ML19345H314

Second Round Questions Re Assumptions & Results of Containment LOCA Analyses

ML19345H314
Person / Time
Site:	Waterford
Issue date:	05/13/1981
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML19345H308	List: ML19345H307 ML19345H310 ML19345H312 ML19345H313 ML19345H314
References
NUDOCS 8105200096
Download: ML19345H314 (16)

Text

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REQUEST FOR ADDITIONAL INFORMATION (Q-2)

O roa wattaroao-3 022.24 (6.2.1.1)

Concerning the assumptions and results of the containment loss-of-coolant accident (LOCA) analyses, verify that the assumption of loss-of-offsite power together with the most severe single active failure was used in the cantainment LOCA analyses.

022.25 (6.2.1.1)

Since the peak calculated containment pressure from the main steam line break (MSLB) analysis is very close to the containment design pressure, provide justification for concluding that the limiting MSLB accident condition has been identified.

In this regard, provide the results of a sensitivity analysis showing the behavior of peak pres-sure and temperature for power levels between 0 and 20%, 20% and 50%,

50% and 75%, and 75% and 102%, and for several different break areas to confinn that the peak pressure and temperature cases for the main steam line break analysis have been selected.

022.26 l

(6.2.1.1)

Provide the following additional infonnation ctacerning the instru-l mentation to monitor the containment atmosphere pressure

• and tempera-ture a,d the sump water temperature following an accident:

a. Discuss and justify the accuracy and response time of the instru-mentation.

b. Discuss your plans for increasing the temperature ranae of the con-1 tainment fan cooler inlet temperature instrumentation (4-270 F, (Note:

See also NUREG-0737, Clarification of TMI Action Plan Re-quirements, II.F.1, Attachment 4, Containment Pressure Monitor)

ENCLOSURE

4

. Table 7.5-1), which is used to monitor containment temperature fol-lowing an accident, to a range encompassing the worst case transient (102 percent power main steam line break with one cooling train fiil-ure,413.5'F).

.s.

022.27 (6.2.1.1)

Provide justification for using the mass and energy release of the 2

9.82 ft DESLS with maximu-safety injection, for the analysis of the secondary containment annulus pressure response to a LOCA.

Compare this response to that for the same break with minimum safety injection available.

022.28 (6.2.1.1)

State the initial containment internal pressure used in the containment external pressure analysis since an inconsistency exists between Table 6.2-11 and Figure 6.2-13, Sheet 1 (14.7 psia) and the response to Question 022.11 (14.25 psia).

Provide any necessary revisions to the FSAR containment vessel and shield building external pressure analysis.

022.29 (6.2.1.1)

There is an apparent conflict between Table 6.2-4 and the text of Sec-tion 6.2 regarding the DBA LOCA. The text states that the 9.82 ft j

DESLS break with maximum safety injection is the worst case, whereas i

the table shows that the same break but with minimum safety injection is slightly worse in terms of both pressure and temperature as shown below:

Pesk Pressure Peak Temperature (psia)

( F)

Max SI 43.1 268.7 Min SI 43.2 269.3 l

l

t

. Confinn which LOCA event is the worst case (DBA), and correct the FSAR as required.

022.30 (6.2.1.2)

Provide the basis for the analysis of vent flow behavior in the sub-compartment models. Show how the models comply with the guidelines of SRP Section 6.2.1.2, Item II.4.

022.31 (6.2.1.2)

The response to Question 022.2 Part (d), describes the simi.larity of the Waterford-3 reactor cavity to the reactor cavity of the Shearon Harris plant and indicates the adoption of the Shearon Harris nadali-zation sensitivity study results for Waterford-3.

It is not clear, however, in the discussion on subcompartment modeling for the steam generator and pressurizer compartments whether these analyses similarly used the Shearon Harris nodalization sensitivity study.

If so, pro-vide the following:

a.

Compare the dimensions of the steam generator and pressurizer compartments for the two plants.

b.

Provide the results of the Shearon Harris study to show the effect of varying the nodal arrangement or number of nodes near the break volume.

If the steam generator and pressurizer compartment nodalization sensi-tivity analyses are not based on Shearon Harris, provide the results of the nodal sensitivity analysis for these compartments, t

. 022.32 (6.2.1.2)

Provide the forward and reverse loss coefficients for the 'ubcompart-ment analyses of the reactor cavity, steam generator, and pressurizer compartments. At present Tables 6.2-14, -15 and -16 only provide s single value for each junction.

