ML19220A806
| ML19220A806 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/03/1975 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904240632 | |
| Download: ML19220A806 (3) | |
Text
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V. T. tore, '_sistaat Director for Light Water Peactors, Groun 2, FL Set 7 '.0UND OCSSTICMS - THREE MILE ISLAND NUCLEAR STATI0'!, U'iIT Plant Nam :
Three Mile Island, Unit 2 Licensing Stage:
CL Cccket No.:
STM 50-320 Responsible Branch LWR 2-2 and Project Manager:
- 3. Washburn Technical Review Branch Involved:
Reactor Systers Branch Recuested Conoletion Cate:
January 31, 1975 Description of Review:
Second Pound Cuestions Review Status:
Awaiting Infer-ation The Reactor Systen Branch has continued the review of the subiect FSAR and reviewed the a~;dicants response to round one questiens.
The following list of questions must be adequately ans',ered before preparation of the SER.
j- - 3 med Ly VcIo S$110, Jr., Assistant Ciractor for Reactor Safety Division of Technical Review Office of Nuclear Reactor Regulation
Enclosure:
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21.46 (Reference S2 21.19)
A parametric study is mentioned in the response, the results of which lead to figure 2.1-2 of the Technical Specification.
Describe the analysis performed in this study.
Include the range of pcwer and other variables considered.
Present the results of the study leading to the aforementioned figure.
The follcwing question is designed to aid in achieving a better understanding of the potential problems in tnic a.~e a :
a) Provide pcwer distributions at increments of pcwer level frcm 30% to 1C0% pcwer. Take into account limits on rod insertion versus pcwe.
b) Select from (a) the worst pcwer distribution for initial conditions. Then demonstrate hcw *he trip system prevents the DNBR from falling belcw 1.3 during a boron dilution excursion leading to over-power.
21.47 (Reference S2 21.20)
As indicated in the response to the reference question, reanalysis in conformance with the NRC Final Acceptance Criteria and updating of the FSAR are required.
Secticns of the FSAR affected inciude 4.4 and figure 2.1-1, Loss of Coolant Accidents in Chapters 6 and 15. The Technical Specifications of Chapter 16 will r. iso requirc ucdating.
21.48 (Reference 52 21.33)
The staff agrees with the applicant as to the apolica-bility to this clant of Critericn 6 of the 1967 general design criteria.
The staff dces recuire substantiation that sufficient condensate is available to maintain decay heat removal by use of the steam generator until a failed isolation valve is repaired or replaced.
Assuming that a normal sequence through hot shu t down has been comoleted and that a single failure is then encountered at actuation of the RHR system:
1.
Define the time that decay heat removal could be continued us;ng water frcm the condensate tank after reinstatement.
Identify other sources available and potential decay heat removal time associated with them.
70~-039
2.
Provide a conservative estimate of the time required to repair or replace either of the two valves.
Provide basis for estimate.
21.49 (Reference S2 21.44)
The applicant's response does not satisfy the concerns of the staff relative to a feecwater line rupture. A rupture between the upstream check valve and the steam generator is postulated to cause a more rapid loss of heat sink than the case referenced (ccmplete loss of normal feedwater). Provide the analysis including the folicwing.
(1) Sensitivity analysis showing that the break size chosen is the most severe one.
(2) The method used to calculate the discharge flow frcm the break and also primary and secondary system safety and relief valves.
(3)
Initial conditions including initial mass in each steam generator.
Provide justification for selection.
(4) Curves as a function of time for:
(a) Average nrimary ccoiant temperature.
(b) Primary coolant pressure.
(c)
Each steam generator inventory.
(d) Fluid conditions and ficw frca the break.
(e)
Each steam generatcr pressure.
(f)
Each steam generator heat transfer area (wetted surface).
(g) Main feedwater ficw to each steam generator.
(h) Main steam ficw frcm eacn steam generator.
(i) Auxiliary feedwater ficw and enthalpy to each steam ceneratcr.
(j) Reactor pcwer.
(k) Raactor vessel water volume and level.
(1) Containment pressure.
(m) Primary and secondary relief rates through safety and relief valves.
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