ML20198F300
| ML20198F300 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 01/08/1974 |
| From: | Tedesco R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| CON-WNP-0941, CON-WNP-941 NUDOCS 8605280466 | |
| Download: ML20198F300 (12) | |
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rfd $0-0 JAN 8 1974 Docket No. 50-460 V. A. Moore, Assistant Director for Light Water Reactors Group 2, L REQUEST FOR ADDITIONAL INFORMATION FOR WPPSS NO.1 Plant Name: WFPSS No. 1 Docket No. : 50-460 Licensing Stage: CP NSSS Supplier: Babcock & Wilcox Architect Engineer: United Engineers and Constructors Containment Type Dry Responsible Branch & Project Manager: LWR 2-3; T. Cox Requested Completion Date: December 28, 1973 Applicant's Response Date: March 4, 1974 Review Status: Incomplete The Containment Systems Branch has reviewed the applicable portions of the PSAR for the Washington Public Power Supply Systen No.1 ('wTPSS-1).
Enclosed is a request for additional information which specifically identifies the information we will need to complete our review.
The :nost significant review item noted is with respect to the methods used by the applicant to determine the energy release to the contain-ment during the reflood phase of the accident.
A draf t of this memorandum was provided the Project Manager on January 4, 1974.
O&ing 31:ned by:
Wrt L. Ted*"'
Robert L. Tedesco, Assistant Director for Containment Safety Directorate of Licensing
Enclosure:
As stated 8605280466 740108 PDR ADOCK 05000460 cc: w/o encl.
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REQUEST FOR ADDITIONAL INFORMATION WPPSS NO. 1 DOCTEr NO. 50-460_
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- 6. 0 Engineered Safety Features
- 6.10 State the minimum containment backpressure that has been used in the analysis of the emergency core cooling systems.
Justify this value to be conservatively low by describing appropriate conservatisms related to the assumptions of energy release to the containment, containment heat removal, containment volumes, initial l
containment conditions, modeling of the heat sinks, heat I
transfer coefficients to the heat sinks, heat sink surface area and any other parameter assumed in the analysis.
Provide the containment pressure, temperature and sump temperature response for the appropriate conservative assumptions.
- 6.11 Describe how the primary system volume which is used in calculating the initial liquid mass contained in the primary system is det ermined.
Provide the temperature assumed in the calculation of the primary system volume, and the assumed pressurizer water level.
Discuss the conservatism of these values from the standpoint of contai nment analysis.
- 6.12 Describe the calculational method used to determine the initial core stored energy.
Provide selected values of the fuel conductivity, gap conductivity, fuel burnup and fuel densificatien including an accompanying discussion of the conservatism of these values from the
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standpoint of containment analysis.
6.13 (1)
Provide an analysis of the containment pressure response for a spectrum of break sizes in the pump discharge piping by using methods and assumptions that are conservative for containment analysis. The break area should range from an area equivalent to twice the cross s.2ction area of the piping being considered to an area where the peak containment pressure has been identified for that particular type break.
l (ii) Include the following information for the cases analyzed; break area, break type, peak containment pressure, time of peak pressure, and energy released to the containment up to the time of peak pressure, t.lso curves should be provided to show the containment pressure response and containment temperature as a function of time for all breaks analyzed.
(iii) For the break producing the highest containment pressure, provide a table of mass and energy released to the con-tainment as a function of time for the entire transient; i.e.,
blowdown, refill, reflood and post reflood.
6.14 Clarify whether the 0.5 ft split break is a cold leg pump suction break or a hot leg break.
(Table 6.2-5 has it listed as a cold leg break while Table 6.2-18 states it a hot leg break.)
6.,15 Clarify the inconsistencies in Table 6.2-2, Table 6.2-6 and Table m
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6.2-8 concerning the amount of mass and energy initially in the reactor system and released to the containment at various times during the containment analysis.
- 6.16 Include in Table 6.2-2, which lists the sources of energy within the containment for the DBA, the following:
a.
The energy transferred to or from the steam generator at the start of blowdown, at the end of blowdown, at the end of the refill period, at the end of the reflood period, and at the time of peak pressure and, b.
The energy specifically contained in the reactor vessel, core internals (thin metals), hot leg piping, pump suction leg piping, pump discharge piping, steam generator tubes, steam generator inlet and outlet plenum, and reactor coolant pu=ps.
