ML19319D576
| ML19319D576 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/02/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Rodgers J FLORIDA POWER CORP. |
| References | |
| NUDOCS 8003170745 | |
| Download: ML19319D576 (7) | |
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Docket File DRL Reading PWR-4 Reading S. H. Hanauer, DR F. Schroeder, DRL T. R. Wilson, DRL
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S. %vd, DRL R. C. DeYoung, DRL Docket No. 50-302 VAY 2 572 R. Tedesco, DRL D. J. Skovholt, DRL H. R. Denten, DRL i
RPS & S&RS B' ranch Chiefs E. G. Case, DRS Florida Power Corporation.,
R. R. Maccary DRS ATTN:
Mr. J. T. Rodgers K. F. Knuth, DRS Assistant Vice President &
PWR Branch Chiefs Nuclear Project Manager R. W. Klecker, DRL P. O. Box 14042 OCC St. Petersburg, Florida 33733 CO (3)
H. J. Faulkner, DRL Centlemen:
F. W.
Karas (2)
D. Davis, DRL On the basis of our continuing review of the Final Safety Analysis b port for Crystal River Unit 3 Nuclear Generating Plant, we find that we need additional information to complete our evaluation. The specific information required is listed in' the esi&losure.
We recognize that some of this information may already have been placed in the public record in the context of our review of the Preliminary Safety Analysis Report (PSAR) or of similar features of other facilities. To the extent applicable, you may incorporate such information in your application by reference.
To maintain the present review schedule, we will need your reply by June 5, 1972.
Please inform as within seven (7) days af ter receipt of this letter as to the date when you will be able to submit the requested information to us so that we may revise our schedule, of necessary.
Please contact us if you desire any discussion if clarification of the meterial Fequested.
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R. C. Deyoung R. C. DeYoung, Assistant Director for Pressurized Water Reactors Division of Reactor Licensing
Enclosure:
8003170 h {
Request for Additional Information ee: See attached 4
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2-Florids Power Corporation ec - enc 1:
Florida Power Corporation ATTN:
Mr. S. A. Brandimore Vice President and General Counsel P. O. Box 14042 St. Petersburg, Florida 33733
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- ,e. Form ABC-Sls (Rev. 9-53) AECM 0240
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!!!O11ti:ST FOR ADDTTTONA1. TNFORi4ATTON F1,ORTDA POWER CORPORATTON CHYSTAT, RTVER !! NIT 3 DOCKET NO. 50-302 1.0 TNTRODtiCTION AND SLTARY 1.5 Tilentify the possibic interactions between Crys tal Fiver l' nit 3 anel the two fossil fired.,lants, and evaluate those interactions of safety significance.
. 6.0 ENGINEERED SAFEGUARDS 6.1 Descrlhe the inservice inspection progran for fluid systers other than those composing the reactor coolant pressure boundary, includinn itens to be inspected, accessibility requirenents, and the frequency and types nf inspection. The fluid systems to he considered are applicable engineered safety systems, reactor nhutdown systems, cooling water systems, and the radioactive waste treatment systems.
6.2 The practice of permitting small dinneter piping for safety related systens to be " field run" should be limited insofar as it is practical to do so.
When it is permitted (1) stringent quality assurance measures should be taken to assure that the installation has been performed in such a manner that the assumptions made for design mad safety assessment purposes remain vn] Ed, and (2) tests should be performed on the completed item to provide a final indication of acceptability.
In view of these requirements, provide the following information:
6.2.1 A discussion of the extent to which you pernitted " field running" of piping for safety related systems, especially for engine 0 red safety features.
6.2.2 A description of the special quality assurance measures and per-formance testa that were conducted to assure satisfactory i ns tal lat i on.
6.2.1 Provisions for incorporating " field run" piping location and construction details on "as built" drawings.
6.3 Describe the preoperational test program that -ill be performed on the engineered safety features to demonstrate that these safety systens will perform their design functions under acci-dent conditions.
6.4 Describe the systems, including flow and instrument diagrams,
that are provided for adding borated water and nitrogen to the core flooding tanks. Discuss the mode of operation of these systems including the actions required for filling the tanks while the reactor is at power.
