 50, Appendix K (undated)

28) Supplement 1 To The Status Report By The Directorate Of Licensing In The, Matter Of General Electric ECCS Evaluation Model Conformance To 10 CFR 50, Appendix K, November 13, 1974

29) Status Report By The Directorate Of Licensing In The Matter Of Westinghouse Electric Company ECCS Evaluation Model Conformance To 10 CFR 50, Appendix K (undated)

30) Supplement To The Status Report By The Directorate Of Licensing In The Matter Of Westinghouse Electric Company ECCS Evaluation Model Conformance To 10 CFR 50, Appendix K, November 13, 1974

31) WREM: Water Reactor Evaluation Model, October 1974, Regulatory Staff-Technical Review

32) Water Reactor Evaluation Model (WRIM): PWR Nodalization And Sensitivity Studics, October 1974, Regulatory Staff-Technical Review

33) Water Reactor Evaluation Model (WREM): BWR Nodalization And Sens! tvity Studies, October 1974, Regulatory Staff-Technical Review 34)' Evaluation Of LOCA Hydrodynamics, November 1974, Regulatory Staff-Technical Review h

,~