If any of the above energy sources are neglected in the containment analysis, justify with regard to the conservatism for containment pressure.
- 6.17 Provide an analysis of the containment pressure response for a postulated steam generator steam line pipe rupture in which the reactor is assumed to be at full power conditions and that consider a single active failure of a feedwater isolaton valve. Provide a 6
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e table of mass and energy release rates as a function of time.
Describe the analytical model used in the analysis.
Also discuss the conservatism in the analysis with regard to maximizing the energy release to the containment; e.g., maximum steam and feed-water valve closure times and actuation delay times.
6.18 In the analysis of the containment pressure response for Secondary System Line 3reaks, Section 6.2.1.3.11, clarify whether or not any credit was taken for heat sinks. This section states that no credit was taken for heat sinks, yet the analysis shows that for the same amount of mass and energy released to the containment over differ-ent time periods, different containcent peak pressures are obtained which indicates that heat sinks were accounted for.
- 6.19 For the hot and cold leg break sizes that result in the highest containment pressures, provide values of the heat transfer co-efficients as a function of time used in the steam generators for forward and reverse heat transfer, for all phases of the loss-of-coolant accident. Discuss the bases for these values and the O
conservatism for containment analysis.
Include in the discussion whether or not DNB was delayed or rewetting of the primary side of the steam generator tubes was accounted for.
- 6.20 (i)
The assumptions for the containment analysis concerning the release of mass and energy into the containment during the g
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. blowdown resulting from a LOCA (pg. 6.2-2, Section 6.2.1.1.b (1)) are based on B&W Topical Report BAW-10065.
BAW-10065 is related to the analysis of the emergency core cooling systems and some of the assumptions in this report may not be conservative for the containment pressure response analysis. Mass and energy release rates to the containment therefore should be maximized for determining containment pressure response analysis. Provide mass and energy release rates that are conservative for containment analyses or justify that the assumptions used in developing the mass and energy releases described in BAW-10065 are adequately conservative for containment analyses.
Consider for example, the effects of extending nucleate boiling in the core.
(ii) Clarify what method was used to develop the mass and energy release rates used in the containment pressure analysis during the core reflood period.
In Section 6.2.1.1.b (2),
pg. 6.2-2, you state the PRIT computer program was used to predict the mass and energy release to the containment after the end of blowdown, while in Section 6.2.1.3.4.c, pg. 6.2-9, you state the BAW-10065 was also used in establishing these mass and energy release rates.
If the results of these codes are different, compare and justify the mass and energy release rates.
- 6.21 Clarify the assumptions made regarding the refill phase of the LOCA. Section 6.2.1.3.6 states that reflooding starts i= mediately after the end-of-blowdown while Table 6.2-7 shows a 3-second delay from the end of blowdown to the initiation of reflooding.
Also provide a curve of core inlet fluid temperature as a function of time during the DBA and justify any corrections made to the fluid temperature and enthalpy if it is different than that predicted by CRAFT at the end of blowdown.
- 6.22 The following questions pertain to the reflood and post-reflood phase of the loss-of-coolant accident.
Describe the analytical model used to predict the mass a.
and energy released to the containment during reflood and post-reflood phase of the loss-of-coolant accident. Discuss the conservatism in the model with respect to maximizing the energy release to the containment.
Include the assumptions made regarding all energy sources, the flow resistance in the broken and intact loop, and the specific volume used in each flow element.
b.
Discuss the assumptions made in the reflooding calculation regarding steam condensation by the emergency injection water.
If condensation is assumed, provide justification based on applicable experimental data; i.e., data corresponding
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t to the conditions in the primary system. Provide a sen-sitivity analysis showing the mass and energy released and the effect on containment pressure if:
(1) no condensation and partial ECCS operation is assumed, (2) no condensation and full ECCS operation is assumed, (3) condensation and partial ECCS operation is assumed, and (4) condensation and full ECCS operation is assumed.
c.
Discuss the assumptions made regarding separation of entrained liquid leaving the core during the reflooding period. We believe a conservative approach would be to assume no liquid separation so that all liquid leaving the core would enter the steam generator and be available for heat.
d.
Discuss the assumptions made in calculating the carryout fraction from the core (ratio of core exit flow to core inlet flow) during reflood.
These assumptions should be justified by comparison with the results of the FLECHT experiments for average core conditions during reflood.
We believe a carryout rate fraction of approximately 0.8 will occur until the 10-foot level in the core is covered.
