ML20058F448
| ML20058F448 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 11/02/1993 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| NUDOCS 9312080120 | |
| Download: ML20058F448 (14) | |
Text
1 o fM (A I
- t..
- [re arsg'o UNITED STATES
[" )
N NUCLEAR REGULATORY COMMISSION g
E W ASHINGTON, D. C. 20555
%...+
November 2, 1993 Docket No.52-003 Mr. Nicholas J. Liparulo Nuclear Safety and Regulatory Activities Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230
Dear Mr. Liparulo:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION (RAI) ON THE AP600 As a result of its review of the June 1992 application for design certifica-tion of the AP600, the staff has determined that it needs additional informa-tion in order to complete its review. The additional information is needed in the areas of quality assurance (Q260.ll-Q260.19), reactor systems (Q440.36-Q440.48), hydrogen control and igniters (Q480.38-Q480.48), and the testing program (Q952.12-Q952.28).
Enclosed are the staff's questions.
Please respond to this request within 90 days of the date of receipt of this letter.
In addition, during a meeting in February 1993, Westinghouse committed to provide a scaling analysis for the CMT test facility based upon a to-be-agreed-upon set of " key phenomena." Westinghouse provided a suggested set of phenomena to the staff shortly after the meeting, and the staff accepted that set, subject to some comments. To date, the scaling analysis has not yet been submitted.
Provide a scaling analysis of the CMT facility, demonstrating that it is capable of providing data that will represent adequately the range of operating conditions and controlling phenomena in the full-scale CMTs in the AP600 plant.
You have requested that portions of the information submitted in the June 1992, application for design certification be exempt from mandatory public disclosure. While the staff has not completed its review of your request in accordance with the requirements of 10 CFR 2.790, that portion of the submit-ted information is being withheld from public disclosure pending the staff's final determination. The staff concludes that this RAI does not contain those portions of the information for which exemption is sought.
However, the staff will withhold this letter from public disclosure for 30 calendar days from the date of this letter to allow Westinghouse the opportunity to verify the
- The numbers in parentheses designate the tracking numbers assigned to h' j the questions.
l 2
B F. H. EHa SWY 9312080120 931102 PDR ADDCK 05200003 A
. l Mr. Nicholas J. Liparulo November 2, 1993 staff's conclusions.
If, after that time, you do not request that all or i
portions of the information in the' enclosures be withheld from public disclo-sure in accordance with 10 CFR 2.790, this letter will be placed in the Nuclear Regulatory Commission's Public Document Room.
i This RAI affects nine or fewer respondents, and therefore, is not subject to review by the Office of Management and Budget under P.L.96-511.
If you have any questions regarding this matter, you can contact me at (301) 504-1120.
Sincerely, (Original signed by)
Thomas J. Kenyon, Project Manager Standardization Project Directorate Associate Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated i
cc w/ enclosure:
See next page DISTRIBUTION:
(
- Central File PDST R/F TMurley/FMiraglia DCrutchfield
- PDR WTravers RBorchardt RArchitzel l
TKenyon RHasselberg TGody, Jr., ED0 JMoore, 15818
~
MSiemien, 15B18 Alevin, 8E23 GHsii, 8E23 TCollins, 8E23 CHoxie, llE22 RLobel, 17G21 FAllenspach, 10A19 RGramm, 9Al l
ACRS (11) w/o encl.
PShea OFC:
LA:PDST:ADAR PM; T':ADAR PM:PDST:ADAR SC:PDST:ADAR,z NAME:
PShd Tk6Nyon:sg RHas N g RArchitzel E l
1Q/ { /9b NT-/93
$/3/93 IdT2 /93 DATE:
i i
0FFICIAL RECORD COPY-DOCUMENT NAME: SRXBSCSB.RAI 4
e V
r -
r r -
w
i Mr. Nicholas J. Liparulo Westinghouse Electric Corporation Docket No.52-003 AP600 cc:
Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit i
P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. John C. Butler Advanced Plant Safety & Licensing Westinghouse Electric Corporation
+
Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 I
Mr. M. D. Beaumont 4
Nuclear and Advanced Technology Division Westinghouse Electric Corporation i
One Montrose Metro 11921 Rockville Pike i
Suite 350 Rockville, Maryland 20852 i
Mr. Sterling Franks i
U.S. Department of Energy i
NE-42 Washington, D.C.
