ML20150E637

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Errata to NEDO-24082, LOCA Analysis for Peach Bottom Atomic Power Station Unit 3
ML20150E637
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 01/31/1988
From:
GENERAL ELECTRIC CO.
To:
Shared Package
ML19292J065 List:
References
NEDO-24082-ERR, NEDO-24082-R07-ERR, NEDO-24082-R7-ERR, NUDOCS 8807150232
Download: ML20150E637 (6)
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Text

4 APPLICABLE TO:

N - 4082 PUBLICATION NO.

ERRATA And ADDENDA T. i . E. " '

SHEET TITLE LOCA Analysis for Peach 7

NO.

Bottom Atomic Power Station DATE

. January 1988 Unit 3 NO TE: Correct al; copies of the applicable tsSUE DATE O'C'"h*T l9I7 publication as specified below.

REFERENCES (SECYlON, PA G E INSTRUCTIONS ITE M PA R AG R APH, LIN E) (CORRECTIONS AND ADQlTIONS) 01 Page v/vi Replace with new page v/vi.

02 Page 3-1/3-2 Replace with new pages 3-1 and 3-2.

03 Page 4-3 Replace with new pagcs 4-3 and 4-4.

04 Page 4-17 Replace with new pages 4-17 and 4-18.

05 Page 7-1/7-2 Replace with new page 7-1/7-2.

(Change brackets in right-hand margin indicate areas where report has been revised.)

.J GENuclearEnergy 115 Criner 4ene S3n Jose. CA 95125 8807150232 880707 \

PDR ADOCK 05000278 P PDC "

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': NEDO-24082 A

4 LIST OF TABLES Table Title Page 3-1 '

Significant Input Parameters to the Loss-of-Coolant Accident Analysis 3-1 4-1 Summary of Results s 4-5 4-2 LOCA Analysis Figure Summary 4-6 4-3a MAPLUGR Versus Average Planar Exposure 4-7

'4-3b MAPLHGR Versus Average Planar Exposure 4-8 4-3c MAPLHGR Versus Average Planar Exposure 4-9 4-3d- MAPLHGR Versus Average' Planar Exposure 4-10 4-3e MAPLHGR Versus Average Planar Exposure 4-11 4'-3 f MAPLHGR Versus Average Planar Exposure 4-12 4-3g MAPLHGR Versus Average Planar Exposure 4-13 4-3h MAPLHGR Versus Average Planar Exposure 4-14 4-31 MAPLHGR Versus Average Planar Exposure 4-15 4-3j MAPLHGR Versus Average Planar Exposure 4-16 4-3k MAPLHGR Versus Average Planar Exposure 4-17 4-31 MAPLHGR Versus Average Planar Exposure 4-13 _

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., 'NED0-24082 s

3. INPUT TO ANALYSIS Y

7 A list of the significant plant input parameters to the LOCA analysis is pre.:eor mf, in Table 3-1.

Table 3-1~

SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters:

. Core Thermal Power 3440 MWt, which corresponds to 105% of rated steam flow Vessel Steam Output 14.05 x 106 lbm/h, which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure 1055 psia Recirculation Line Break Area for Large Breaks - Discharge 1.9 ft2 (DBA)

- Suction 4.1 ft2 (DBA)

Number of Drilled Bundles 432 Fuel Parameters:

Table 3-1 continued on Page 3-2.

3-1

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_s NEDO-24082 .

-Table 3-1 (Continued)

Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Type Geometry (kW/ft) Factor Ratio

  • A. IC Type 2 7x7 -18.5 1.5 - 1. 2 B. IC Type 3 7x7 '18.5 1.51 1.2 C .' 8D274L 8x8 13.4 1.4 1.2 D. 8D274H 8x8 13.4 1.4 1.2 E. PTA260 8x8 13.4 1.4 1.2 F. 8DRB283 8 x 8R 13.4 1.4 1.2 G. P8DRB284H 8 x 8R 13.4 1.4 1.2 H. B/P8DRB299 8 x 8R 13.4 1.4 1.2 I. P8DQB326LTA 8 x 8R 13.4 1.4 1.2 J. BP8DRB299H 8 x 8R 13.4 1.4 1.2 _.

