ML19308E072

SER Input from Containment Sys Branch.Containment & Related Sys Acceptable,Pending Demonstration That Structural Capability of Subcompartments Adequate for Calculated Pressures

ML19308E072
Person / Time
Site:	Crystal River
Issue date:	12/20/1973
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML19308E071	List: ML19308E070 ML19308E072
References
NUDOCS 8003200802
Download: ML19308E072 (10)

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DRAFT SAFETY EVALUATION' (CONTAl? MENT SYSTEMS)

CRYSTAL RIVER, UNIT 3 DOCKET NO. 50-302 6.2 CONTAINMENT' SYSTEMS 6.2.1 CONTAINMENT FUNCTIONAL DESIGN The primary containment structure (reactor building) for Crystal River 3 is a free-standing steel-lined, reinforced concrete struc-3 ture with a net free volume of approximately 2,000,000 ft The structure houses the reactor and the primary sys tem including the pressurizer and steam generators, as well as certain ecmponents of the engineered safety features provided for the facility.

The primary containment structure is designed for an internal pressure of 55 psig and a temperature of 231 F.

The applicant has described in the Safety Analysis Report and knendments 27 and 32 the results and methods used to analyse the containment pressure response for a number of design basis loss-of-coolant accidents.

The applicant has analyred the containment for loss-of-coolant accidents for a spectrum of both hot leg and cold leg breaks, up to and including the double-ended rupture of l

the largest reactor coolant pipe to determine the containment pres-sure responses.

Minimum containment cooling was assumed in the 1

analysis and reactor building 1.e., one of the three emergency l

building cooling units, and one of the two spray trains of the Reactor Building Spray System were assumed to operate, and the core reflood energy and s team generator stored energy were included, 8003200

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As discussed below, we have reviewed the.results of these analyses and verified by our analyses that the calculational methods used by the applicant to determine the containment pressure response from postulated loss-of-coolant accidents are conservative.

Mass and energy release rates were calculated using the CRAFT com--

_puter code.

These mass and energy addition rates were than used-as inputs to' CONTEMPT which is the applicant's computer program to calculate the. containment pressure response.

The CRAFT computer code was used by the applicant to determine the mass and energy addition rates to the containment for cold leg breaks during the blowdown phase of the accident; i.e., the phase

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of the accident during which most of the energy contained in the reactor coolant system including the coolant or water, me tal and.

core stored energy is released to the containment.

The applicant has, however, increased the energy release rate to the containment by conservatively extending the time that the core would remain in nucleate boiling; 1.e.,

the time when the energy removal rate from the core is highest.

By using this me thods, the core would trans-fer more heat to the containment for containment analysis than for emergency core cooling analysis

-Since this additional energy release from the core will increase the c:ntainment pressure, the

' calculation is conservative. The CRAFT computer code has been ap-provedbytheAECforcalculatdgenergyreleaseduring'aLOCA.

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D A The applicant has identified the 7.0 f t2 split break at the pump suction as the cold leg break dhat results in the highest contain-ment pressure.. The applicant calculates a peak pressure about 50 psig for this break.

The largest break (S.55 ft ) results in a 2

peak calculated pressure of about 49 psig. We have analyzed the contain=ent pressure response - for the 7.0 f t2 break in the suction leg of the reactor coolant sys tem using the CONTOLDT-LT computer code and included the. energy addition to tne containment from the steam generators.

We calculated a peak containment pressure es-sentially the same as the applicant's.

To detarsine the mass and energy release to the containment, we used the applicat t's blowdown mass and energy release rates calculated by CRAFT and the mass and energy release rates during :e reflood phase of the accident de-termined by our computer program FLOOD 2.

Blowdown mass and energy releases for hot leg breaks were also cal-culated by the applicant using the CRAFT computer code.

The appl'icant's analysis indicates that a 14.1 f t2 break of the hot leg results in

-the highest hot leg break containment pressure of 49 psig.

The applicant has also analyzed the containment pressure response due to postulated failures of a main steam line.

The applican t has conservatively assumed that the energy in a s team genera tor -was instantaneously released and has not taken credit for the stergy

. removal capability of the structural heat sinks.

The applican t calculates a peak containment pressure of 23 psig for this 4 aident.