20585 Mr. S. M. Modro EG&G Idaho Inc.
~
Post Office Box 1625 Idaho falls, Idaho 83415 i
Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
i Room 8002 Washington, D.C.
20503 Mr. Frank A. Ross U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 Mr. Victor G. Snell, Director Safety and Licensing i
AECL Technologies 9210 Corporate Boulevard Suite 410 Rockville, Maryland 20850 3
a
REQUESTS FOR ADDITIONAL INFORMATION ON THE WESTINGHOUSE AP600 DESIGN QUALITY ASSURANCE 260.11 Section 17.3 of the SSAR states that Revision 12 to WCAP-8370, " Energy Systems Business Unit -- Power Generation Business Unit Quality Assur-i ance Plan," will apply to future design, procurement, fabrication, inspection, and testing activities.
This revision, now labelled Revision 12A, dated April 1992, has been accepted by the NRC by letter dated April 23, 1992. Clarify which revision to WCAP-8370 is applica-ble to the AP600.
260.12 Table 16.2-1 of the SSAR describes those D-RAP non-safety-related systems that provide defense-in-depth or that are used in the PRA evaluation to provide credit for event mitigation.
Table 3.2.3 shows, by AP600 class, those systems for which Appendix B of 10 CFR Part 50 applies including most of the D-RAP systems. Describe the quality assurance requirements for those D-RAP systems of the AP600 that are not covered by Table 3.2-3.
260.13 Note (2) for Table 3.2-1 of the SSAR states that the NNS defined in the ANSI 51.1 standard is divided into two AP600 equipment classifica-tions; namely, Class D and Class NNS.
It is also noted in Table 3.2-1 that Appendix B to 10 CFR Part 50 does not apply to Class D.
- However, Class D contains items important to safety for which a quality program should be implemented that will be sufficient to ensure that the functions of the equipment shall be achieved.
Also, Criteria II of Appendix B to 10 CFR Part 50 states that the quality assurance program shall provide control over activities affecting the quality of the identified structures, systems, and components (SSCs) to an extent consistent with their importance to safety. Describe, in general, the quality assurance requirements for SSCs of the AP600 that are not required to meet Appendix B of 10 CFR Part 50, but that are important to safety, such as noted in Q260.12 above or in AP600 Class D (See also 0260.14 and Q260.15 below).
260.14 Generic Letter 85-06, " Quality Assurance Guidance For ATWS Equipment i
That is Not Safety-Related," provides specific quality assurance guidance required by 10 CFR 50.62.
Provide a commitment to meet this guidance or describe some other way of meeting this requirement for i
the AP600.
260.15 Describe the quality assurance program for fire protection for the AP600 that meets BTP CMEB 9.5-1.
260.16 Justify the exclusion of specific quality assurance program require-ments (AP600 Class D) for items in Table 3.2-1 of the SSAR that are designated as seismic Category 1.
Enclosure i
i
4 l
f 260.17 State whether the quality assurance requirements for construction are considered a COL responsibility by Westinghouse.
260.18 Describe the quality assurance provisions that are being applied to snftware for the design phase of the AP600.
260.19 Describe what quality assurance provisions apply to the design phase l
test programs for the AP600.
4 REACTOR SYSTEMS 440.36 The Reactor Coolant System (RCS) Piping and Instrumentation Diagram (P&lD) on p. P5.1-9 of Chapter 5 of the SSAR shows, on the parallel Automatic Depressurization System (ADS) valve groups (stages 1-3),
additional solenoid-operated valves, V007A, -B, -C, and -D.
The purposes of these valves do not appear to be described in the SSAR.
a.
What are the functions of these valves?
b.
It appears from the P&lD that the normal state of these valves is i
" closed." Under what conditions do the valves open, and is the operation automatic or manual?
?
c.