- 6.23 In the post-reflood phase of the LOCA for a pump suction break, when the core has been recovered with water, a two-phase mixture of steam and water will be generated. Provide an analysis showing height that the two phase mixture will rise above the core.
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any water is calculated to enter the steam generator, provide the energy release rate to the containment as a function of time.
6.24 The following questions pertain to the subcompartment pressure analyses:
a.
State and justify what critical flow model was used in the subcompartment analysis.
b.
Provide the blowdown mass rate (1b/sec) and energy (Btu /lb) up to the time of peak pressure for each subcompartment analyzed.
(Approximately 25 values should be provided) c.
Provide and justify with regard to conservatism the orifice type equation used to calculate the flow between adjacent compartments.
Include in the justification the effect of choking if it occurs.
d.
Since the area of the surge line is about 1 f t, justify the assumption of a 0.72 ft break in the analysis of the pressurizer compartment pressure response.
Justify the size of the subcompartment nodes (volumes) e.
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used in the analysis. Discuss the consideration given to the effects of possible flow blockage by components within a compartment.
f.
The reactor cavity is analyzed as a single node. Provide i
a multi-node analysis assuming the pipe penetration, vessel i
l to shield wall annulus, and upper and lower cavities as f
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separate nodes with appropriate modeling of the vent areas.
Provide the design pressures for each of these subcompartments.
6.25 Describe the method of analysis and provide the results of analyses of the jet forces which cad impinge on the containment and internal structures. Discuss the structural design capability of each to withstand the differential pressure and jet forces resulting from the LOCA.
6.26 Describe the initial plant conditions, assumptions and analysis (including the time) that would cause the containment pressure to be decreased to.a negative pressure of 4 psi.
(Section 6.2.1.lf.)
- 6.27 Provide the value of the pressure setpoint that was assumed in the containment analysis to actuate the containment spray system.
Justify the adequacy of the delay time assumed in the analysis.
- 6.28 Discuss the method and accuracy of the method used to calculate the containment free volume. Provide a sensitivity study of the effects of the uncertainty in calculating the containment free volume on the containment pressure response under DBA conditions.
Describe any tests that could be performed to verify the free containment volume.
6.29 In each of the containment pressure transients, described in Section 6.2.1.3.4.a, the engineered safety features are assumed to operate at the time indicated in Tabic 6.2-7.
Tabic 6.2-7 is for t
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- t a double-ended hot leg split break. Justify the adequacy or conservatism of these times for breaks other than a double-ended hot leg split.
6.30 Justify the use of the following containment initial conditions used in the containment analysis:
a.
relative humidity of 75%..
b.
initial heat transfer coefficient on inside wall of 5.8 o
Btu /hr-ft' *F.
6.31 Provide justification for the lack of an interface resistance (contact or gap resistance) between steel and concrete for the steel-lined hemispherical dome heat sink (Table 6.2-20, Item 2) used in the containment pressure calculation.
6.32 Justify the amount of exposed surface area available for the concrete ficar heat sink (Table 6.2-20).
The value indicates that credit was taken for the area occupied by the reactor vessel, ring wall and other structures.
- 6.33 With respect to the modeling of the heat sinks in the CONTRAST-S computer program, provide and justify the mesh spacing used for the concrete, steel, and steel-lined concrete heat sinks.
6.34 Clarify what equation was used for h sinc Se ti ns 6.2.1.3.4.d atas and 15.B.1 have different equations.
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f 6.35 Describe the procedure for transferring the spray system pump suction from the Borated Water Storage Tank to the reactor building sump.
Include in the discussion the information that
- 6.40 will be available to the operator to guide him in making a timely decision.
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j 6.37 Provide the sizes of the coarse and fine screening in the sump j
to prevent debris entering the pump suction and of the smallest i
spray header nozzle orifice. Also provide a description of the I
screen cap which is provided over the pump suction opening.
Include the type of material and the size of the openings.
I 6.38 With regard to the containment isolation system, identify those containment isolation valve arrangements that deviate from AEC t,
General Design Criteria SS. 56 and 57 dated July 7, 1971, and i
i indicate where the valving arrangements do not fully conform with i
these criteria.
Discuss the bases for concluding that the design meets the intent of the CDC.
- 6.39 Provide the design pressure of the guard pipe for the piping from
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r the containment sump to the suction of the decay heat pump and
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