022.33 (6.2.1.2)

Appendix 5.4A, " Dynamic Analysis for Waterford-3 Reactor Vessel Support Locds Under LOCA Conditions," is inadequate in satisfying NRC needs for force and moment calculations as outlined in NUREG-0609, "Assymetric Blowdown Loads on PWR Primary Systems." The Appendix deficiencies and requests for additional information are as follows:

a.

The results shown in the Appendix do not correspond to nodalization assumptions used in the reactor cavity subcompartment analysis (i.e., do not provide the flow entering the lower reactor cavity).

Therefore, orovide a description of the model used to compute forces in the horizontal and vertical directions on the outside of the reactor vessel and provide results of the analysis. The model discussion should include a group of figures showinc. direction of l

reference axes (X, Y and Z), areas used in determining forces, and comparison to subcompartment nodalization used in pressure calcula-l tions.

l b.

There was no :alculation of moments included in Appendix 5.4A.

Pro-vide moment calculations about the X, Y, and Z axes and show the reference axes used.

c.

Provide justification for assuming that the entire break flow could be dumped into a single compartment volume in the reactor cavity l

s' wa

-+,

+ - - -

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-w subcompartment analysis.

Specifically show how this method will satisfy the requirements for spacial pressure variation as dis-cussed in SRP 6.2.1.2.

022.34 (6.2.1.2)

The results of the pressurizeq scbcompartment analysis show that a large differential pressure results between Volumes 1 and 2.

Provide the design value for the steel boundary between these volumes which make up the support area and show that the calculated values do noi exceed the design value.

Also, provide peak differential pressures for the surge line break bed tween Volumes 2 and 8,1 and 4, 3 and 5, and 3 and 8, or provide scme justification to show that these peak pressures will be no greTter than those given in the present FSAR or will not exceed the stated design pressure of 10 psid.

022.35 (6.2.1.3)

Verify that calculations of the energy available for release during a loss-of-coolant accident were done in accordance with the requirements of 10 CFR Part 50, Appendix K, paragraph 1. A.

Also verify that gon-siderations of fuel clad swelling and rupture, prescribed by paragraph 1.8 in Appendix K to 10 CFR Part 50, have NOT been included in the evalu-ation model, in order to maximize the energy available for release from the core in the evaluation of the functional capability of the cuitain-ment structure.

022.36 (6.2.1.3)

It is not clear that FSAR Table 6.2-1 contains the complete spectrum of possible pipe breaks.

Define " double-ended slot" and explain how this l

(

L-

=..

l l correlates with the terms " double-ended pipe break" and " longitudinal split." Verify that the spectrum of possible pipe breaks analyzed for mass and energy release for postulated loss-of-coolant accidents in-cludes a longitudinal split in,,the largest pipe with break area equal to the cross-sectional area of the pipe, or justify its exclusion (Reference SRP Section 6.2.1.3).

022.37 (6.2.1.4)

Verify that stored energy in the metal of the feedwater and steam lines connected to the affected steam generator was considered as a source of energy in the mass and energy release analysis for postulated secondary system pipe ruptures.

022.38 (6.2.1.4)

Concerning the secondary system pipe break analysis, verify that mass release rates during postulated secondary system pipe ruptures were calculated using the Moody model or a model that is demonstrated to be equally conservative. What flow multiplier was used (if any) at the break?

022.39 (6.2.1.4)

Provide the following information regarding the main steam line break analysis:

(1) Discuss how the feedwater flow into the steam generator, during the period from the initiation of the MSLB event until the feedwater isolation valve is completely closed, has been included in the blow-down inventory.

(2) Since a single failure of the Main Feedwater Isolation Vai'.'e is backed up by tha main feedwater regulating valve, specify the valve closure time for the regulatir.g valvo.

. 022.40 (6.2.1.5)

Justify as conservative the initial primary containment atmosphere temperature of 80*F used in the minimum containment pressure analysis for ECCS perfomance capability.

022.41 (6.2.1.5)

Discuss why the 0.6 x DES /PD break is the limiting break in evaluating the minimum containment pressure analysis for ECCS perfomance capabil-ity studies.

022.42-(e.2.2)

FSAR Figures 6.2-41, 42, 43 and 44 show a considerable amount of over-lapping of the spray patterns of the containment spray system. Specify the extent of overlapping of the sprays and provide an analysis of the effect of this degree of overlapping on the heat removal effectiveness of the sprays.

Provide the above infomation and analysis for both full and partial spray system operation.