What happens to the valves on the generation of an "S" signal and on actuation of the ADS?
d.
Has this configuration been analyzed to determine whether it is possible to create an unisolated leak from the pressurizer, beyond the capability of normal makeup? If so, present the analysis of the event; if not, perform the analysis or justify the assumption that an unisolable leak cannot occur.
e.
In the failure Modes and Effects Analysis, Table 6.3-6 of the SSAR, p. P6.3-51, and on the P&lD for the Passive Core Cooling System, Figure 6.3-1 of the SSAR, p. P6.3-59, valves V007A/B are identified as " pressurizer to CMT line check valves." This definition appears to be inconsistent with the RCS P&lD cited above.
Reconcile this inconsistency, f.
In the Failure Mades and Effects Analysis, Table 6.3-6 of the SSAR, p. P6.3-51, valves V002A/B and V003A/B are identified as "CMT inlet and outlet isolation A0Vs," while on p. P6.3-57 of the same table, these valves are identified as ADS MOVs.
The same inconsistency is carried through on the associated P&lDs for the Passive Core Cooling System and the Reactor Coolant System.
Reconcile this inconsistency.
i
~
i t
. 440.37 Table 6.3-5 of the S5AR, p. P6.3-49, lists parameters for the passive RHR heat exchanger. Theproductoftheoverallheattrgnsfercoeffi-cient and heat transfer area (UA) is given as 3.80 x 10 BTU /hr*F.
l a.
On what heat transfer area (tube inner surface or tube outer t
surface) is the UA product based, and what is that area?
b.
Are the numbers for UA and for the design heat transfer for each heat exchanger or for both heat exchangers taken together?
l c.
What are the individual heat transfer coefficients on the inside and outside of the heat transfer tubes, and how were they calcu-lated?
d.
What are the tube thermal conductivity, and inner and outer diameters?
e.
What provisions in the UA product were made to account for scale on the tube inner and outer surfaces?
f.
Justify the use of 120*F as both the inlet and outlet temperature on the shell side of the heat exchanger.
g.
The outlet temperature on the primary (tube) side of the heat exchanger is shown as 224*F. However, it is stated in Sec-tion 6.3.2.2.5 of the SSAR, p. P6.3-9, that the vertical run of the C-tubes is approximately 18 ft. At a submergence of about 18 ft., assuming the containment is at atmospheric pressure, the static pressure at the bottom of the tube would be approximately 21.9 psia.
From steam tables, the saturation temperature at 21.9-psia is approximately 230*F.
Is the shell side coolant in surface r
(subcooled) boiling along the entire length of the C-tube?
If so, explain how the tube side temperature at that point nn be less than the saturation temperature of the shell side coolant.
Provide a table or plot of shell-side and tube-side water tempera-tures as a function of length along the C-tubes.
440.38 Clarify the operation of the hydrogen addition system and the provi-sions for venting hydrogen that accumulates in the pressurizer during normal plant operation.
In addition, explain the impact of these operations on the performance of reactor and containment systems, and on plant accident response.
Specifically:
l a.
The text of Section 9.3.6.1.2 of the SSAR and the associated figures of the Chemical and Volume Control System indicate that
+
hydrogen for oxygen control is introduced directly into the RCS, i
maintaining a concentration of 25-35 cc per kg (at STP) using a i
gas cylinder.
Confirm that this is the case, b.
Specify the projected losses and consumption of hydrogen that comprise the specified rate of addition of 4000 cc per minute.
L n
a i
+
t c.
Section 5.6.5 of the SSAR describes operation of the pressurizer and associated subsystems. The spray subsystem operates con-stantly to balance boron concentration and to keep the spray piping hot.
The coolant that is sprayed into the pressneizer will be stripped of much of its dissolved hydrogen, with this gas
~
building up in the pressurizer. Degassing of the pressurizer is described as being required only on ta occasional basis, and is accomplished using the stage 1 valves of the automatic depressuri-l zation system. How often is it projected that degassing of the pressurizer will be required, and what is the basis for selecting that frequency of operation?
d.
What is the volume of hydrogen projected to be removed from the pressurizer during each degassing operation?