6.5 Section 6.1.] of the FSAR indicates that both core flooding tanks are requi red to provide 1007. core protection for inter-mediate and large reactor coolant system pipe f ailures.
Describe the degree of core protection provided by only one of the two core flooding tanks.
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6.6 Discuss the degree of protection of critical engineered safety features, such as ECCS and reactor building spray nump motors, from flooding.
.6.7 in Appendix 14.B. of the FSAR containment purging in evaluated a means of controlling pos t-accident hydrogen accumulation.
as For this application the hydrogen purge system performs as an engineered safety feature. In regard to this:
6.7.1 niscuss the design and construction standards for the hydrogen purge system ann their conformance to engineered safety feature standards.
6./.2 th scribe the requirements for accident environnent testing, redundancy, periodic functional tes ting, and single failure analysis.
6.8 Describe the means of detecting coolant leakage into the auxiliary building from the engineered safety feature systens during post-accident recirculation operation.
6.9 For the reactor building emergency cooling system, describe the verification testing that has been or will be performed on com-ponents of this system under the combined environmental effects of high humidity, pressure, temperature, radiation, and applied chemical concentrations. Indicate whether an assembled system of fans, motors, coolers, and filters will be tested under these environmental conditions.
6.1D Discuss the net positive suction head (NPSH) that is available for reactor huilding spray and low pressure injection pumns l
us ing the assumptions of Safety Guide No. 1 (published November 2,- 1970), and identify the margin between the necessary NpSH and that which is available. I f you are unable to demonstrate that suf ficient NpSH is available using these assumptions, state the extent to which containment pressure is relied upon to achieve the necessary NPSH.
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. 7.0 TNSTittfMI'NTATION AND CONTROL
/.1'l The I5 Alt descrlhes brief1v the capabill tv for prompt hot shut-dvun an.I subsequent cold shutdown of the reactor from outs Ide the control room.
In regard to th is, provide the followine:
/. l').1 A detailed explanation of the actions and their timing that would he taken to af fect both hot and cold shutdown from outside th e control room.
7.19.2 The locations of all auxiliary control stations that would he used in these actions. Include as stations the location of valves which need to he manually adjus ted.
7.20 To shut down the reactor from outside the control room, section 7.4.6 of the FSAR states that eight operations reauiring 1 or 2 minutes are performed initially from inside the control room.
I t appears that some of these operations can be performed fron outside the control room.
Although it may be nreferable to execute some of these actions from inside the control room, speci fy those actions and the associated time span which mus t be performed from inside the control room prior to evacuation if such an emergency should arise.
7.21 providn a discussion of the availability of your communication facilities, both internal and external to the plant, in the event of a loss of AC power.
Indicate the locations of the plant's paging and handset stations that will be functional under loss of AC power.
7.22 Tdentify the information readouts or indications provided to the operator for monitoring conditions in the reactor coolant system, safety related systems, and containment for anticipated operational occurrences, and accident and pos t-accident conditions. Your response should include the design criteria and associated bases, type of readout, number of channels pro-vided, range, accuracy, location and a discussion of the ade-quacy of the desian for the full spectrum of nostulated accidents.
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. 14.0 S A FETY ANAI.YS TS 14.II The results of the locked rotor accident presented in Fictire 14-14 of the FSAR nre identical to those renorted for the Oconee pinnts; however, your assumption of the leneth of time, 0.1 seconds, for the rated flow to decrease from 100% to 75% is significantly less than the 2 seconds used in the Oconee analysis. Resolve this disparity of seemingly identical results using different flow assumptions. Also, justify this assumed rf me for whatever value you select.
14.12 Evaluate the possibility and subsequent consecuences of nine whip from a ruptured tube in the steam generator precipitating other tube failures.
- 14. I 'l Using the maximum positive moderator temnerature coe ficient allowed by your Technical Specifications, +0.5 x 10 {(ak/k)/*F,
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specify the percent of rods experiencing Dtm for the rod ejec-tion accident at rated power using a rod worth or 0.65% ok/k.
14.14 l'rovide a plot, similar to Figure 14-3 of the FSAR, of peak thermal power versus rod withdravel rate for the rod withdrawal accident at rated power.
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