ML19319D581
| ML19319D581 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/10/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Rodgers J FLORIDA POWER CORP. |
| References | |
| NUDOCS 8003170751 | |
| Download: ML19319D581 (13) | |
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FWKaras, DRL (2 DRL Reading HJFaulkner, DRL l
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FSchroeder, DRL Docket No. 50-302 RWKlecker, DRL TRWilson, DRL RSBoyd, DRL RCDeYoung, DRL APR 101972 DJSkovholt, DRL Florida Power Corporaties HRDenton, DRL ATTN Mr. J. T. Rodgers ECCase, DRS Emelear Project Manager
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P. O. Boa 14042 St. Petersbarg, Pierida 33733 RBMinogue, DRS n<d; PWR Branch Chiefs 1,-
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On the basis of our soatinuing review of the Final Safety Analysis Report
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additional information to semplete our evaluation. The specific information required is listed in the enclosure.
We resogaise that some of this information may already have been placed in the publie record in the sentent of our review of the Preliminary Safety Analysis p.y 4
, Report (PSAR) er of similar features of other facilitiies. [To the.ortent 13,]
E sppliamble, you may inaerporate such information in your applicaties by.
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-To mair.tain the present review sehedule, we will amed your reply by May 8,1972. '
please inform us within seven (7) days after receipt of this letter as to the date when you will be able to submit the requested information to us se that we may revise our schedule, if assessary.
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Please eestact us if you desire any diseossion or clarification of the material ,':(
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R. C. DeTeung, Assistant Director J,
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1 Florida Power Corporation ATTN:
Mr. S. A. Brandimore l
Vice President and General Counsel P. O. Box 14042 l
St. Petersburg, Florida 33733 l
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REQUEST FOR ADDITIONAL INFORMATION FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET No. 50-302
1.0 INTRODUCTION
AND
SUMMARY
1.1 The safety review of the Crystal River Unit 3 Nuclear Steam Supply System (NSSS) is based in large extent on a comparison with other Babcock & Wilcox units which have been reviewed previously. For Table 1-1 of the FSAR, tabulate those design parameters for the Oconee I and Three Mile Island I nuclear units which are dif ferent from the parameters for the Crystal River Unit 3.
For each difference, summarize the safety significance including any considerations of changes of safety j
margins for the Crystal River Unit 3.
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1.2 For any component within the reactor coolant pressure boundary, the engineered safety feature systems, the steam and power con-version system, and other systems associated with radiological safety that has been designed or fabricated outside of the United States, provide the following information:
L.2.1 Identify the manuf acturer and describe his qualifications, experience in the construction of nuclear power plant com-ponents, and experience in furnishing components for nuclear power plants in the United States.
1.2.2 Describe the steps that assure that the quality levels achieved in the fabrication of foreign components are acceptable.
1.3 Recent experience indicates that the bodies of valves and other cast compcnents important to nuclear safety may have areas where the wall thickness is less than required.
Describe the quality control procedures that you are using to verify wall measurements to demonstrate that such components meet design requirements. If such verification can not be demonstrated, provide an engineering justification to support the acceptance of the corponents.
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. L.4 With respect to your quality program, the quality assurance aspect, Level 2, during the operations, maintenance, and refueling phase is treated only briefly in the FSAR. Level 3, the quality surveillance aspect, is omitted entirely from the program. Provide a discussion that treats the quality program during the operational life of the plant in sufficient depth to clearly describe the functions and re'sponsibilities of the organizations comprising the program. Specify lines of authority between quality program, plant and corporate organi-zations. Specify the frequencies of review and audit and the criteria which will be used to judge operational plant and personnel performance. Justify the deletion of the surveillance aspect of the program.
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3-1.0 H F.A(*TO R 3.2 For this type of reactor plant, Oconee Unit I has been tentatively accepted as the prototype for the reactor internals vibration testing program, by the Regulatory Staff.
3.2.1 With reference to Safety Guide 20. Vibration Measurements of Reactor Internals, discuss the vibration testing program which you propose for Crystal River Unit 3.
3.2.2 Provide justification for those aspects of Safety Guide 20 with which your program does not comply.
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. 5.0 CONTAINMENT SYSTEM AND OTHER SPECIAL STRUCTURES 5.28 previous programs for tendon surveillance that have been recently accepted for licensed f acilities have met the follow-ing guidelines which the proposed Crystal River Unit 3 program (see Section 15.4.3.1) does not address and/or meet. Present justification for those guidelines that will not become a part of the tendon surveillance program for Unit 3.
5.28.1 Specify by number or other identification the three tendons in each of the three group's to be included in the surveillance p ro gram.
5.28.2 Specify that the three dome tendons selected will be approximately 120 degrees apart.
5.28.3 Specify that at each surveillance period there will be one l
tendon in each group detensioned and one previously stressed wire removed for inspection.