I e.
The hydrogen vented from the pressurizer will flow through tia ADS sparger into the in-containment refueling water storage tanA (IRWST). What actions will be taken to avoid a buildup of hydro-gen above the IRWST or in the containment that could eventually result in a combustible concentration of the gas?
f.
What is the in-containment radioactive source that results from the venting of hydrogen from the pressurizer?
g.
As hydrogen builds up in the pressurizer, it will also accumulate in the piping associated with the passive safety injection sys-tems, such as pressure balance lines.
How will accumulation of hydro'en affect the performance of the passive SI systems?
Present analyses to support your evaluation.
j h.
How has the use of the ADS 1st stage valves for pressurizer degassing been accounted for in assessing the likelihood of inadvertent ADS actuatio;. as an accident initiator?
440.39 Clarify the meaning of " failure of a CMT discharge valve" discussed in the analysis of a steam system piping failure in Section 15.1.5.2.1 of the SSAR. Does this mean that the CMT flow is reduced, or does it l
mean that a single CMT is assumed to be unavailable?
l 440.40 The analysis of a steam system piping failure presented in Sec-tion 15.1.5.2.3 of the SSAR discusses 10 seconds of flow from the unaffected steam generator before the main steam isolation valves i
- close, is this 10 seconds of flow included in the transient analysis?
l r
]
440.41 In the analysis of a steam system piping failure presented in Sec-tion 15.1.5.2.4 of the SSAR, the departure from nucleate boiling ratio (DNBR) is discussed, but no figure is provided to show results for the limiting case.
Provide such a figare.
i I
i
i !
440.42 In the analysis of a feedwater system pipe break presented in Sec-l tion 15.2.8.2.1 of the SSAR, a 17-second delay is assumed for passive residual heat removal (PRHR) system actuation. What is the rationale for this assumption?
Is the assumption conservative?
i 440.43 In the analysis of the locked rotor event presented in Sec-tion 15.3.3.3 of the SSAR, an assumption is made for the analysis of radiological consequences that 18% of the fuel rods fail. According I
to the Standard Review Plan (SRP) guidance for this event, fuel rod failure is to be assumed for any rod for which DNB is predicted to occur.
Provide results of the analysis of this event that demonstrate that the assume fuel failure percentage is consistent with SRP requirements.
i 440.44 Provide a rationale for the assumption that "e locked rotor event is most limiting for the section addressing decre se in reactor coolant system flow rate. (Chapter 15.3)
I 440.45 In the analysis of a rod ejection event presented in Section 15.4.8.3 of the SSAR,15% of the fuel rods are assumed to fail.
This is contrasted to the analysis in Section 15.3.3.3, where 18% of the fuel rods are assumed to fail. What is the rationale for the difference in these assumptions?
440.46 In the analysis of the inadvertent operation of the ADS presented in Section 15.6.1.1 of the SSAR, it is stated that "the design of the fourth ADS stage valves is such that the valves cannot be opened while i
the reactor coolant system is at nominal operating pressure." How-ever, Westinghouse has stated in meetings with the NRC staff that a v
final design for the fourth stage valves has not yet been determined, and no test data currently exist for these valves demonstrating the stated characteristics. Accordingly, present an analysis for the
[
inadvertent opening of one fourth stage ADS stage, or show that the consequences of this event are bounded by other hot leg LOCA analyses.
l 440.47 In the analysis of a steam generator tube rupture presented in Sec-
[
tion 15.6.3.2 of the SSAR, steam generator overfill is discussed as i
part of the event sequence.
The assumptions appear to be reasonable, t
but no analytical results are shown (e.g., what calculational method).
Provide these results.
i 440.48 The analysis of a single steam generator tube rupture (SGTR) event is presented in Section 15.6.3 of the SSAR, with a deta11ed description of the assumptions and results in Sections 15.6.3.2.1.2 and 15.6.3.2.1.3.