022.43 (6.2.2)

Explain the basis for the assumed volume from each of the coatributors to the minimum calculated safety injection system sump water depth, used to detemine the elevation head in the containment spray system pump NPSH calculations.

l 022.44 (6.2.2)

Identify and f.:stify the conservatisms applied in determining the water level in the containment for the containment spray system pump NPSH cal-culations, such as the uncertainty in determining the free volume in the lower part of the containment that may be occupied by water, and the quantity of water that may be trapped by the reactor vessel cavity and the refueling cavity.

022.45 (6. !'. 2)

The grating installed at elevation -4.0 ft is considered an acceptable substitute for the outer trash rack reconinended by Regulatory Guide 1.82.

However, the major opening-for the stairs proximate to the safety injec=

tion system pump is cause for. concern.

Provide assurance that the open-ing for the stairs will not allow the passage of large debris into the sump area.

022.46 (6.2.3)

Propose a Technical Specification to maintain the Shield Building annu-lus pressure at or below the negative pressure assumed as the initial annulus pressure in the Shield Building annulus pressure analysis, as a limiting condition for cperation.

022.47 (6.2.3)

The Shield Building annulus pressure analysis presented in the FSAR has several non-conservative assumptions. Therefore, provide a new analysis of the Shield Building annulus pressure for the worst case loss-of-coolant accident assuming:

t a.

Adiabatic boundary conditions at tne surfaces of the Shield Building exposed to the outside environment.

l l

b.

Inteakage rates from the containment and the outside environs. con-stant at their design values (see previous Question 022.12.(1)).

c.

Loss-of-offsite power and the most severe sinale active failure; identify and justify the most severe single active failure assumed.

Additional guidance on performing the analysis of the pressure and tem-perature response of the Shield Building annulus to a loss-of-coolant ac-cident is provided in SRP Section 6.2.3.

l l

)

l 022.48 (6.2.3)

Provide additional infomation to justify the conservatism of the 30-second delay used in the Shield Building annulus pressure analysis for the Shield Building Ventilation System fan (s) to reach rated speed after a. loss-of-coolant accident coincident with loss-of-off, site power.

022.49 (6.2.3)

Describe the pre-operational and periodic tests (i.e., Technical Speci-fication Surveillance Requirements) that will be conducted to determine the ability of each Controlled Ventilation Area System (CVAS) to draw down the Reactor Auxiliary Building areas exhausted by the CVAS to a prescribed negative pressure in a prescribed time. Also, describe the pre-operational and periodic tests of the Shield Building Ventilation System (SBVS) and the Controlled Ventilation Area System (CVAS) to de-temine inleakage rates and uniformity of negative pressure in the Shield Building annulus and in Reactor Auxiliary Building areas ex-hausted by the CVAS.

1 1

022.50 -

(6.2.3)

Provide additional infomation on the periodic tests similar to those for the Shield Building (i.e., Technical Specifications similar to 3.6.8.2 and 3.6.8.3) to ensure the leakage integrity of the Reactor Aux-i iliary Building areas exhausted by the Controlled Ventilation Area Syttem.

022.51 l

(6.2.3)

The response to Question 022.4 is not sufficient to allow adequate review of bypass leakage paths.

Identify and list by type (i.e., isolation valves, seals and gaskets, welding joints, etc.) all leakage paths which

(

. could bypass the annulus ulume treated by the Shield Building Ventila-tion System. Guidance in determining bypass leakage paths in dual con-tainment plants is provided in Branch Technical Position CSB 6-3.

In-dicate where all bypass leakage paths terminate and if that area is part of the Reactor Auxiliary Building that is exhausted by the Controlled Ventilation Area System.

022.5?

(6.2.3)

Provide justification that if the Shicid Building Ventilation System can draw down the annulus to a negative pressure greater than or-equal to 0.25 in, water gauge in one minute following a Safety Injection Actu-ation test signal (Technical Specilication 4.6.8.1.d.4) that the Shield Building annulus pressure would be maintained below -0.25 in, water gauge following the worst case loss-of-coolant accident.

022.53 (6.2.3)

Identify all openings, such as personnel doors and equipment hatches, ':nto the Shield Building annulus and the Reactor Auxiliary Building areas ex=

hausted by the Controlled Ventilation Area System.

Confim that ill open-ings are under administrative control and are provided with position in-dicators and alarms having read-out in the main control room.

Evaluate the effect of an open door or hatch on the functional capability of the Shield Building Ventilation System or the Controlled Ventilation Are; System and provide information on plans for confimatory preoperational tes ts.