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i I 'Je have evaluated the containment system in comparison to the Commission's General Design Criteria stated in Appendix A to 10 CFR Part 50 of the Commission's Regulations and, in particular, to Criteria 16 and 50.-

As a result of our evaluation, we have concluded that the calculated pressure and temperature conditions resulting from any design basis loss-of-coolant accident will not exceed the design conditions of the containment structure. The highest calculated containment pressure and temperature are about 50 psig and 230 F, respectively. The containment design pressure of 55 psig provides a 107. margin above the peak calculated pressure.

• Je conclude that the maximum containment pressure is correctly cal-t culated to be below the design pressure and that there is sufficient margin between the maximum containment pressure and the design pres-sure of the containment structure to assure - that the health and i

safety of the public is adequately protected.

I The applicant has analyzed using the FLASH-2 program the pressure I

l response within the containment interior compartnents,_ such as the reactor vessel cavity, the primary shield pipe penetration, and the s team generator compartnents during loss-of-coolant accidents. The l

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- applicant calculates peak pressures of 123.5 psi in the reactor cavi ty,. 1152.2 psi in the oice penetration and 17.5_osi in a steam i

generator compartment.

The applican t's calculated pressures exceed the compartment initial i

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. 5 design' pressures i.e., the design pressure at the construction permit s tage of review, in several' cases.- For these cases,- the applicant has reanalyzed the as-built structural capability of the compartments.

The applicant's design pressures, ts-built capability, and calculated pressures are presented in Table 6-1.

TA3LE 6-1 SUBCOMPARTMENT DESIGN AND CALCULATED PPSSSURES Cecpar==en t Initial As-built

~ Applicant's Design Dif f erential Calculated Differential Pressure Structural Differential Pressure Capability Prassure (psi)

(psi)

(psi) 123.5 Reactor Cavity 170 Pipe Penetrati:n 1200 2000 1152.2 South Stein Generator Ccepar tment 15 30.3 17.5 North Stesa Generator Cocpar t=en t 15 30.3 16.8 We have perfor=ed pressure response calculations using the RELAP-3 program and co= pared our results to the applicants's.

Our results indicate reasonable agreement with the applicant's. We conclude tha t the applicant's calculated design differential pressure for the subcompartments are acceptable.

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2 6-6.2.2-REACTOR 2UILDING MEAT REMOVAL SYSTEMS' The Reactor Building Spray System (RSSS) and the Reactor Buidling Emergency Cooling System (RSCS) are provided to recove heat from the containment following a loss-of-coolant accident. Any ofL the following combina*. ions of eculpment will provide adequata heat removal capability:

(a) both spray trains of the RSSS, (b) three fan-cooler units of the RSCS, and (c) one spray train of the R3SS and two fan-cooler units of the RSCS.

The RSSS serves only as an engineered' safety features and herform no normat operating function.

It is a seismic Category-1 systam con-r'. sting' of redundant piping, valves, pumps and spray headers. All active components of the R3SS are located outside the reactor building.

Missile protaction is provided by direct shielding or physical se-paration of equipment.

The reactor building sump screen assembly is designed to prevent debris from entering the spray system that could clog the spray no: les.

The RBSS includes a system for injecting sodium thiosulfate into the spray water for iodine removal fran the containment atmosphere-following an accident, and a system 'for injecting sodium hydroxide into the spray water for. pH adjustment.

The sodiud hydroxide will raise the pH of the spray water into the alkaline range.

Both systems are designed to pennit gravity draining of the solutions into the ' spray pump suction piping.

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A high reactor building pressure signal ~.from the engineered safety features actuation system will automatically place the RSSS in

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operation.. The spray penps will initially take suction from the borated water storage tank. When the water in the tank reaches a low level, the spray pump suction' is manually transferred to the reactor building' sump..

The Reactor Buidlin7 Cooling System (R3CS) is used during both no: mal and accident conditions.

Three equal capacity fan-cooler units are provided. Lach unit contains a coisture -seperator, a cooling coil, and a two-speed f an.

Under pos t-accident energency cooling conditions, the unit will operate at a reduced ' speed.

Under normal plant operating conditions, water from the industrial cooler is circulated through the cooling coils.

Under accident

' conditions, following receipt of an engineered safety features actu-ation signal, the high speed portion of the air handling units are -

de-energi:ed and the slow " eed portion of the air handling-units are energized.

For e=ergency cooling, heat will '_e rejected to the nuclear services closed cycle cooling system.

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The RBCS is a seismic Category I sys tem.