K. BD319A 8 x 3R 14.4** 1.4 1.2 L. BD321A 8 x 8R 14.4** 1.4 1.2

  • To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e. ,1.2 divided by 1.02) for a bundle with an initial MCPR 1.20. __
    • NRC approval is provided in References 13 and 14. _

3-2

t NEDO-24082 jg L 4-4.5 RESULTS OF THE CHASTE ANALYSIS This code is used,-with suitable inputs from the other codes,-to calculate the fuel cladding heatup rate, peak cladding temperature, peak local cladding oxidation, .and ' core-wide metal-water reaction for large breaks. The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and

-channel wetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly. Iterative applications of CHASTE determine the maximum permissible planar power where required to satisfy the requirements of 10CFR50.46 acceptance criteria.

- The CHASTE results presented are:

o Peak Cladding Temperature versus time o Peak Cladding Temperature versus Break Area o Peak Cladding Temperature and Peak Local Oxidation versus Planar

  • Average Exposure for the most limiting break size o Maximum Average Planar Heat Generation Rate (MAPLHGR) versus Plar.ar Average Exposure for the most limiting break size A summary of the analytical results is given in Table 4-1. Table 4-2 lists the figures provided for this analysis. The MAPLHCR values for each fuel type in the PB-3 core are presented in Tables 4-3a through 4-31. For the GE8x8EB fuel designs, the MAPLHGR values reported in this document are a composite of the most and least limiting lattice specific MAPLHGR values (enriched lattices only) documented in "Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 3 Supplement 1," NEDE-24082-P Supplement 1, December 1987.

The maximum peak cladding temperature is s2089'F and the maximum peak local 4-3

k NEDO-24083 oxidation fraction 1s110.048 at all exposures for the GE8x8EB fuel designs. o.

For single' recirculation loop operation, the MAPLHGR multiplier for these bundle designs is 0.73. --

4.6 -METHODS In the following sections, it will be useful to refer to the methods used to analyze DBA, large breaks, and small breaks. For jet-pump reactors, these are defined as follows:

a. DEA Methods. LAMB / SCAT / SAFE /DBA-REFLOOD/ CHASTE. Break size: DBA.
b. Large Break Methodr. (LBM). LAMB / SCAT / SAFE /non-DBA REFLOOD/ CHASTE.

2 Break sizes: 1.0 ft 5 A < DBA.

c. Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Heat transfer coefficients: nucleate boiling prior to core uncovery,.25 Btu /hr-ft 2 _.7 after recovery, core' spray when appropriate. Peak cladding temperature and peak local oxidation are calculated in non-DBA-REFLOOD. Break sizes A s 1.0 ft .

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Table 4-3k MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Planti PB Fuel Type: BD319A Average-Planar Least Limiting Most Limiting '

Exposure MAPLHCR MAPLHGR (GWd/t) (kW/ft) (kW/ft) 0.2 11.68 11.22 2.0 11.87 11.47 4.0 12.20 11.85 6.0 12.62 12.34 8.0 13.04 12.84 10.0 13.13 13.06 12.5 13.07 12.91 15.0 12.77 12.71 20.0 12.17 12.16 50.0 6.69 6.63 I

I 4-17

s-NED0-24082 -

Table 4-31 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: PB-3 Fuel Type: BD321A Average Planar Least Limiting Most Limiting-Exposure MAPLHGR HAPLHGR (mwd /t) (kW/ft) (kW/ft) 0.2 11.51 :11.05.

2.0 11.75 11.33 4.0 12.04 11.67 6.0 12.27 11.95 8.0 12.70 12.47 10.0 13.13 12.78 12.5 13.05 12.98 15.0 -12.76 12.74 20.0 12.18 12.16 50.0 6.68 6.62 4-18

NEDO-24082

7. REFERENCES
1. Letter, George Lear (NRC) to Edward G. Bauer, Jr. (PECO), "Re: Peach Bottom Atomic Power Station Unit No. 2," dated March 11, 1977.
2. Letter, Darrel G. Eisenhut (NRC) to E. D. Fuller (GE), Documentation of the Reanalysis Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-Lead Plants." June 30, 1977.
3. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CRF50 Appendir K, NEDO-20566 (Draft), submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to SAFE code Description) transmitted to the USAEC by letter, G. L. Gyorey (GE) to Victor Stello, Jr. (NRC), dated December 20, 1974.
4. "Safety Evaluation for General Electric ECCS Evaluation Model Modifica-tions," letter from K. R. Goller (h1C) to G. G. Sherwood (GE), dated April 12, 1977.
5. Letter, A. J. Devine (GE) to D. F. Ross (NRC) dated January 27, 1977, "General Electric (GE) Loss of Coolat Accident (LOCA) Analysis Model Revisions - Core Heatup Code CHASTE 05."
6. Letter, A. J. Levine (GE) to D. B. Vassallo (NRC), dated March 14, 1977, "Request for Approval for Use of Loss of Coolant Accident (LOCA) Evalua-tions Model Code REFLOOD05."
7. "Supplemental Information for Plant Modification to Elimina', "*;nificant  !

