ML20210T131

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Forwards Containment Sys Branch Request for Addl Info After Review of Applicable Portions of Psar.Significant Concerns Listed
ML20210T131
Person / Time
Site: Satsop
Issue date: 01/23/1975
From: Tedesco R
Office of Nuclear Reactor Regulation
To: Deyoung R
Office of Nuclear Reactor Regulation
References
CON-WNP-1742 NUDOCS 8605290751
Download: ML20210T131 (15)


Text

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  • Docht Nos. 50-508/509 l

. R. C. DeYoung, Assistant D11 rector for Light Water Reactors. Grcup 1., RL /_- .n. ,;a l . REQUEST POR ADDITIONAL INPORMATION 701 WASHINGTON FUBLIC POWER SUPPLY e,8YSTEMsWP-345 3. p. 9tW Mx tiQ. M2 A.-;;.: om o ~.R& i -l4a ? & f. W M... & V.l.'l~0.l;k W $ fQ & i'O.'.9', b W G V' ::':,. r & %.h Q ,l ~ l. .s M P1mt Names 7Weshlagteur hablie Power Supply;$ystems, WNP-345 ' L.q.MMtb 9 i .t c % 1%cket Wee.31 50.505/50it.. cj. f r.OM': ai*~*N *:. 5*5,2y' % 6 N N. y~*". 4- , h. Y '., y,. .' VM

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7. O'Reilly Requested Conplation Datet January 24. 1975 Applicant's Response Dates !!srch 28,1975 4-Raview Statu.s: Awaiting Informagion

'i The enclosed request for additional inferriat. ion (Q-2) fev the subject i s .plaats has been prepared by. the contatsmant Systems Branch of ter reviewing the appliembla portf ens 'of t;he TfP-3 PSAR.' ' -+ g m, a e. ; q a., 5,., s v' '+ .c yc y4 w.4, ^ c,1 4

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Our signifier.nt concaens are related to the followings g , 1. The applicant has"not provided an analysis of the functional . capahtlity of the cantainment vacuum relief systers. 2. The applicant has act proposed to conform to all the requiro- . monts of Appedix J to 10 CFR Par; 100 - Containment thak Testing. N 0, ~ l .3. The applicant has not provided sufficient infounation t5 justify I _ that purging of tt.a containment during nonaal plant opa' ration i s. l Lis so infraquant that t!ve pura valves are not required to be 'j/.h y l designed for dynamic loadin3s MFelcped during a IOCA. q: j 9y. p 77: w-;,, i. t 4. ' The,eyplicant has met p;;ov.idad. a itinimous centairnsent pressure r V.'. analysia as t'equirst by Section 50 34'of 30 CT1 Part 5th, i [ $ < [ y,,$ h. @ W' m..' :y, c-9 ; G,[h[0rigf aal Sf[rneS7 Y 14 .a ~* 'y ' ;.i. Ls: EcbertL %wco i i l Robert T.. Todasco, Aasistant Diracter t for Contaisuant Safety Division of Technical Enview i Enclostr.re: . 0ffice cf Nuclear Reacter P.egulat. ion '/' As stated ,_ f h&W