The housirrs for the cooling units and the supply ducts are designed to withstand an inward pressure differenti.al of 2 psi..The system was analyzed by the applicant to determine that the duct sizes and outlet locations are such that a two pound differential is not exceeded during the transient period.

The cooling units are located outside the secondary concrete shield for missile protection. The R3CS equip-cent is accessible for periodic testing and inspection during nor-

=al plant operation.. We have reviewed the containment heat recoval sys tems for 'conformance to General Design Criteria 33, 39, -and 40, and Safety Guide L.l. We conclude that the sys tems are acceptable and meet the intent of the GDC.

6.2.3 CONTAIWDTP ISOU. TION SYSTE'4S The Reactor Building Isolation System is designed to isolate the containment atnosphere from the outside environment under accident conditions.

Double barrier protection, in the form of closed systems and/or isolation valves, is provided so that no single, credible failure or calfunction of an active component can result in loss-of-conta4-ent' integrity. Reactor building penetration piping and the associatad isolation valves are designed as Category I (seismic) equipment, and are protected against tsiles which could be gener-ated under accident conditions

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g 9-n Reactor bu.iding isolation will automatically occur on a signal of high reactor building pressure (4 psig) or a high radiation

- signal. All fluid penetrations not required for ~ operation of the engineered safeguards equipcent will be isolated.. Remotely operated-

. isolation valves will~ have position indication inJthe control rooQ.

We have reviewed the containment isolation'systL, for conformance.

to General Design Criteria'53, 36, and 57.

We conclude that the systems meets the intent of the General Design Criteria.

6.2.1 COM3USTI3LE GAS CONTROL SYSTEMS Following a loss-of-coolant accident, hydrogen may accu =ulate inside the reactor building. The major sources of hydrogen generation include: (1) a chemical reaction between the fuel rod cladding and-the steam resulting from vaporization of the emergency core cooling wa te r, (2) corrosion of aluminum by the alkaline spray solution, and (3) radiolytic decompostion of the cooling water in the reactor core and the building sump. The. generation of sufficient hydrogen could lead to potentially ccmbustible mixtures in the containment.

Crystal River Unit 3 will utilize a containment purge system which is designed to =aintain the hydrogen concentration below its lower flammability limit. This is accomplished by introducing outside air into the containment building and allowing the displaced con-tainment atnosphere to be discharged through. the purge exhaus t filters to - the plant vent. The hydrogen purge system consists of a contain-ment atmosphere monitoring 'subsystec (hydrogen-and radioactivity),

a fresh air makeup subsystem and a discharge' subsystem.

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The applicant s tates in Amendment.17 and _27 and Appendix 143,

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-the hydrogen concentrations will' be limited to 3~.57. by volume, and

. purging of the containment will be initiated at a rate of 25.

SCFM about:ll days following an accident.

We'have reviewed the system using the' guidelines of the supplement to Regulatory Guide 1.7, " Control of Combustible Gas: Concentrations Considerations".

Our independent calculations of the -hydrogen con-centrations are essentially in agreement with those of the applicant.

We conclude that the =e thod of purging - for c)ntrol of combus tible gases is acceptable.

The reactor building cooling fan circulates the atmosphere within x-the contain=ent to provide mixing and prevent stra:ification fol-lowing a loss-of-coolant acciden t.

We conclude that this assucp -

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tion of =ixing is acceptable; however, we are continuing studies to verify :nis. conclusion on a generic basis.

We have reviewed the combus tible gas control systems for conformance to General Design Criteria.41, 42, and 43 and Safe ty Guide 1.7 We conclude that the sys tems are acceptable.

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m W. P. STEWART, DIRECTOA POWER PRODUCTION 6 o.To.*;,To,,

August 30, 1978 Mr. Robert W. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

Enclosed are forty (40) copies of " Crystal River Unit 3 Impingement Report, March 13, 1977 to March 13, 1978" for review by the Staff. This report justifies the termination of the impingement study under Appendix B Technical Specification 3.1.3 as it shows the actual impingement at Crystal River Units 1, 2 and 3 is similar to the impingement projected from the preoperational studies.

At the present time, the impingement program is bcing waived during the intake canal modification. The equipment has been mobilized to the site, the work will begin upon receipt of the Corps of Engineers Permit and it will be completed within 18 months.

So that Florida Power Corporation may be fully prepared to continue, if necessary, the study subsequent to that work, we request an expeditious review of this report.

If we can be of any assistance, please contact this office.

Sincerely, FLORIDA POWER CORPORATION

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