ML19329C610
| ML19329C610 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/12/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19329C606 | List: |
| References | |
| NUDOCS 8002180127 | |
| Download: ML19329C610 (16) | |
Text
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SAFETY EVALUATION (CONTAIWEST 3YSTEMS)
DAVIS-BESSE NUCLEAR POWER STA L' ION. UNIT 1 DOCKET NO.50-34n 6.2 Containment Sys tems 6.2.1 Containment Functional Design The containment system for the Davis-Besse Nuclear Power Station, Unit 1 includes an A5ME Code,Section III, Class B, f ree-s tanding steel containment vessel surrounded by a reinforced concrete shield building, con tainment heat removal sys te.n, containment i s ola tion sy s tem, combus tible gas control sys tem, and shield building ventila-tion system.
1,92',000 Tha c raal cen tain an t va = =al ha c a nat frae voluma cubic feet. The containment vessel houscs the nuclear s team supply system, including the reactor, s teaa genera tors, reac tor coolant pumps and pressurizer, as well as certain components of the plant's enaineered safe ty feature systems.
The containment vessel is de-signed for an internal pressure of 40 psig and a temperature of 264 F.
The applicant has described in the Safety Analysis Report the methods used to analyze the containment pressure response to pos tulated loss-of-coolant accidents and reported the results. Various break locations and sizcs were evaluated to determine tha t a 14.1 f t-, hot l
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The applicant has analyzed the containment pressure response to postulated loss-of-coolant accidents in the following manner.
The Babcock and Wilcox CRAFT computer code was used to calculate mass and energy releases to the containment during the blowdown, core reflood, and post-reflood phases of the accident. The mass and energy addition rates calculated in this manner were then used as input to the COPATTA computer code to calculate the contain-ment pressure response.
As described above, tce CRAFT code was used to calculate blowdown mass and energy elcases. The blowdown phase of the accident is the phase during which most of the energy contained in the reactor coolant sys tem, including the s tored energy in the wa ter, me tai and core; is released to the containment. To obtain a conserva-tively high energy r'elease rate, the applicant assumed nucleate f
boiling in the core until the quality of the coolant was approxi-mately 1.0, and full ECCS operatic.i.
The CRAFT program was also used by the applicant to predict mass and energy releases to the containment during the core reflood c
phase of the accident. The reflood phase is important when analyzing postulated pipe rupturcs in the reactor coolant system cold legs since the steam and entrained liquid carried ou't of the core for these break locations can pass through the s team generators and be suoerheated to the temperature of the steam generator secondary l
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fluid. During core reflood the carryout rate fraction, which determines the amount of steam and entrained water leaving the core and therefore the amount of energy that can be transferred from the s team genera tors, is calculated based on a correlation inherent in CRAFT. CRAFT calculates average carryout rate frac-tions in excess of 0.S.
Results of the FLECHT experiments indi-cate that the carryout fraction of fluid leaving the core during reflood is about 80'. of the incoming flow to the core, which con-firms the CRAFT approach. The rate of energy release to the con-tainment during,this phase is proportional to the flow rate into the core, and thus through the s team generators.
Af ter the core is completely co ered with water, decay heat generation will produce boiling in the core and a 2-phase mixture of s team and water will exis t.
This mixture can enter the s team generators and superheated steam will be generated. The appli-cant's analytical model accounts for this additional energy.
About 500 seconds af ter a large break accident essentially all of the available sensible heat is removed f rom the primary sys tem and i
the steam generators.
f The CRAFT computer program has been accepted by the NRC for l
calculating mass and energy releases to the containment during the blowdown phase of the postulated accident. However, in applying t
the CEMT mh to the reficad an:! post-re flood rh ms of n^
W ?u loss-of-coolant acciuta.;;
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following blowdown, t;w aget.cani e
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. included the quenching action of the ECCS fluid on the exiting steam. We will require that this effect be neglected. The appli-
- cant has committed to providing a reanalysis of the design basis
-loss-of-coolant accident assuming no quenching. The results of this analysis will be reported in a supplement to the Safety Evaluation Report.