Among the assumptions listed in Section 15.6.3.2.1.2 is j
the availability of offsite power or onsite ac power, since both the i
chemical and volume control system (CVS) and startup feedwater system (SFWS) are assumed to be available. These assumptions are inconsis-tent with the guidance in the Standard Review Plan to include consid-
)
eration of a loss of offsite power. They are also in conflict with Commission guidance on the use of non-safety equipment for limiting design faults, and with the EPRI Passive Plant Utility Requirements i
1
u
. Document (URD), Volume III, Chapter 5, Paragraph 1.2.2, that states that only safety-related equipment is assumed to be available for LDB events.
In addition, Westinghouse's analysis includes opening the steam generator PORV, which appears to be in conflict with Para-graph 4.2.5 of Chapter 3 of Volume III of the URD, which indicates that passive plants should be able to sustain a single SGTR without lif ting steam generator relief valves.
Provide the following informa-tion:
a.
The staff understands that the assumption of CVS flow tends to increase the primary-to-secondary break flow, thus maximizing inventory loss from the primary system.
However, makeup from the CVS also delays the injection of makeup flow from the core makeup tanks (CMTs), and thus reduces the possibility that the CMTs will drain to the automatic depressurization system (ADS) first stage actuation setpoint.
Present a new analysis assuming no onsite or offsite ac power availability to demonstrate that the AP600 is capable of being brought to a stable condition with (a) no opera-tor action and (b) use of passive safety systems only.
b.
Why does the AP600 design comply with the EPRI URD requirement for no steam generator relief valve actuation in the event of a single SGTR?
c.
The staff notes that Westinghouse's SGTR analysis requires flow from the secondary system back into the primary to stabilize the system near the end of the event.
Describe how the effects of boron dilutior, on reactivity that can result from this backflow are accounted for in the analysis model.
HYDROGEN CONTROL AND IGNITERS When reviewing the AP600 design and igniter placements, some of the 480.38 criteria that the staff uses are:
a.
The optimum placement of igniters is nearest to the source.
b.
Compartments adjacent to the break compartment should have igniter coverage.
c.
Compartments that are considered as lower probability break compartments should have igniter coverage if the compartments are accessible.
d.
Apr.ropriate test data are the NTS, HCOG 1/4 scale, and HDR pro-grams.
e.
Computer analyses, such as MAAP and WG0THIC, are a valuable tool in assessing general trends.
However, they are not sufficient in determining whether or not igniters are needed in a specific location.
i
t
., f.
Detonation calculations have an important role in the overall assessment of the design. They should be used in two capacities:
(1) to show that the compartment can survive the bounding calcula-tions, or (2) to determine the capability of the surrounding i
structures.
g.
Placement of igniters should be near the source, but not directly above so as to not be in the immediate burning zone.
h.
Equipment survivability should be addressed by determining the environment in the burning zones.
Does Westinghouse have some similar criteria that were used to decide upon the placement of the 58 igniters in the AP600 design? Describe that criteria.
480.39 As stated in Q480.38, the staff considers the optimum igniter place-ment to be near the source. For the AP600, sources include:
release through stages 1-3 of the ADS into the IRWST release through stage 4 of the ADS directly into the containment atmosphere release from the high vent points release through breaks and seal failures.
Discuss the igniter coverage for each of these locations. How was the placement and number of the igniters decided upon?
480.40 Are there any subcompartments (such as the reactor cavity) that do not
. have igniter coverage?
If so, what is the justification for not having coverage in those subcompartments?
480.41 What was the criteria used by Westinghouse to decide on the separation between igniters? On page 6.2-27 of the SSAR, it states that the separation between igniter locations is selected to prevent the i
velocity of a flame front initiated by one igniter from becoming i
significant before being extinguished by a similar flame front propa-gating from another igniter. WCAP-13388 (pg. 4-5) states that the igniters are distributed such that the distance a flame can travel before it encounters another flame front (or exits the compartment) is less than about 25 feet. What is the experimental support for choos-ing this number?
480.42 Provide details of the " multiple electrical power sources" (WCAP-13388, pg. 4-4) that may power the 58 igniters. Are they ac-powered or dc-powered? What is the pedigree of the power (Class-lE, non-safety, etc.)? Are the igniters divided into any divisions / trains?