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^ G, 042-1 y 04,2.0 CONTAINMENT SYST22fS BRANCH 042.1 Describe and justify the method to be used during isolation (6.2.1) valve leakage testing to convert water leakage rates to air leakage races. 042.2 Section 6.2.1.4.4.2, Type B Local Leak Tests, is not consis-(6.2.1) tent with the requirements of Appendix J to 10 CFR Part 50. Consistent with the requirements of Appendix J, we will require that the personnel airlocks be leak tested after each opening and at six-month intervals. 042.3 Section 6.2.1.4.4.3, Type C Isolation valve Leak Tests, is (6.2.1) not consistent with the requirements of Appendix J to 10 .CFR Part 50. Consistent with these requirements, we will require that isolation valves be leak tested at intervals not exceeding two years. 042.4 Provide an analysis of the containment negative pressure (6.2.1.1) response assuming inadvertent actuation of the containment spray system during normal power plant operation. Assume ~ ? containment initial conditions of 0% and 100% relative . humid 4.ty, and the vacuum relief system is operable and not operable. Describe in detail the analytical model used, in-cluding assumptions. Also, provide the following information: (1) Graphically show the containment pressure response as a function of time for each of che above cases. (2) Graphically show the flow rate as a function of s differential pressure across a vacuum relief valve. (3) Describe the periodic test program which will demon-strate operability of the vacuum relief system. 042.5 It is stated on page 6.2-176 that the purge valves will not (6.2.4) be designed for the dynamic loads associated with a pipe break accident, since the purge system is not expected to be used on a routine basis and would rarely be in operation during normal plant operation. Discuss why purging is necessary and under what conditions. When during normal plant operation, would purging of the containment be required? Indicate the frequency and duration of purge operations. Estimate the fraction of time during a plant operating cycle that the purge system would be operated. 042.6 Provide ar. analysis of the amount of containment atmosphe e (6.2.5) that would be released to the environment assuming the con-t.61nment purge system is operating at the time of a loss-of-ccolant accident. The analysis should be done for a spectrum

F_ g ~)j 042 - 042.6 of reactor coolant system pipe break sizes. Identify the ' (6.2.5) instrumentation and setpoints that initiate purge system isolation and justify that the setpoints vill be reached. + Also, provide an analysis of the effect cne loss-of-contain-ment atmosphere would have on the minimum containment pressure used in the evaluation of emergency 6ere cooling system per-l formance. 042.7 Provide an analysis of the pressu[e response of the ECCS area i (6.2.1) outside containment following a loss-of-coolant accident. Discuss and justify the assumptions made in calculating heat transfer to the ECCS area. Graphically show the ECCS area pressure response as a function of time following a LOCA. I 042.8 Provide an inventory of aluminum in the containment. Specify (6.2.5) the surface area, thickness, and mass of each item. 042.9 Graphically show the containment vessel temperature responses (6.2.1) as a function of time for the loss-of-coolant accident, or main steam or feedwater line break accident whichever results in the highest containment atmosphere temperatures. Discuss the method of analysis and provide justification that the assumptions used in the analysis are conservative. 042.10 Revise Figure 6'.2.2-1 to show the arrangement of the recir-(6.2.2) culatica line intikes; i.e., containment' sump used for ECCS and containment >3 pray. t 04'2.11 Figure 1.2-6 indicates that the containment sump is at a (6.2.2) higher elevation than the safety injection system racirculation sump. Provide justification for this arrangement in light of the discussion in Regulatory Guide 1.82 which suggests that sumps for routine building drainage should be at a slightly, lower elevation than sumps intended to serve engineered safeguard systems pumps. 042.12 Section 50.34 of 10 CFR Part 50 requires that an analysis and (6.2.1) evaluation of ECCS cooling performance following postulated loss-of-coolant accidents be performed La accordance with the requirements of Section 50.46. Appendix K, "ECCS Evaluation Models," to 10 CFR Part 50 sets forth certain required and acceptable features of evaluation models. Appendix K states, in part, that the containment precaure used for evaluating cooling effectiveness during reflood and spray cooling shall { not exceed a pressure calculated con'servatively for this purpose. It further requires that the calculation include the effects of operation of all hastalled pressure _ reducing systems and processes. Branch Technien1 Position CSB ' 6-1, " Minimum i F v4. L a e-

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042-3 2 042.12 Containment Pressure Model for PWR ECCS Performance Evaluatiott," (6.2.1) which is attach 6d, provides additional guidance for the perfer-mance of a minimum containment pressure analysis and should be used when the analysis is performed. Therefore, state the l minimum. containment pressure that has been used in the analysis of the emergency core cooling system. Justify this value to be j conservatively low by describing the conservatism in the ~! , assumptions of initial containment conditions, modeling of the containment heat sinks, heat transfer coefficients to the heat sinks, heat sink surface area and any other parameter assumed in the analysis. Graphically show the containment pressure and temperature responses and sump temperature response, as functions of time for the most conservative assumptions. + i a O f \\ f i 4 e

