ML20037A163
| ML20037A163 | |
| Person / Time | |
|---|---|
| Site: | Midland, 05000514, 05000515 |
| Issue date: | 10/11/1978 |
| From: | Schroeder F Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 7810190054 | |
| Download: ML20037A163 (4) | |
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'o un TcD STATES
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j WASHINGTON, D. C. 20555
\\'D4 Docket Nos. 50-514/515 OCT t1 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for LWRs, DPM FROM:
Frank Schroeder, Acting Assistant Director for Reactor Safety
SUBJECT:
SER SUPPLEMENT INPUT FOR PEBBLE SPRINGS UNITS 1 AND 2 Pebble Springs, Units 1 and 2 Plant Name:
i Docket Numbers:
50-514/515 Milestone Number:
27-21 Licensing Stage:
CP Responsible Branch LUR-4 and Project Manager:
C. Stable Systems Safety Branch Involved:
Reactor Systems Branch Description of Review:
SER Supplement Review Status:
Complete
References:
1.
Letter from S. A. Varga (NRC) to W. J. Lindblad (PGE) dated May 18, 1978, " Outstanding Issue Regarding Capability for Cold Shutdown of Pebble Springs Design."
t 2.
Applicant's Testimony Relating to the Atomic Safety and Licensing Board's Questions of April 12, 1978 by John L. Frewing.
Z 3.
Letter from W. J. Lindblad (PGE) to S. A. Varga (NRC) dated September 14, 1978.
Reference 1 provided the applicant with our requirements regarding cold shutdown capability. We have reviewed the applicant's' response (references 2 and 3) and find their proposal acceptable at the construction pennit stage.
Our safety evaluation is enclosed.
ank
- hroeder,t g / Acting sistant Director for Reactor Safety Division of Systems Safety
Enclosure:
l Safety Evaluation cc:
S. Hanauer V. Benaroya l
R. Mattson W. ' efave F. Schroeder T. havak
Contact:
Scott Newberry, NRR S. Varga G. Mazetis 49-27341 C. Stable S. Newberry i
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e Enclosure Our requirements for taking the plant from power operation to cold shutdown are described in reference 1.
The applicant provided additional information in references 2 and 3.
Our review of the Pebble Springs design with respect to our requirements in reference 1 follows:
The Pebble Springs design must have the capability to cooldown to the decay heat removal system cut-in temperature with only safety-grade systems,
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assuming the most limiting single failure and loss of offsite power, in approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. To meet this requirement, two additional modulating atmospheric dump valves (MAD valves) were added to the main steam system for additional heat removal capacity. The Pebble Springs design now consists of two MAD valves connected to cach steam generator.
Each valve will be seismically qualified for manual local handwheel operation 'such that a single failure of one MAD valve, and the loss of offsite power, will not preclude the cooldown of the plant in a reasonable time using both steam generators. The applicant has coninitted to demonstrate to our satisfaction that a controlled cooldown can be accomplished by handwheel operation of the PAD valves.
We require that the Pebble Springs design have the capability to depressurize the reactor coolant system with only safety-grade systems, assuming a single failure and loss of offsite power.
The applicant has made a preliminary evaluation of depressurizing the reactor coolant system using the emergency cbre cooling system. As the reactor coolant temperature is reduced by heat removal through the steam generators via the MAD valves, the water volume will contract and the pressurizer water level will decrease.
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The emergency core cooling system would be used to pump borated water from the Borated Water Storage Tank into the reactor coolant system, causing the pressurizer to fill with cooler water and thus create a pressure reduction.
The applicant comits to confinn this depressurization process in a manner acceptable to the NRC staff.
We require that the Pebble Springs design provide the capability for boration with only safety-grade systems, assuming a single failure, and a loss of offsite power.
Boration requirements must be met assuming the most reactive control rod is stuck out of the core. The applicant comits to provide this capability to our satisfaction.
The applicant agrees to reference approved prototypical tests to confirm i
adequate baron mixing and natural circulation cooling. Specific details will be developed to the satisfaction of the staff. The applicant also commits to provide operating procedures for the natural circulation cooldown process at the operating license stage. These procedures will be detailed to our satisfaction.
The applicant has discussed the Pebble Springs capability for the collection and containment of decay heat removal system relief valve discharge for an appropriate spectrum of overpressure transients.