5.23.4 Specify that three tensile tests will be performed on each of the wires removed, and that the test samples are to be cut from each of the two ends and the center portion of the length of wire.
5.28.5 Specify that the three detensioned tendons will be retensioned to the stress level measured at lift-of f and then checked by a final IIft-off reading.
5.24.6 Speelfy the frequency of tests to be 1 and 3 years af ter the initial structural acceptance test with the interval thereaf ter to be 5 years or as modified, based on experience.
I 5.28.7 Specify that a report will be submitted which quantitatively describes the inspection results. The report should address the broken wires, force-time trend line along with the predicted bounds and any changes in tendons or sheathing filler material.
5.29 Describe the measures which have been instituted to assure that adequate seismic input, including any necessary feedback f rom structural and system dynamic analyses, is specified to vendors of purchased Class I (Seismic) components and equipment.
Identify the responsible design groups or organi-zations who will assure the adequacy and validity of the analyses and tests employed by vendors of Class I (Seismic) components and equipment. Provide a description of the review procedures to be utilized by each group or organization.
. 10.0 STEAM AND POWER CONVERSION SYSTEM 10.1 Provide a flow and instrument diagram of the Emergency Feedwater System.
10.2 Describe the locations of both emergency feedwater pumps and describe how the electrically driven pump can be manually con-nected to the diesel generator busses or to battery power.
Estimste the time associated for such transfers of power sources.
10.3 The turbine stop valves serve: t) as containment isolation valves and, b) for isolation of the unaf fected steam generator for a s team linc break accident. Provide information on the design, operation, including closing time, inspection and testing cf these valves relative to the above functions. Discuss the manner in which the control system that closes these valves meets IEEE-279 Criteria.
10.4 Technical Specification 15.4.8.2 addresses the periodic testing of main steam stop valves.
10.4.1 Specify the leakage rate and testing frequency of these valves.
10.4.2 The closure time of 15 seconds for these valves is unsupported.
Provide the valve closure time used in the Accident Analysis which is referenced as 'the bases for this Technical Specification.
10.5 From the FSAR it is not completely clear as to the amount of energy the steam system has been designed to dissipate. Provide a tabulation of the relief capacity for:
a.
The s team bypass to the condenser l
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The controlled atmospheric relief valves c.
The atmospheric safety valves d.
The total energy dissipation capability of all of these.
10.6 Provide an evaluation of a turbine trip without steam bypass to the condenser.
In the evaluation demonstrate that the emergency feedwater pump capability and the condensate storage capacities are sufficient for residual heat removal during emergency conditions.
. 10.7 Figure 10-3 of the FSAR indicates that a portion of the main steam piping is designed to Class I (Seismic) requirements and the remaining piping to Class III (Seismic) requirements. From the material of Section 5 of the FSAR it appears that the main steam isolation valves (turbine stop valves) and the turbine building are not designed to withstand seismic loads. Provide a complete and detailed analysis of the ability to safely and orderly shut down the plant when the plant with this steam nystem Ls subject to ground accelerations of the same magnitude na the design bases for your Class I (Seismic) structures and systems. As an alternative to this detailed analysis, identify those design changes which you will make to provide full protec-tion to the steam system f rom the design basis earthquake.
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. 11.0 RADI0 ACTIVE WASTE AND RADIATION PROTECTION 11.11 Instrumentation for prompt detection of gross fuel failure is described in the FSAR with the intent to satisfy an area of ACRS concern.
It is stated that "this instrumentation should be' capable of rapidly detecting fuel failure in the presence of fission products already in the coolant due to ' leakage' through the clad and other normally expected sources and to scram on the signal." From our review, the instrunentation does not appear to fulfill these requirements. In regard to this, provide the following:
11.11.1 The detection capability expected in terms of the number of failed fuel rods. Relate this capability to the situation where acceptable leakage, i.e.,1% failed fuel, has occurred and other normally expected sources such as " crud bursts" have also occurred.
11.11.2 The indication, alarm, and readout locations associated with the system and ita instrumentation.
11.11.3 The action to be taken upon determination that " gross" fuel failure has occurred. Specify the minimum criteria suen as ins trument settings and activity levels for certain isotopes required to take this action. Provide the maximum time after the event has occurred in which this action will be initiated.
11.11.4 Provide the action to be taken upon outage of this system.
12.0 CONDUCT OF OPERATI(IIS 12.1 Provide the following information with regard to the corporate level technical staf f that will support the operation of Crystal River Unit 3.
12.1.1 Identification of the organizational group (s) that vill provide this support.
12.1.2 Identification of the organizational responsibilities and lines of authority for the group (s) identified in 12.1.1 above.
12.1.3 The numbers of personnel in the group (s) defined above and a general summary of their qualifications (previous experience, training and education). Nuclear reactor experience is of special interest.