We have performed a confirmatory containment analysis for a postulated cold leg (pump suction) break based on the mass and energy release data for a similar plant which neglected the quenching effect. Using the CONTEMPT computer code (References 1 and 2), we calculated a peak containment pressure of 35.8 psig.
The Davis-Besse 1 containment vesset is designed for a maximura pressure of 40 psig. Although we do not expect the peak calculated pressure to change significantly using revised mass and energy re-lease data, we will defer our conclusions on this plant based on the applicant's containment analysis until additional information is received.
The applicant has also analyzed the containment pressure respense to a postulated main steam line failure. The applicant calculated a peak containment vessel pressure of about 22 psig for this accident.
l The soplicant has not comniated tSe nrecer c -acncr=e n.nsly=is o' the containment vessel intericr ccmpartm?nts, seca la tae r c ] c. n-w a.,
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- vessel cavity, the steam generator compartments and the primary shield pipe penetration annulus. The applicant has committed to providing the results of the analysis and the applicable mass and energy release data. We will report on this in a supplement to the Safety Evaluation Report.
We have evaluated the containment system functional design in accordance with the Cencral Design Criteria stated in 10 CFR Part 50 of the Commission's Regulations and, in particular, Criter!1 16 and 50.
However, before we can conclude that the contaidment vessel and interior compartment design pressures are adequate, we will need revised mass and energy release data for the containment vessel anslycir 'thich doce not include the quenching cetion of the ECCS water on the exiting steam following blowdown, revised mass and energy release data for the containment vessel interior compart-ment analysis that adequately describes the blowdown for each pos tu-lated pipe break over the time scale of interes t, and the results of the containment vessel and interior compartment analyses based on the revised mass and energy release data. We will report our conclusica; on the acceptability of analyses and adequacy of design pressures in a supplement to the Safety Evaluation Report.
6.2.2 Containment Heat Removal Sys tems The containment spray system and the containment air cooling sys-tem are provided to reduce the containment vessel pressure follow-in; po:'"5 a>d S';h energ n 2 br;-
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I air cooling system is also used during normal plant operation, whereas the containment spray system has no normal operating function.
f The containment spray sys tem consists of two separate spray trains of equal capacity. All active components of the system are located outside the containment vessel to facilitate maintenance operations.
Missile protection is provided by direct shielding or physical separation of equipment. The system is seismic Category I.
The containment spray pump recirculation intakes from the containment emergency sump are enclosed by a scrcen assembly to prevent the entry of debris which could clog the spray nozzles.
I A high containment pressure signal from the safety features actuation sys tem will automatically actuate the con tainment spray sys tem.
The system pumps and valves can also be manually cperated from the control room. The spray pumps initially take suction from the borated water s torage tank. When the wa ter in the tank reaches a low lavel, a switchover from injection to recirculation is manually initiated.
The applicant has provided an analysis which demonstrates that sufficient net positive suction head will be available to the spray pumps for both the injection and recirculation modes of operatien.
The analysis performed is consistent with the guidelines of Regu' ate: Gaide 1.1.
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, The containment air cooling system consists of three equal capacity air cooler units. The sys tem components and equipment required to remain operable following an accident are located outside the secondary concre te shield for missile protection at an elevation that precludes flooding, are designed to withstand the dif ferential pressures resulting from a loss-of-coolant accident, and are seismic Ca tegory I.
A high containment pressure signal or a low reactor coolant system pressure signal from the safety features actuation system will automatic $lly actuate the containment dir cooling systcm.
The system can also be manually operated from the control room.
Based on our review of the containment heat removal systems, we conclude that the sys tem designs are consis tent with the require-ments of General Design Criteria 38, 39, and 40, and are therefore accep tabl e.
