ML19289G320
| ML19289G320 | |
| Person / Time | |
|---|---|
| Site: | Crane, Midland  |
| Issue date: | 10/26/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19289G307 | List: |
| References | |
| NUDOCS 7908220499 | |
| Download: ML19289G320 (3) | |
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211-39 211.0 REACTOR SYSTEMS BRANCH 211.166 Your response to first round question 222.1-2 is insufficier.t.'
(150)
We requested a description of the detailed caiculational method used, however, Section 15.1.5.3.2 of-the FSAR provides only a brief description of TRAP-2 code with reference to RADAR code. Also, recent discussions indicate that the Midland sterm line break analysis does not consider the effects of a stuck rod on the power distributions assumed in this analysis. We require that the power distr *bution dis-tortions caused by a stuck rod be considered during both the initial portion of your analysis and the later return to sub-critical power. Provide the detailed calculational: method used for the steamline break analysis.
211.167 Describe how all input parameters were_cottinei, trcluaing t;ie (15D) initial values. Other computer codes usid to generata input.
variables should also be identified.
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211.168 Dascribe now the radial, axial and local r.over distributions (15D) were calculated and used in the RADAR ede. First round 1
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question 222.1-5 requested transient axial end. radial power disse:butions instead of design peaking fact.ms. P rovide the answers to this question.
211.169 Provide a detailed description of how the radial, axial and (lSD) local hot channel factor; are applied in the RADAR code for the hot shannel and the core averzge crannel. Describe how tima dependence of the peaking factors is taken into account.
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211.170 The nodali ation diagrin show en Figrre 150-1 does not include (15D) dead volume in the reactor vessel vpper bcad. Justify that the use of this volume is not necessary in tea modeling of the steam e
lir.n break cnalysis. Dest ibe Lcw flashi.g in the primary system following emptying of the pressur!:er is handled.
211.171 Describe how the pressure drop and coolant flow rates through the
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(150) hot channel vere obtained and used in the RADAR code.
211.172 In additic.$ to the total time dependent rezerivTty fc3dback, (150) provide each component of reactivity feedsack -(Modarat]r, Doppler, red worth, boron injection).
3 211.173 Provide the core average coolant. densf ry.. rad care averace.
(15D) boron concentration for the first 15 seconf; 'or botn BDL i
and EOL cone;itions from full power.
211.174
?rovide a detailed description of the certted water flow path (150) into the core following a steamline break accident including a o
discussic, of the boro.' transpor: deley time.
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21T-41 211.175 Your response to question 211.48 with respect to demonstration (14.2) of ECCS recirculation flow from the reactor building sumo to (RSP) the Reactor Coolant System in accordance with Section C 1.b(2) of Regulatory Guide 1.79 is not acceptable. We require that you perform or reference tests which verify vortex control, available net positive suction head and acceptable pressure drops across screening, suction lines and valves, daring the recirce'ation mode of ECCS operation. Temporary holding facilities and/or scaled testing may be appropriate if suf tably justified.
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