ML19319D791
| ML19319D791 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/02/1974 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8003250789 | |
| Download: ML19319D791 (6) | |
Text
AEC DIST2 M TION FOR PART 50 DCCKET "ATERIAL (TE2iPOP.ARY F0EM)
CONTROL No: 8175 FILE:
FROM:
DATE OF DOC DATE REC'D JLTR TWX RPT CTEER Sacramento Municipal Utility Dist Sacramento, Cal.
95813 8-2-74 8-7'74 X
J. J. Mattimoc T02 ORIG CC OTHER SENT AEC PCR X
A. Giambusso 1 signed SENT LOCAL PCR X CLASS UNCLASS
?RCP INFO INPUT NO CYS RIC'D DOCKET NO:
XXXX XXX 50-312 DESCRIPTION:
ENCLOSURES:
i Ltr submitting Appendix "K" Calculations & Proposd Amdt to the OL: Consisting of....,......
ed rev to Tech Specs...re loCFR Part 50, Appendix (1) Loss of Coolant Accident ant. lyses con-
"K".........
formance with 10CFR50, ApNndix "K" DO NOT REMOVE (2) Progesed rech Spec aev4.1on (3) RE Topical Reports 3t's-10092, BAW-10093, BAW-10094, BAW-10095, BAW-10091 ACKNOWLEDGED PLANT' NF : Rancho Seco
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Regulatory Docket File 3 SACRAMENTO MUNICIPAL UTILITY DISTR SMUD O 6201 S Street, Box 15830, Sacramento, California 95813; (916) 452-3211 l
l August 2,1974
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Mr. Angelo Giambusso Deputy Director for Reactor Projects 1
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Directorate of Licensing i
U. S. Atomic Energy Commission
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Washington, D. C.
20545 x-
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a, AEC Docket No. 50-312 l
ECCS Final Acceptance Criteri~a Analysis Rancho Seco Nuclear Generating Station, Unit No.1
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Dear Mr. Giambusso:
On January 4,1974 the Atomic Energy Commission issued the Final Acceptance Criteria (FAC) for ECCS evaluation as revisions to 10 CFR 50.45.
In compliance with these revisions, Babcock & Wilcox has developed an evaluation model which meets the requirements of Appendix K of 10 CFR 50. The description of this model is contained within the their nonproprietary topical report BAW-10091, "B&W's ECCS Evaluation Model Report with Specific Application to 177 FA Class Plants with Lowered Loop Arran Directorate of Licensing (DOL)gement," which has been submitted to the on August 5,1974.
In addition, B&W has provided supporting documentation for the computer codes utilized in this model in the following nonproprietary topical reports:
1.
BAW-10092, " CRAFT 2-Fortran Program for Digital Simulation of a Multintie Reactor Plant During Loss of Coolant" 2.
BAW-10093, "REFLOOD - Description of Model for Multinode Core Reflood Analysis" 3.
BAW-10094, Babcock & Wilcox Revisions to THETAl-B, a Computer Code for Nuclear Reactor Code Thermal Analysis - IN-1445, 4.
BAW-10095, Babcock & Wilcox Revisions to CONTEMPT -
Computer Program for Predicting Containment Pressure -
Temperature Response to a Loss-of-Coolant Accident, The analysis presented in BAW-10091 for the B&W 177 FA class plants with lowered loop is generic in nature, since the plant parameters utilized in the analysis (such as the rated power level, fuel densification and containment building volume) are taken to be the most conservative values for all the plantt of this type. Thus, the results contained in q j ~ ;-
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Mr.' Angelo Giambusso August 2,1974 BAN-10091 provide an overly conservative analysis for all plants of this type and can be applied to the Rancho Seco plant. As such, we adopt the results contained in BAW-10091 to this plant. These results
-demonstrate conformance to the criteria of 10 CFR 50.46 under the following operating conditions:
1.
The peak linear heat rate is less than or equal to 17.8 kW/ft. at the six foot elevation.
2.
The plant will operate within the attached revision to the proposed. Technical Specifications for the loss-of-coolant limits. These revisions are based on the most restrictive LOCA limits as established from the revised ECCS analysis (FAC) (BAW-10091) and from the Interim Acceptance Criteria.
Justification for the use of the previously proposed control rod position limits for the Rancho Seco plant is attached and as such, operation for this plant will be in compliance with 10 CFR 50.46.
Sincerely yours,
- .'}ps W x c
' J. J. Mattimoe Assistant General Manager and Chief Engineer
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Attachments 1
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w LAC Loca Limit I
18 e a wa~
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FAC i 'f 16 t
Operating Envelope for Rod
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Position Limits in Previously 1
g Proposed Technical Speolfications b
14 J
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0 12
.9 10 0
0 2
4 6
8 10 12 Distance from Inlet, ft CollP ARIS0ll 0F RAXI!40R llNE AR HE AT RATE ENVELOPES VERSUS AX1Al ELEVATID!?
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS c
Limiting Cone.. ions for Operation 3.5.2.7 The control rod drive patch panels s.he.ll be locked at all times with limited access to be authorized by the superintendent.
Bases The power-imbalance envelope defined in figure 3.5.2-2 is based on LOCA 28 analyses which have defined the maximum linear heat rate (see figure 3.5.2-3)
.o I such that the maximum clad temperature will not exceed the Final Acceptance 281 Criteria. Operation outside of the power imbalance envelope alone does not 9 I constitute a situacion that would cause the Final Acceptance Criteria to be 28l exceeded should a LOCA occur. The power imbalance envelope represents the o i boundary of operation limited by the Final Acceptance Criteria only if the 281 control rods are at the position limits as defined by figure 3.5.2-1 and if a 4 percent quadrant power tilt exists.
Additional conservatism is introduced by application of:
A.
Nuclear uncertainty factors.
24 B.
Thermal calibration uncertainty.
C.
Fuel densification effects.
D.
Hot rod manufacturing tolerance factors.
The 30 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or bcnks defined as follows:
Group Function 1
Safety 2
Safety 3
Safety 4
Safety 5
Regulating 6
Regulating 7
Regulating 8
APSR (axial power shaping bank)
Control rod groups are withdrawn in sequence beginning with group 1.
Group 5 is overlapped 25 percent with groups 6 and 7, which operate in parallel.
The normal position at power is for groups 6 and 7 to be partially inserted.
The minimum available rod worth provides for achieving hot shutdown by reactor trip at any time assuming the highest worth control rod remains in the full out position.(1)
Inserted rod groups during power operation will not contain single rod worths 4
greater than 0.65 percent ak/k.
This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident.(2)
A single inserted control rod worth of 1.0 percent ak/k at beginning of life, hot, cero power would result in the same transient peak thermal power and therefore the same environmental' consequences as a 0.65_ percent ak/k ejected rod worth at rated power.
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10 12 AXIAL LOCATION OF PE AK POWER FROM BOTTOM OF CORE FT FIGURE 3.5.2-3 LOCA LIMITED MAXIMUM RANCHO SECO UNIT 1 ALLOWABLE LINEAR HEAT RATE TECHNICAL SPECIFICATIONS
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