022.54 (6.2.3)

Provide a description of the indications and alarms in the main control room which provide the operator information on the satisfactory alignment of valves and operation of the Controlled Ventilation Area System and the maintenance of a negative pressure of greater than or equal to 0.25 in.

water in the controlled ventilation areas.

022.55 (6.2.3)

Verify that following an accident initating a Safety Injectior. Actuation Signal the environmental conditicm (i.e., temperature, humidity) in the

~

Reactor Auxiliary Building areas served by the Controlled Ventilation Area System will not exceed the environmental qualification conditions of the contained engineered safety feature equipment.

022.56 (6.2.3)

Provide additional information on the Shield Building Ventilation System (SBVS) characteristics (i.e., pressure drop versus flow) for the different operating modes and varying Shield Building annulus pressure to be used in conjunction with Figure 6.2-51 (SBVS Fan Perfomance Curves) to perform confirmatory analysis of the Shield Building annulus pressure response following a LOCA.

022.57 (6.2.4)

Information. provided in response to Question;022.9 regarding thetcontain-ment leak testing program is deficient in the following respects:

a.

No justification is given for Penetrations 53 and 54 (Instrument H&V) ren.aining fluid-filled (i.e., not being vented and drained) during Type A tests.

Provide justification.

l b.

The justification given in Table 6.2-43 for not including Penetrations i

53 and 54 in Type C leak tests is inadequate. Show that containment isolation valves associated with these penetrations do not constitute potential containment atmosphere leak paths following a loss-of-coolant accident.

. c.

FSAR Section 6.2.6.3 does not provide evidence to show accept-ability of testing the valves listed in Table 6.2-44 with pressure applied in the reverse direction.

Provide evidence in the form of test results or design descriptions of the applicable valves.

022.58 (6.2.4)

All power-operated containment isolation valves should have position in-dication in the main control room.

The information provided in FSAR Tables 6.2-32 and 7.5-1 and in FSAR Section 7.5 does not show that this requirement has been met.

Demonstrate that all power-operated contain-ment isolation valves listed in Table 6.2-32 have position indication in the main control room, by listing each valve tag number along with the number of the main control room panel on which in' ication is available.

d 022.59 (6.2.4)

Table 6.2-32 indicates that in the event of loss of power to the valve operator of valves 2MS-V602A and 2MS-V604B (main stsam); 2MS-VG70, 2MS-V671, 2MS-V663, and 2MS-V664 (main steam sample and drain); and 2FW-V823A and 2FW-V8248 (main feedwater), the valves fail as is.

Pro-vide the necessary justification that this is the " safe" position as opposed to failing closed.

022.60 l

(6.2.4)

Manual valves which serve as an isolation barrier should be under the administrative control required for " sealed closed barriers," as de-i fined in SRP Section 6.2.4, Item II.3.f.

This also applies to manual valves in test, vent, and drain lines, although leak testable blind flanges may be used in place of sealed closed isolation valves.

Pro-l vide assurance that the above guidance will be followed for all manual valves employed as containment isolation barriers.

l

. 022.61 (6.2.4)

Table 6.2-32 for penetrations #1 and #2 does not list the safety relief valves or, valves 2MS-V697 and 2MS-V698 which, according tc Figure 10.2-4 Sheet 1 of 2, serve as containment isolation barriers.

Explain or correct the absence of these valves from Table 6.2-32.

022.62 (6.2.4)

There are several inconsistencies between the actuation signals listed for valves in Table 6.2-32, the Section 7.3 tables listing components actuated by the various signals of the ESFAS, and the FSAR text (Section 6.2.4 and elsewhere). Provide the needed FSAR corrections to remove these inconsistencies.

022.63 (6.2.4)

Provide a fluid system piping drawing showing the main steam line valves 2MS-V670, -V671, -V663, and -V669 (not shown in Figurt' 10.2-4 Sheet 1 of 2 referenced in Table 6.2-32a).

1 022.64 (6.2.4)

Concerning containment isolation of the chemical and volume control charg-ing line (penetration #27):

A.

Provide the justification for locking open the outside isolation valve (2CH-E1529A/B).

Describe how the valve will be locked open and how quickly it can be isolated if leakage is detected from this line outside containment.

B.

Specifically describe the provisions for detecting possible leakage from this line outside containment.

C.

Provide the justification that failing open is the " safe" position for the outside isolation valve (2CH-E1529A/B).