Considering 1) the above design modification and the evaluation by the applicant of the cooldown, depressurization and boration mechanisms and i
- 2) their commitment to provide a capability which meets our requirements I
-3 to our satisfaction, we find the Pebble Springs preliminary design acceptable with respect to our cold shutdown requirements.
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o UNITED STATES
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WASHINGTON, D. C. 20555 x
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V OCT 2 G B78 Docket Nos. 50-329/330 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for LWRs, DPM FROM:
Frank Schroeder, Acting Assistant Director for Reactor Safety, DSS
SUBJECT:
SUPPLEMENTAL SECOND ROUND QUESTIONS - MIDLAND PLANT UNITS 1 AND 2 Plant Name:
Midland Units 1 and 2 Docket Numbers:
50-329 & 50-330 Milestone Number:
12-21 Licensing Stage:
OL Responsible Branch LWR-4 and Project Manager:
D. Hood Systems Safety Branch Involved:
Reactor Systems Branch, Analysis Branch Core Performance Branch Description of Review:
Round Two Questions Requested Completion Date:
11/1/78 Review Status:
Incomplete
'N We have reviewed the applicant's responses to First Round Questions on
(.y theMainSteamLineBreak(MSLB)andhavemetwiththeapplicant(September 20, 1978) to discuss the stuck rod assumptions in the Midland MSLB analysis. Our position on the stuck rod assumptions and request for additional information regarding the MSLB analysis are enclosed.
In addition, a position regarding ECCS recirculation testing in accordance with Regulatory Guide 1.79 was inadvertently onitted from RSB Round T40 Questions, dated September 28, 1978, and is also enclosed.
h Assistant Director Fraa Schr for Reactor Safety Division of Systems Safety k
Enclosure:
Supplemental Second Round Questions
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cc:
S. Hanauer D. Hood S. Newberry R. Mattson T. Novak Z. Rostoczy F. Schroeder S. Israel P. Norian S. Varga G. Mazetis S. Salah K. Kniel l
Contact:
Scott Newberry, NRR 49-27341
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211-39 211.0 REACTOR SYSTEMS BRANCH 211.166 Your response to first round question 222.1-2 is insufficient.
(150)
We requested a description of the detailed calculational method used, however, Section 15.1.5.3.2 of the FSAR provides 4
only a brief description of TRAP-2 code with reference to RADAR code. Also, recent discussions indicate that the Midland steam 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 distribution 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 steaaline break analysis.
211.167 Describe how all input parameters were obtained, including the (15D) initial values. Other computer codes used to generate input variables should also be identified.
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t 211.168 Describe how the radial, axial and local power distributions (150) were calculated and used in the RADAR code.
First round question 222.1-5 requested transient axial and radial power distributions instead of design peaking factors. Provide the answers to this question.
211.169 Provide a detailed description of how the radial, axial and (15D) local hot channel factors are applied in the RADAR code for the hot channel and the core average channel.
Describe how time dependence of the peaking factors is taken into accc,unt.
211.170 The nodalization diagram show on Figure 15D-1 does not include (15D) dead volume in the reactor vessel upper head.
Justify that the use of this volume is not necessary in the modeling of the steam line break analysis.
Describe how flashing in the primary system following emptying of the pressurizer is handled.
211.171 Describe how the pressure drop and coolant flow rates through the (150) hot channel were obtained and used in the RADAR code.
211.172 In addition to the total time dependent reactivity feedback, (150) provide each component of reactivity feedback (Moderator, Doppler, rod worth, boron injection).
4 211.173 Provide the core average coolint den'siEyI ~a~nd'co're avda~o^e~.-
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(15D) boron concentration for the first 15 seconds for both BOL and E0L conditions from full power.
211.174 Provide a detailed description of the borated water flow path (15D) into the core follow.ng a steamline break accident including a discussion of the boron transport delay time.
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211.175 Your response to question 211.48 with respect to demonstration (14.2) of ECCS recirculation flow from the reactor building sump to (RSP) the Reactor Coolant System in accordance with Section C.l.b(2) of Regulatory Guide 1.79 is not acceptable. We require that i
you perform or reference tests which verify vortex control, 4
j available net positive suction head'and acceptable pressure drops across screening, suction lines and valves, during the recirculation mode of ECCS operation. Temporary holding y
facilities and/or scaled testing may be appropriate if suitably justified.
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