In-Core Vibrations," Supplement 1, NEDE-21156-1, September 19n.,.

8. "Supplemental Information for Plant Hodification to Eliminate Significant In-Core Vibrations, Supplement 2, NEDE-21156-2, January 1977,
9. Letter, R. Engel (GE) to V. Stello (NRC), "Answers to NRC Questions on NEDE-21156-2," January 24, 1977.
10. Letter, G. L. Gyorey (GE) to V. Stello, Jr., dated May 12, 1975, "Compli-ance with Acceptance Criteria for 10CFR50.46."
11. Letter, George T. Berry (PASNY) to Robert U. Reid (NRC), "James A.

FitzPatrick Nuclear Power Plant ECCS Analysis Docket No. 50-333," dated

-July 29, 1977.

12. Peach Bottom Atomic Power Station Units 2 and 3, Response to June 18, 1975 NRC Request for Additional Information on Appendix K Analysis, dated August 1975.
13. Letter, C. O. Thomas (NRC) to J, S. Charnley (GE), "Acceptance for Refer-encing of Licensing Topical Rei, ort NEDE-24011-P-A-6, Amendment 10,

' General Electric Standard Application for Reactor Fuel,'" May 28, 1985.

14. Letter, H. N. Berkow (NRC) to J. S. Charnley (GE), "Acceptance for Approval of Fuel Designs Described in Licensing Topical Report NEDE-14011-P-A-6, Amendment 10 for Extended Burnup Operation," December 3, 1985.

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7. REFERENCES'
1. Letter, George Lear (NRC) to Edward G. Bauer, Jr. (PECO), "Re: Peach Lottom Atomic Power Station Unit No. 2," dated March 11, 1977.
2. Letter, Darrel G. Eisenhut (NRC) to E. D. Fuller (GE), Documentation of the Reanalysis Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-Lead Plants." - June 30, 1977.
3. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CRF50 Appendix K, NED0-20566 (Draft), submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to SAFE code Description) transmitted to the USAIC by letter, G. L. Gyorey (GE) to Victor Stello, Jr. (NRC), dated December 20, 1974.
4. "Safety Evaluation for General Electric ECCS Evaluation Model Modifica-tions," letter from K. R. Goller (NRC) to G. G. Sherwood (GE), dated April 12, 1977.
5. Letter, A. J. Devine (GE) to D. F. Ross (NRC) dated January 27, 1977, "General Electric (GE) Loss of Coolat Accident (LOCA) Analysis Model 6 Revisions - Core Heatup Code CHASTE 05."
6. Letter, A. J. Levine (GE) to D. B. Vassallo (NRC), dated March 14, 1977, "Reques' for Approval for Use of Loss of Coolant Accident (LOCA) Evalua-tions M.edal Code REFLOOD05."
7. "Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations," Supplement 1, NEDE-21156-1, September 1976.
8. "Supplemental Information for Plant Modification tc Eliminate Significant In-Core Vibrations, Supplement 2, NEDE-21156-2, January 1977,
9. Letter, R. Engel (GE) to V. Stello (NRC), "Answers to NRC Questions on NEDE-21156-2," January 24, 1977.
10. Letter, G. L. Gyorey (GE) to V. Stello, Jr. , dated May 12, 1975, "Compli-ance with Acceptance Criteria for 10CFR50.46."
11. Letter, George T. Berry (PASNY) to Robert W. Reid (NRC), "James A.

FitzPatrick Nuclear Power Plant ECCS Analysis Docket No. 50-333," dated July 29,1977.

12. Peach Bottom Atomic Power Station Units 2 and 3, Response to June 18, 1975 NRC Request for Additional Information on Appendix K Analysis, dated August 1975.
13. Letter, C. O. Thomas (NRC) to J. G. Charnley (GE), "Acceptance for Refer-encing of Licensing Topical Report NEDE-24011-P-A-6, Amendment 10,

' General Electric Standard Application for Reactor Fuel,'" May 28, 1985.

14. Latter, H. N. dirkow (NRC) to J. S. Charnley (GE), "Acceptance for Approval of Feal Designa Described in Licensing Topical Report NEDE-14011-P-A-6, Amendaent 10 for Extended Burnup Operation," December 3, 1985.

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