. 022.65 (6.2.4)

Table 6.2-32 states that the component cooling water inlet valve (2CC-F146A/B) and outlet valves (2CC-F147A/B and 2CC-F243A/B) for the reactor coolant pumps and CEDM cooler fail open.

Provide the necessary justification that this is the " safe" position as opposed to failing closed.

022.66 (6.2.4)

For the safety injection lines from the SIS sump (penetrations #32 and

1. 33) and from the LPSI pumps (penetrations #36, #37, #38, and #39) and the chemical and volume control charging line (penetration #27), provide the following (see SRP Section 6.2.4 II.3.e):

A.

Confirm that the closed system outside containment is protected from missiles, designed to seismic Category 1 standards, classified Safety Class 2, and has a design temperature and pressure rating at least equal to that for the containment.

B.

Verify that the closed system outside containment will be leak tested, unless it can be shown that the system integrity is maintained during nomal plant operation.

C.

For the one outside isolation valve, provide information showing the piping between the containment and the valve is enclosed in a leak l

tight or controlled leakage housing or that the design of the piping l

and valve precludes a breach of piping integrity.

l I

D.

Provide information demonstrating the design of the outside isolation valve and/or piping compartment provides the capability to detect leakage from the valve shaft aH/or bonnet seal and teminate the l

l leakage.

I

. 1 022.67 (6.2.4)

Regarding the safety injection lines from the LPSI pumps, explain why Table 6.2-32 does not include the check valves located immediately in-side containment on each line as part of the containment isolation sys-tem.

022.6B (6.2.4)

Table 6.2-32 states that the CARS exhaust valves 2HV-F254B and 2HV-253A fail as is. Provide the necessary justification to show that this is the fail safe position as opposed to failing closed.

022.69 (6.2.4)

Provide further evidence that all containment isolation valves located outside containment have been placed as close to the containment as practical as required by GDC-55, 56, and 57, as some of the distances listed in Table 6.2-32 appear to be excessive.

022.70 (6.2.5)

The normalized decay power curve (Figures 6.2-13a and 6.2-13b) utilized to compute radiolytic hydrogen generation is not conservative compared to the NRC decay energy model given in Branch Technical Position ASB 9-2.

Either revise your radiolytic hydrogen generation calculations using the BTP ASB 9-2 model or one more conservative, or quantify the effect on the results from your use of a less conservative model.

022.71 (6.2.5)

Identify the source of the cooling water for the hydrogen analyzer system coolers and indicate if loss of cooling water will lead to failure or shut-down of the hydrogen analyzers due to the temperature of the sample gases.

022.72 (6.2.5)

Provide additional information on containment internal structures demon-strating that areas do not exist where pockets of hydrgen could collect.

. Additional information is also needed on the design features of con-tainment internal structures that promote the free circulation of the atmosphere.

022.73

.s, (6.2.5)

Clarify whetner or not the containment cooling system is used to mix the combustible gases within containment following a 1.0CA.

If so, present an analysis showint; how it will prevent excessive stratification of com-bustible gases within the containment or within a containment subcompart-ment.

022.74 (6.2.5)

Define more clearly the safety related portion of the containment cooling system duct work since it is not clear on Figure 6.2-37.

Also state 1

whether the safety related portion 07 the containment cooling system duct work has been protected against damage from missiles.

022.75 (6.2.5)

Specify whether or not the hydrogen analyzer system panel located in the Reactor Auxiliary Building is in an area which is accessible following a loss-of-coolant accident.

1 I

l l

-, ~... -. _ - - _,

Mr. D. L. Aswell Vice President, Power Production Louisiana P0::Or & Light Company 142 Delaronde Street N w Orleans, Louisiana 70174 cc:

W. Malcolm Stevenson, Esq.

Monroe & Lemann 1424 Whitney Building New Oricans, Louisiana 70130 Mr. E. Blake Shaw, Pittman, Potts and Trowbridge-1800 M Street, N. W.

Washington, D. C.

20036 Mr. D. B. Leste'r Production Engineer Louisiana Power & Light Company 142 Delaronde Street New Orleans, Louisiana 70174 Lyman L. Jones, Jr., Csq.

Gillespie & Jones P. O. Box 9216 Metairie,. Louisiana _. 70005 2

Luke Fontana, Esq.

Gillespie & Jones 824 Esplanade Avenue New Orleans, Louisiana 70116 Stephen M. Irving, Esq.

One American Place, Suite 1601 Baton Rouge, Louisiana 70825 Resident Inspector /Waterford NPS P. O. Box 822 Killona, Louisiana 70066 e

-.