ML20040A437
| ML20040A437 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 12/02/1975 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Moore V C Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201210087 | |
| Download: ML20040A437 (4) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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W ASHINGTON, D. C. 20555
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DEC 2 1975 oore, Assistant Director for Light Water Reactors, Group 2, RL BABC CK & WILCOX ACCEPTANCE REVIEW PLANT NAME:
Babcock & Wilcox (B-SAR-205)
PROJECT NUMBER: P-566 PROJECT MANAGER:
T. Cox, LWR 2-3 REQUESTED COMPLETION DATE: November 18, 1975 REVIEW STATUS: AAB Acceptance Review Complete This memo contains the evaluation of the Babcock & Wilcox Standard Safety Analysis Report (B-SAR-205) by the Accident Analysis Branch (AAB). This review was coordinated by Charles Ferrell, Site Analyst, AAB.
Brandt members who participated in this review were:
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Section B-SAR-205 Description B-SAR-205 AAB Reviewer 3.5 Missile Protection K. Campe/H. Fontecilla 15.4 Design Basis Accident C. Ferrell/H. Fontecilla Analysis Attached is our partial first list (Acceptance Review) questions for this standard plant. Additional questions by H. Fontecilla on tornado design, accidents, and plant interfaces will be provided at a later date. We con-clude that the B-SAR-205 report is complete enough to permit docketing.
/#f H. R. Denton, Assistant Director for Site Safety Division of Technical Review Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: w/o enclosure R. Boyd W. Mcdonald J. Panzarella
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W. Pasedag K. Campe K. Murphy C. Ferrell 0
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i 301.1 With respect to justifying the direct-leakage rate assumptions (6.2) used in calculating the radiological consequences of a LOCA, l
provide the following:
A specification of the fraction of total containment through line leakage assumed to (a) terminate in creas served by filtra-l tion and vent systems and (b) discharge directly to the atmosphere.
The specific leakage rates for each through-line leakage path.
A discussion of the tests and frequency of tests proposed to detect and limit the through-line leakage.
Identify all process lines (steam lines, electrical penetrations, instrument lir s, etc.) which are open to the primary containment and which terminate in plant areas not served by penetration rooms or enclosure building filtered ventilation systems after a postulated accident.
110.2 Section.15.1 of the B-SAR-205 PSAR indicates that 64 out of 264
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i.1) fuel rods were assumed to be damaged in a fuel handling accident. __
j Our position is that this analysis should be based on the failure of all of the rods in a E el assembly.
t 310.3 It is noted in Table 15.120-1 that the fuel handling accident is t
(15.6) based on a power level of 4100 MWt, and in Table 15.1.13-1 that the LOCA analysis is based on 4100 MWt. Table 1.3-1 indicates that the Babcock-205 design is 3820 MWt. Please explain the discrepancy.
310.4 Assume that in the event of a loss-of-coolant accident the ECCS (15.4) equipment leaks at a maximum possible leakage rate (i.e., postu-late a damaged seal, or packing, or some other leakage path in which leakage would be at a maximum but not great enough to cause the pump or equipment to be defined inoperable). Calculate the fission product enventory available for release to the environment from ECCS leakage.
Include the following parameters in your analysis:
Concentration of iodine and noble gas activity in the primary a.
containment sump water following a LOCA.
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Temperature curve vs. time for water being circulated through ECCS pumps following a LOCA.
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Expected maximum leak rate through pump seals, flanges, valves, c.
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Partition factor for iodine.
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(Cent'd)
Provide an estimate of the total amount of leakage that could occur prior to isolation of failed ECCS equipment such as a pump seal.
310.5 Provide data in Section 3.11 on the radiation environment, (3.11) both for normal and post-accident conditions, for safety-related equipment located within the containment or immedi-ately adjacent to recirculation lines outside containment that may carry radioactive fluids in a post-accident environ-ment.
Provide tables listing the radiation dose rates and integrated doses as a function of equipment location within containment, time after a postulated accident, and by type of radiation (neutrons, beta and gamma rays).
State all your assumptions and list that equipment important to safety that will be qualified and the radiation levels to which this equip-ment will be qualified.
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AUG 2 31976 t
e Docket No. 50-561 MEMORANDUM FOR:
J. F. Stol::, Chief, Light Water Reactors Branch No.1, DPM John T. Collins, Chief, ETSB, ESE FROM:
SUBJECT:
ROUND TWO QUESTIONS FOR B-SAR-205 PLANT NAME:
B-SAR-205 LICENSING STAGE: CP DOCKET NUMBER: 50-561 MILESTONE NUMBER: 03-01 RESPONSIBLE BRANCH:
LWR No. 1 PROJECT MANAGER:
T. Cox DESCRIPTION OF RESPONSE: Round Two Questions REQUESTED COMPLETION DATE: August 23, 1976 REVIEW STATUS: Under Review We have reviewed the revised Chapter 11.0, " Radioactive Waste Management",
i and related sections of B-S.an-205, and have determined that additional information is necessary to cor.plete our review.
Enclosed is the addi-tional information concerning lic,uid and gaseous source terms that we
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will need to continue our renew.
We will need the additien91 information by October 22, 1976 to meet our schedule.
4/%h ohn T. Collins, Chief Effluent Treatment Systems Branch Division of Site Safety and Environmental Analysis
Enclosure:
Round Two QuestiBns cc:
S. Hanauer H. Denton D. Muller F. Miraglia J. Miller S. Varga R. Boyd
,R. DeYoung
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Docket File 50-561 NRR Reading File DSE Reading File
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AUG 2 31976 320.0 EFFLUEhT TREATMENT SYSTEMS BRANCH Your response to Question 320.7 states'that expected releases 320.10 (11.1.5) from the boron recovery system are accomplished by bleeding to the liquid waste system a portion of the recovered distillate downstream of the distillate test tanks.
Provide the expected and maximum volumes, radioactive concentrations, and chemical content of this stream as interface information to the BOP designer.
320.11 In Amendment 2, you state that the airborne tritium concentra-(11.1. 2. 2) tion is based on a 175 lb/hr evaporation rate from the re-fueling canal during the 14-day refueling periods. Provide the bases for this rate, preferably citing operating experience.
Provide the bases for the tritium concentration given in Figures 11.1-1 thru 11.1-5, including assumed mixing between reactor water 'and fuel pool 'during refueling. Cite previous pertinent experience from operating reactors. You should also consider the use of Li-0H containing less than 99.99%
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