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~. Q ) 7 s ( 3 Branch Technical Position CSB 6-1 J MINIMUM CONTAINMENT PRESSURE MODEL FOR PWR ECCS PERFORMANCE EVALUATION A. BAGGROUND Paragraph I.D.2 of Appendix K to 10 CFR Part 50 (Ref.1) requires that I the contain= ant pressure used to evaluate the performance capability of a pressurized water reactor emergency core cooling system does not exceed a pressure calculated conservatively for that purpose. It further requires that the calculation include the effects of operation of all installed pressure-reducing systems and processes. Therefore, the followhg Branch Technical Position has been developed to provide guidance in the performance of a minimum containment pressure analysis. B. BRANCH TEGNICAL POSITION 1. Input Tnformation for Model a. Initial Containment Internal Conditions, The minimum containment gas temperatura, minimum containment pressure, and maximum htsnidity that may be encountered under limiting normal operating conditions should be used. b. Initial Outside Containment Ambient Conditions A reasonably low ambient temperature external to the containment l should be used. c. Containment Volume The maxistan net free containment volume should be used. This maximum frse volume should be determined from the gross contain-ment volume minus the volumes of internal structures such as walls and floors, structural steel, major equipment, and piping. The individual volume calculations should reflect the uncertainty in the component volumes. 2. Active Heat Sinks a. Spray and Fan Cooling Systems l The operation of all engineered safety feature containment heat removal systems operating at maximum heat removal capacity; ( 3 -ee e m,,, ~

'3 ) I i.e., with all containment spray trains operating at maximum flow conditions and all emergency fan cooler units operating, should be assumed. In addition, the minimum tu perature of the stored water for the spray cooling system and the cooling water supplied to the fan coolers, based on technical specification limits, should be assumed. t i Deviations from the foregoing will be accepted if it can be shown that the worst conditions regarding a single active i failure, stored water temperature, and cooling water temperature have been selected from the standpoint of the overall ECCS model. 1 b. Containment Steam Mixing With Spilled ECCS Water The spillage of subcooled ECCS water into the containment' provides an additional heat sink as the subcooled ECCS water mixes with the steam in the containment. The effect of the ste.am-water mixing should be considered in the containment pressure calculations. c. Containment Steam Mixing With Water From Ice Melt The water resulting from ice melt in an ice condenser containment provides an additional heat sink as the subcooled water mixes with the steam while draining from the ice condenser into the lower containment volume. The effect of the steam-water mixing should be considered in the containment pressure calculations. 2. Passive Heat Sinks a. Identification The passive heat sinks that should be included in the containment evaluation model should be established by identifying those l structures and components within the containment that could influence I the pressure response. The kinds of structures and components that I should be included are listed in Table 1. t l 1

f3 3 ~ - An envelope of passive heat sink data has been developed, based on information obtained from safety analysis reports, that would be acceptable for use in performing minimum containment pressure analyses until such time (i.e., at the OL review) that a complete identification of available heat sinks can be made. This simplified approach has been followed on operating plants by licensees en-gaged in performing minimum containment pressure analyses to comply with Section 50.46 of 10 CFR Part 50. For such cases, and for CP reviews where a detailed listing of heat sinks within the con-tainment cannot be provided, the following procedure may be used to model the passive heat sinks within the containment: (1) Use the surf ace area and thickness of the primary containment steel shell or steel liner and associated anchors and concrete, as appropriate. (2) Estimate the exposed surface area of other steel heat sinks in accordance with Figure 1 and assume an average thickness of 3/8 inch. (3) Model the internal concrete structures as a slab with a thickness of 1 foot and an exposed surface area of 160,000 ft. The heat sink thermophysical properties that would be acceptable are shown in Table 2. At the OL stage, applicants must justify the claims they have made for heat sinks. b. Heat Transfer Coefficients The following conservative condensing heat transfer coefficients for heat transfer to the exposed passive heat sinks during the blowdown and post-blowdown phases of the loss-of-coolant accident l should be used (See Figure 2): (1) During the blowdown phase, assume a linear increase in the condensing heat transfer coef ficient from h,f tg,1=8 Btu /h M t - 4, g at t=0, to a peak value four times greater than the maximum calculated condensing heat transfer coefficient at the end of blowdown, using the Tagami correlation (Ref. 2), l l t ~ ~ ~ ~ ~