6.2.3 Secondarv Containment Functional Desien The secondary containment (shield building) is a reinforced concre te s truc ture surrounding the s teel containment vessel.
Potential leakage from the containment vessel to the shield building and adjoining penetration rooms is collected and processed by the emergency ventilaticn system, which is a seismic Category I system.
The emergency ventilation system consists of redundant trains, each carabic of th t'=cci m i r
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eme tgency ventilation system will maintain the areas it serves at a negative pressure to assure the collection of leakage from the containment vessel.
The applicant has attempted to identify potential leak paths from the containment vessel which bypass the volumes treated by the emergency ventilation system. The bypass leak paths identified by the applicant and the total allowable leakage frcm these bypass leak paths have been ir.cluded in the plant technical specifications.
However, all potential bypass leak paths have not been identified.
The applicant has ccm=itted to provide additional informatica re-garding this matter. We will conclude on the acceptability of identifieu potentist bypass leak paths in a supplement to the sate ty Evaluation Report.
The applicant has analyzed the pressure response of the shield 4
building following a postulated less-of-coolant accident.
Basea on our review, we conclude tha t the applicant has underes tima ted the time required to depressurize the shield building and reach a negative pressure of 0.25 in. w.g.
The applicant calculates that l
about 20 seconds is required to establish a negative pressure af ter the emergency ventilation system becomes operational which is cbcu t i
45 seconds af ter the accident; our calculations it -*icate that this will not occur until about 50 seconds af ter the emergency ventila-tion system. bece-*= operational asso-ine ont" one tr'in f= anarable.
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the sys tem to maintain a negative pressure within prescribed limits. We will also require the applicant to verify the time required to depressurize the shield building and establish a negative pressure.
6.2.4 Containment Isolation System The containment isolation system is designed to automatically isolate piping systems that penetrate the containment to prevent outleakage of th'e containment atmosphere following postulated accidents. Double barrier protection, in the forn of closed systems and isolation valves, are provided to assura that ec ringle acti.ve failure will result in the loss of containment integrity.
The containment isolation provisions, including the isolation valving and penetration piping, are scismic Category I.
Containment isolation will automatically occur upcn receipt of containment high pressure signals or reactor coolant system low pressure signals from the safety features actuation system.
High radiation signals are also used to isolate the containment vessel purge sys tem lines.
Based on our review, we conclude tha t the containment isolation system design conforms to General Design Criteria 54, 55, 56 and 37, and tne
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_ 6.2.5 Combustible Gas Control Sys tem Following a loss-of-coolant accident, hydrogen may accumulate inside
. the containment as a result of (1) a chemical reaction between the fuel rod cladding and the steam resulting from vaporization of emergency core cooling water, (2) corrosion of construction materials by the spray solution, and (3) radiolytic decomposition of the cooling water in the reactor core and the contain=cnt sump.
The combus tible gas control system is designed to control the concentration of hydrogen within the containment vessel following a loss-of-coolant accident.
The sys tem consis ts of the contain-ment hydrogen dilution system, hydrogen purge system, recirculation sy s tCm, 7"d ga0 anCl**2cr 373 tJm.
The contain=cnt hydrogen dilu tion sys tem contrsis the hydrogen concentration within the containment vessel by the addition of air.
The system is seismic Category I and consists of redundant trains.
The system blowers have a 100 SCEM capacity.
The maximum pressure that the system blowers are capable of repressurizing the coatain-ment vessel to is 13 psig.
P The hydrogen purge system serves as a backup to the hydrogen dilution system, and consis ts of a sin' '
train.
It releases the containment vessel atmosphere ab s HPEA and charcoal filters to the s ta tion vent. The sys tem 1: setee Lc Ca tegory I.
