ML19319D932
| ML19319D932 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 09/13/1978 |
| From: | Kaplan D, Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| References | |
| NUDOCS 8003270608 | |
| Download: ML19319D932 (3) | |
Text
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DISTRIBUTION FOR INCOMING MATERIAL 50-312 REC: REID R W ORO: MATTIMOE J J DOCDATE: 09/13/7S NRC SACRAMENTO MUN UTIL DIST DATE RCVD: 09/18/7E DOCTYPE: LETTER NOTARI ZED: YES COPIES RECEIVED
SUBJECT:
LTR 1 ENCL 40 FORWARDINO LIC NO DPR-54 APPL FOR AMEND: TECH SPEC PROPOSED CHANGE NO 54 CONCERNING REVISIONS TO THE SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS. TO ASSURE OPERATION OF UNIT 1 CYCLE 3 CORE WITHIN APPLICABLE FUEL DESIGN AND PERFORMANCE CRITERIA.
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- no************** DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS ******************
GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE A001)
FOR ACTION:
BR CHIEF ORD#4 BC**W/7 ENCL INTERNAL:
FILE **W/ ENCL (2 NRC PDR**W/ ENCL A a t**W/2 ENCL OELD**LTR ONLY HANAUER**W/ ENCL CORE PERFORMANCE BR**W/ ENCL AD FOR SYS & PROJ**W/ ENCL ENGINEERING BR**W/ ENCL REACTOR SAFETY BR**W/ ENCL PLANT SYSTEMS BR**W/ ENCL EEB**W/ ENCL EFFLUENT TREAT SYS**W/ ENCL J MCGOUGH**W/ ENCL EXTERNAL:
LPDR'S SACRAMENTO, CA**W/ ENCL REGION V**W/ ENCL TERA **W/ ENCL NSIC**W/ ENCL ACRS CAT B**W/16 ENCL l
i 8003270 h DISTRIDUTION:
LTR 41 ENCL 40 CONTROL NBR:
%W SIZE: 2P+24P+49P r
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THE END
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9 gSMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT C 1708 59th Street. Box 15830, Sacramento, California 95813;(916)452-3211 September 13, 1978 Director of Nuclear Reactor Regulation ATTN:
Mr. Robert W. Reid, Chief Operating Reactors, Branch 4 U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Re:
Docket No. 50-312 Proposed Amendment No. 60 Rancho Seco Nuclear Generating Station, Unit No. 1
Dear Mr. Reid:
In accordance with 10 CFR 50, Section 50.90, the Sacramento Municipal Utility District hereby proposes to amend its Operating License DPR-54 for Rancho Seco Nuclear Generating Station, Unit No. 1, by submitting forty (40) copies of Proposed Amendment No. 60.
The purpose of these revisions is to assure operation of the Rancho Seco Nuclear Generating Station, Unit No. 1, Cycle 3 core within applicable fuel design and performance criteria. The proposed change, are shown in Attachment I as replacement pages for the Rancho Seco Nuclear Generating Station Technical Specifications:
2.1 SAFETY LIMITS, REACTOR CORE 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING AND REACTOR BUILDING SPRAY SYSTEMS 3.5.2.4 THE QUADRANT POWER TILT Also attached is Babcock & Wilcox Report BAW-1499, " Rancho Seco Nuclear Generating Sta tion, Uni t No. 1, Cycle 3 Reload Report".
This report includes a summary of Cycle 2 operating parameters and contains the safety analyses supporting the operation of Rancho Seco No. 1 Cycle 3 core at rated power in accordance with the Technical Speci fications provided.
TN_Y
T Mr. Robert W. Reid September 13, 1978 Section 3.3.3 -is being changed to be in agreement with methods used within the Standard Technical Specifications and interests of safety. The
.present Specifications ' require that a component that is to have maintenance be removed from service and the redundant component tested. The testing of the redundant system requires that the system be removed from operation to have an _ operational. test performed. During this time interval the system is not available for safety features operation. When the testing is completed it'Is returned to service. We believe, and are in agreement with the' Standard Technical Specifications, that tests other than those specified in the Surveillance Standards should not be required for proof of operability.
The Surveillance Standards test is sufficient to prove operability of a component for the duration specified in Technical Sr scifications Section 4.
If you have any questions concerning this matter, please contact Mr.
R.W. Colombo at the Rancho Seco Nuclear Plant.
Respectfully submitted, 0 -
J
%'. jr a V.s.R J. J. Mattimoe Assistant General Manager and Chief Engineer 3
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GeneralCounselffor Sacramento Municipal Utility District Subscribed and sworn to before me this 13th day of September, 1978.
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. SACRAMENTO COUNTY County of Sacramento, State of California My Commission Expires Janusy 13.1981
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RANCHO SECO I TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.
Objective To meintain the integrity of the fuel cladding.
Specification 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1.
If the actual pressure / temperature point is within the restricted region the safety limit is exceeded.
2.1.2 The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2.
If the actual-reactor-thermal power / reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded.
Bases The safety limits prpsented have been generated using BAW-2 critical heat flux (CHF) correlation \\Il and the actual measured flow rate (2).
This deve is discussed in the Rancho Seco Unit I, Cycle 2 Reload Report, reference {oment 2
The flow rate utilized is 104.9 percent of the design flow (369600 spm) based on 60 four pump operation.
(2,3)
To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions.
This is accomplished by operating within the nucleate boiling region of heat transfer, wherein the heat transfer coef ficient is is only slightly greater of the nucleate boiling (DNB). At this point, DUPLICATE DOCUMENT cefficient, which would result of cladding failure. Although Entire document previously ope ra t i on, the observable entered into system under:
temperature, a pressure 790f
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No. of pages:
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Proposed Amend nent No. 60
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