,~) m 0 0.62 h = 72.5Lv'_ where h,,= maximum heat transfer coefficient, Btu /hr-ft

  • F Q = primary coolant energy, Btu V = net free containment volume, ft t = time interval to end of blowdown, sec.

P (2) During the long-term stagnation phase of the accident, characterized by low turbulence in the containment atmosphere, assume condensing heat transfer coefficients 1.2 times greater than that predicted by the Uchida data (Ref. 3) given in Table 3. (3) During the transition phase of the accident between the end of blowdown and the long-term post-blowdown phase, a reasonably conservative exponential transition in the condensing heat transfer coefficient should be assumed (See Figure 2). The calculated condensing heat transfer coefficients based on the above method should be applied to all exposed passive heat sinks, both metal and concrete, and for both painted and unpainted l surfaces. Heat transfer between adjoining materials in passive heat sinks should be based on the assumption of no resistance to heat flow at the material interfaces. An example of this is the containment liner to concrete interface. C. REFERENCES 1. 10 CFR 50.46, " Acceptance Criteria For Emergency Core Cooling Systems For Light Water Nuclear Power Reactors", and 10 CFR Part 50, Appendix K, "ECCS Evaluation Models". 2. T. Tagami, " Interim Report On Safety Assessments and Facilities Establishment Project In Japan For Period Ending June 1965 (No.1)", prepared for the National Reactor Testing Station, February 28, 1966 (unpublished work). I 1 1

. 3 ) l 3. H. Uchida, A. Oyama, and Y. Toga, " Evaluation of Post-Incident Cooling Systems of Light-Water Power Reactors", Proc. Third Inter-national Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, United Nations, Geneva (1964). .i e 8 a .mm,a y w----

q ) L TABLE 1 IDENTIFICATION OF CONTAINMENT HEAT SINKS 1. Containment Building (e.g., liner plate and uternal concrete walls, floor and sump, and liner anchors) 2. Contain= mat Internal Structures (e.g., internal separation walls and floors, refueling pool and fuel transfer pit walls, and shielding walls) 3. Supports (e.g., reactor vessel, steam generator, pumps, tanks, major components, pipe supports, and storage racks) 4. Uninsulated Systems and Components (e.g., cold water systems, heating ventilation and air conditioning systems, pumps, motors, fan coolers, recombiners, and tanka) 5. Miscellaneous Equipment (e.g., ladders, gratings, electrical cable trays and crar.es) ~ e 9 ~ m .s,

y TABLE 2 HEAT SINK THERM 0 PHYSICAL PROPERTIES i Specific Thermal Density Heat Conductivity Material lb/ft3 Btu /lb 'F Btu /hr-ft 'F l Concrete 145 0.156 0.92 Steel 490 0.12 27.0 I e l w er-w., -,.w- .y-w.., - m -

m TABLE 3 UGIDA HEAT TRANSFER (X)EFFICIENTS Mass Heat Transfer Mass Heat Transfer Ratio Coefficient Ratio Coefficient -l (1b air /lb steam) (Btu /hr-f t _*F) (1b air /lb steam) (Btu /hr-f t _op) 2 2 i 50 2 3 29 i 20 8 2.3 37 { 18 9 1.8 46 1 14 10 1.3 63 10 14 0.8 98 7 17 0.5 140 5 21 0.1 280 4 24 i f e 5 4 w

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