If at least some of the igniters are not dc-powered, provide a justi-fication for not having dc-powered igniters (especially to handle a station blackout).
.l l
f r ;
480.43 Has Westinghouse performed any structural response / detonation calcula-tions for compartments in the containment? For example, if igniters are not included in the reactor cavity, a structural response calcula-tion could show that igniters are not needed there because the struc-ture can withstand the effects of a hydrogen detonation.
i 4
480.44 Discuss an event in which there is a large release of hydrogen because of a break in the ADS valve package at the 174 ft. elevation.
This appears to be at a very high point, with no igniters encountered until the containment dome is reached. What is the potential for large hydrogen concentrations and detonations from such a release?
i 480.45 Review igniter performance in the industry today.
It appears that the igniters are not very reliable, and that a considerable percentage of them are replaced at each refueling outage when they are surveilled.
i Given this poor performance / reliability, and, in addition, given that pipe whip or jet impingement may disable an igniter, why is one igniter per room sufficient (as proposed for many of the AP600 subcom-partments)?
480.46 The staff's position is that the igniters must be covered by technical specifications.
Provide a justification for not submitting a techni-cal specification for the AP600 igniters.
480.47 Has Westinghouse reviewed the following experimental databases?
Discuss the application of these programs to the AP600.
The NTS program, which is an excellent source for evaluating the
=
igniter performance within a single compartment.
I HCOG, which may provide the most appropriate data for evaluating
=
the performance of transients involving an intact primary system.
i Most notable of this group is the station blackout sequence.
This group could also be used to gain insights into hydrogen release into the spargers and out into the IRWST.
HDR, which was a test p: 4.am in which multi-compartmented large i
scale tests were conducted.
The most significant test is the one selected for the standard problem. This test is of particular interest in that the released hydrogen quickly went to the dome.
This test is also good for demonstrating code accuracy. WG0THIC calculations of HDR hydrogen distributions would provide good insights into the design.
480.48 Discuss equipment survivability and burn zones caused by igniters.
1 Within the burn zone, the environment is likely to exceed the criteria of 10 CFR 50.49.
Have all such burn zones been considered, and is safety-related equipment located outside of such burn zones?
i t
F
t l
_ g_
TESTING 952.12 The staff has reviewed the design and test matrix for ADS Phase B testing at ENEA Laboratories.
It appears that only one flowmeter is included in the apparatus for measurement of steam flow.
It is recommended that an additional flowmeter be installed on the saturated water lines. Address this comment.
j i'
952.13 The effects of non-condensible gases on the operation of the passive safety systems need to be addressed in Westinghouse's test programs for the AP600.
The presence of hydrogen in the pressurizer and its introduction into the passive safety injection piping is one example of this concern; a second is the behavior of nitrogen injected into the RCS following injection of water from the accumulators.
How will these effects be studied in the SPES-2 and OSU test programs?
952.14 The staff has previously raised the question of applicability of the passive residual heat removal (PRHR) heat exchanger (HX) testing performed at Westinghouse to the current C-tube PRHR design in the AP600. Westinghouse's responses have concentrated on conditions inside the tubes, and have attempted to demonstrate that the thermal-hydraulic conditions in the test were comparable to those expected i
i during operation of the HX. The staff has evaluated the test results 4
and Westinghouse's responses to previous RAls; the staff continues to have questions about the ability of the tests to adequately represent i
the primary side behavior of the HX.
In addition, the staff is concerned about the ability to model behavior on the exterior of the PRHR HX tubes, since the C-tube configuration (bundle) is substanti-l ally different from the 3-tube configuration tested. The staff notes that a C-tube bundle will be included in the OSU facility. However, the relatively short length of this bundle and low operating tempera-l ture of the OSU loop will allow only a very limited range of condi-l tions in the PRHR HX to be investigated in a representative geometry; i
these conditions are not those of greatest interest, such as film boiling and vapor blanketing of the upper portion of the bundle.
l i
a.