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ The cow:ainment recirculation system is designed to draw air from the containment vessel dome and discharge it toward the con-tainment air coolers, to provide a more uniform dispersion of hydrogen. The system is seismic Category I and consists of redundant trains. The gas analyzer sys tem is designed to monitor the hydrogen concentration within the containment vessel following a loss-of-coolant accident. The system is seismic Categcry I and consis ts of redundant trains.
Samples can be drawn from four points in the containment vessel.
The applicant has performed.a analysis of the pos t-loss-of-coolant accident hydrogen generation in the containment vessel following a loss-o f-coolan c acciden t chat is consis tenc with the ;;u l u e li c,e 3 ot-Regula tory Guide 1.'.
The applicant calculated that the hydro;cn concentra tion in t'he con tainmen t <ill no t reach the icwer fla=:abi ti ty limit of four volume percent until about 44 days af ter the accident, and that the control limit of thiee volume perecnt eill occur :tcut 24 days af ter the accidcat. The hydrogen ccnccntration in the con-tainment will be maintained below three volume percent by actuatin; trains of the hydrogen dilution f stem ehen the control one of the limit is reached. We have performed similar calculations for the hydrogen generation in the containment following a loss-of-ccolant accident and our results have confirmed thoss of the applicent.
Based on our review of the sys tems orovided for combus tible tas caacrou -. :.: 1 :..,
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conclude that the sys tems conform to the guidelines of Regulatory Guide 1.7 and the requirements of General Design Criteria 41, 42 and 43, and are therefore accep table.
6.2.6 Containment Leakaze Tes tinc Program The containment design includes provisions and features which satisfy the testing requirements of Appendix J to 10 CFR Part 50.
The design of the containment penetrations and isolation valves will permit periodic leakage rate testing at the pressure specified in Appendix J.
Included will be those penetrations tha t have resilient seals and expansion bellcws, such as personnel airlocks, equipment ha tch, refueling tube blind flange, hot process line pene tra tion s and nicc tri:21 pene tra tions.
The proposed reactor con tainment leakage tes ting program complies with the requirements of Appendix J to 10 CFR Part 50.
Such compliance provides adequate assurance that containment in teg ri ty can be verified throughout the service lifctime of the plant and that the leakage rates will be periodically checked on a timely basis to assure that they are within specified 3(mi ts.
Maintaining s
con tainmen t leakage rates within such limits provides reasonable assurance that, in the event of any radioac tivity releases within the containment vessel, the los s-o f-con tainmen t a tmosphere thrcugh potential leak paths will not be in excess of acceptable limits specified for the site.
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. We have concluded that the containment leakage testing program complies with the requirements of Appendix J to 10 CFR Part 50, and that such compliance cons titutes an accep table basis for stisfying the requirements of General Design Criteria 52, 53, and 54 I
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l BIBLIOGRAPHY OF REFERENCE 51ATERIAL
+
1.
R. J. Wagner and L. L. Wheat, CONTEMPT-LT Users Manual, Interim Report I-214-74-12.1, Aerojet Nuclear, August, 1973.
2.
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3.
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5 4
R. C. Schmitt, G. E. Bingham, and J. A. Norbert, Simulated Desien 3 asis Accident Tests of the Carolina Vir2 inia Tube Reactor Containment - Final Report, IN-1403, Idaho Nuclear Corpora tion, Dececher, 1970.
5.
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1 6.
T. Tagami, Interim Recort on Safety Assessments and Facilities Establishment Project in Japan for Period Ending June, 1965 (No. 1), Prepared for the
~ L ~. a. :T. *065 P;tr.ali::ed
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7.
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9.
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i i
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I 11.
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l 12.
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I r
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A. O. Allan, The Radi a ti en P5 --i*- tr-- ~ ~ Un ter -- >
t-"-
ue e'"-4...
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l
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~.
14 ANS Standard A'is-5.1, Decay Enerzy Release Rates Followina Shutdown of Uranium-Fuel Ther-al Reactors (DRMT), American Nuclear Society, Hinsdale, Illinois, Oc tober, 1971.
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