12.2 Designate the specific succession to responsibility for overall operation of the facility in the event of the absence of the Nuclear Plant Superintendent and the Assistant Nuclear Plant Superintendent.
12.3 Describe the duties, if any, that the Nuclear Plant Superin-tendent or any of the staf f shown in Figure 12-3 has for the operation of Crystal River Units 1 and 2 12.4 The minJmum qualifications for positions on the plant staf f are not adequately described in the FSAR. State your intentions as I
to meeting the minimum qualifications of ANSI N.18.1, " Selection and Training of Nuclear Power Plant Personnel" for all plant personnel.
12.5 At the plant level, specify the position responsible for the
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administration and evaluation of the training program.
12.6 Describe your provisions for the periodic retraining of the plant s taf f.
12.7 The information presented in Section 12.3 regarding emergency plans is insufficient. Submit a revised, comprehensive emergency plan that meets the requirements of 10 CFR 50, Appendix E.
12.8 Provide a list of the titles of all safety-related procedures to be developed for plant operations.
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. 13.0 INITIAL TESTS'AND OPERATION 13.1 Your initial tests and operation program is not in accord with the acceptable positions described in the Guides for the Plan-ning of Preoperational Testing and Initial Startup Programs, dated December 7,1970. Modify your preoperational and startup programs to comply with the positions of these guides or with positions that provide an equivalent degree of protection.
13.2 Review of the qualifications of your proposed operating staf f indicates that they are well educated and trained and strong in fossil plant operating experience, but they are not deeply experienced in nuclear plant operation.
Consequently, your normal plant operating staff will need to be augmented during the initial tests and operation period for about six months.
In this regard describe the following:
13.2.1 The organizational functions, responsibilities, and authorities of the various greups that will augment the plant operating staff during preoperational and startup testing.
13.2.2 The functions, responsibilities, and authorities of key posi-tions in the organizations identified above.
13.2.3 The qualifications of the appointees to the positions above.
13.3 Describe your system for preparing, reviewing, approving, and executing test procedures; for making any necessary changes to a test procedure while performing the test; and for evaluating, documenting, and approving the test results.
, 14.0 SAFETY ANALYSIS 14.6 For the steam line failure accident:
14.6.1 Indicate the time sequence for significant events such as initiation of accident, reactor trip, turbine valve closure, feed water valve closure, steam bypass valve operation, safety and relief valve operation, and high pressure injection actu-ation.
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14.6.2 in addition to the time-history traces of Figure 14-22 of the FSAR, provide additional traces of DNBR, pressurizer level, pressurizer pressure, af fected s team generator pressure, and l
unaffected s team generator pressure.
i 14.6.3 We note that you have analyzed this accident assumming certain operator actions in one case and without operator action in another case. Provide a detailed clarification of the actions which occur and their timing both with and without operator action. If the time-history traces of Figure 14-22 of the FSAR are not for the worst case, provide a set of traces for the wors t case. Discuss the effects of the dif ferences in action on 1
the potential consequences of the accident such as fuel failures and radiological dose.
14.6.4 For the worst case above extend your analyses to include the event where the tripped rod worth is the minimum available with the rod of maximum worth stuck out, and indicate whether the reactor can become critical and achieve significant power during the transient.
14.6.5 Evaluate the effect of the secondary system pressure on steam generator leak rate.
If a constant leak rate was used, revise Table 14-21 to reflect the increased leak rate due to the reduced secondary system pressure. State the langth of time fo::
the primary to secondary leakage to be terminated.
14.6.6 We assume the doses and release activities presented in Table 14-21 of the FSAR assume complete loss of condenser vacuum.
If not, revise Table 14-21 to include this assumption.
14.6.7 Indicate each term used in the calculation of the doses and release activities in Table 14-21.
If steady s tate decontam-inatica factors are used, justify their validity under accident conditions.
. 14.7 Two loss-of-electric-power situations are considered in the FSAR. Provide time-history traces of the parameters shown on Figure 14-22 of the FSAR and specified ir. request 14.6.2 above for each situation.
14.8 For the loss of all a-c power discuss the capability of safely shutting down the plant. Include a list of all necessary components, control systems, instrumentation, and other electrical systems that must be supplied frca the station batteries. Discuss how power will be provided from these batteries. Estimate the power consumption of each component and Indicate the maximum time the plant could maintain this condition safely.
L4.9 For the loss of flow accident:
14.9.1 Specify the Doppler and moderator coef ficients used in the analysis.
14.9.2 Provide assurance that the reactor coolant pumps, piping and restraints have been designed in such a manner as to prevent the locked rotor accident from initiating a more serious accident.
14.10 The rod drop accident was analyzed assuming automatic run back to 60% demand at 30% per minute. Figures 14-20 and Figures 14-21 show automatic runhack at 50% per minute.
Please reconcile this inconsistency.
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