Provide details of the data base used to predict the heat transfer i
on the outside of the HX tubes, accounting for the geometry (length, tube diameter, pitch, etc.) and fluid conditions in the
[
IRWST (stratification, subcooled natural convection through pool 3
boiling and possible tube dryout conditions, thermal and vapor plumes, etc.).
t b.
Aspects of primary side behavior, such as plenum flow distribution and flow distribution among the tubes due to the "C" configuration (with each tube having different pressure drop characteristics and i
thus different flow) were not studied in the PRHR HX tests. What data base will be used to ass'st in the modeling of these para-meters over the range of expe ted operating conditions in the AP600?
i f
h m,
m o
. i 952.15 Given the experience in ADS Phase A testing of trying to hold the facility pressure constant, how will tests in Phase B be run (e.g.,
B3-6) so as to hold the pressure constant, as specified in the test
^
matrix?
Note:
Q952.16 through Q952.28 have resulted from the NRR/ Westinghouse Meetings in Italy in April 1993.
952.16 Update the information in the SPES-2 design documentation to reflect any new hardware or instrumentation, and to indicate how staff com-ments on the test matrix have been taken into consideration. With respect to the test matrix, describe the key phenomena to be studied in the SPES-2 tests, and justify the choice of these phenomena.
952.17 The following questiors refer to information obtained during the cold and hot preoperational tests on the SPES-2 facility:
a.
Summarize important insights on facility behavior that has been gained to this point during preoperational testing.
Include discussion of flow distribution and pressure losses; facility heat loss characteristics; instrumentation response and data acquisi-tion; and actual system behavior compared to analytical predic-tions.
t b.
Discuss test procedures for matrix testing, with emphasis on any insights gained from preoperational tests.
952.18 Provide the data from the cold and hot preoperational tests on SPES-2.
952.19 Provide details on the flow-matching orifices that will be used in SPES-2 to simulate steam generator outlet and cold leg pressure drops.
952.20 Provide a plot and/or table comparing volume vs. elevation in SPES-2 to that for the AP600.
952.21 Provide an analysis to show if tWe is any difference in plant response to steam generator tubt ptures originating on the cold leg side of the SG, as opposed to SGlRs originating on the hot leg side.
-l 952.22 Summarize the important information gained from the Phase A tests performed on the ADS sparger, includi, discussion on:
a.
stratification behavior in the tank and its impact on the choice of sparger elevation for Phase B testing; b.
tank / internal loads due to condensation and air clearing at the sparger, and how these test results apply to the AP600; c.
experience with water hammer during the tests; and d.
unexpected behavior of the sparger during the tests.
t
. i 952.23 The operation of the passive RHR heat exchangers in the IRWST is i
expected to have an impact on ADS operation, especially with regard to changing the temperature profile of the water in the IRWST before initiation of depressurization. Were these effects simulated in the i
Phase A tests? Will they be simulated in Phase B?
952.24 Discuss the test procedures to be used during Phase B testing of the ADS valves, including a.
duration of tests, and tank temperature profiles before and during testing (see Q952.14 on PRHR effects);
j b.
valve opening times and their relationship to AP600 operation; and c.
measurement of key parameters during tests, such as mass flows, void fractions, and transition from critical to subsonic flow.
952.25 Describe the analyses to be performed using ADS test data from both test phases. Have analyses of Phase A tests been initiated?
If so, discuss the analytical results and compare these to the test data.
952.26 Provide the following information concerning the ADS tests:
a.
How will the fluid entering the depressurization valves be condi-tioned before reaching the valves? This includes single-phase liquid, two-phase mixtures, and single-phase steam.
b.
What is the basis for choosing the specified conditions?
c.
Will the fluid experience a range of thermal-hydraulic conditions similar to that expected in the AP600 during the course of the test program?
If appropriate, refer to any pretest analyses performed on the test facility or on the AP600 design.
952.27 The staff is concerned that the full range of possible conditions under which ADS may operate has not yet been considered. Justify your selection of operating conditions. A " map" of RCS behavior of the AP600 vs. ADS test facility operation would be useful in making this assessment.
952.28 Provide information on dynamic forces associated with reflection waves off ridged surfaces during ADS operation.
i r
t E