ML20147G545
| ML20147G545 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 03/05/1997 |
| From: | Novendstern E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Huffman B NRC |
| Shared Package | |
| ML20147G451 | List: |
| References | |
| NUDOCS 9703280190 | |
| Download: ML20147G545 (86) | |
Text
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' FAX S
H E
E T
U dr,an A Mc intvre To:
Bill Huffman - NRC
Subject:
NRCAV NOTRUMP Meeting Date:
March 5,1997
)
Pages:
13, including this cover sheet.
COMMENTS:
- Bill, a
Attached is a partially complete table on the " road-map" to close OIS/RAls/etc. for NOTRUMP V&V related issues. Please give Ralph Landry a copy so he can have a chance to review before i
we meet. We'll discuss this as part of Thursday meeting. Thanks.
4 i
1 1
J cc: L. Hochreiter (FAX), B. Osterrieder, M. Young, From the desk of...
B.MCINTYRE (NRC Informal correspondence
,,,,g,y,,,,,,,,,
File), A.Gagnon Manager, Advanced and WER Plant Safety Analysis Westinghouse PO Box 355 Pittsburgh, PA 15235 (412) 374 -4790 Fax: (412) 374-5744 9703280190 970321 PDR ADOCK 0520 3
E
i i SDSER Confirmatory Description of item Reference Where Answered I
Item #
I DSER-CN 21.6.2.4-1 The t@ plication of SIMARC drift-flux is The NOTRUMP FVR (WCAP-acceptable pending confirmation of the 14807, Revision 1) has been model through benchmark and submitted.
assessment of code to be provided in NOTRUMP Final Validation Report (FVR).
DSER-CN 21.6.2.4-2 The modifications made to the The NOTRUMP FVR (WCAP-NOTRUMP drift-flux correlations are 14807, Revision 1) has been acceptable pending confirmation of the submitted.
model through benchmark and assessment of code to be provided in NOTRUMP FVR.
~
DSER-CN 21.6.2.4-3 Westinghouse needs to verify that the
.NOTRUMP code does not use the Bjornard and Griffith modification.
DSER-CN 21.6.2.4-4 Westinghouse needs to verify that heat link methodology for transition boiling is not used in AP600 NOTRUMP calculations.
DSER-CN 21.6.2.5-1 The acceptability of the PRHR model used in NOTRUMP is contingent on a finding that the PRHR data are applicable.
{
DSER-CN 21.6.2.7-1 Comparisons of the NOTRUMP code The NOTRUMP FVR (WCAP-simulations to the OSU and SPES-2 test 14807, Revision 1) has been data in the NOTRUMP FVR should submitted.
confirm the applicability or insensitivity of the NOTRUMP flow regime models to the key system response parameters.
t
- l
l SDSER Open item o Descnption of item Reference Where Answered i
DSER-Ol 21.6.2.2-1 Westinghouse needs to identify which This table identifies where RAI information from the NOTRUMP-related information is captured and RAI responses will be incorporated into closes the 01. Note that the l
NOTRUMP-related documentation.
NOTRUMP FVR is intended to be
{
the only NOTRUMP related i
documentation summarizing the NOTRUMP code for use on t
AP600 plant calculations.
DSER-Ol 21.6.2.2-2 Westinghouse needs to submit the The NOTRUMP FVR (WCAP-NOTRUMP FVR.
14807, Revision 1) has been submitted.
DSER-OI 21.6.2.4-1 Westinghouse needs to explain provisions to ensure that volumetric-based momentum equations will be used for all AP600 calculations.
l DSER-Ol 21.6.2.4-2 Westinghouse needs to submit the The NOTRUMP FVR (WCAP-l assessment cases demonstrating 14807, Revision 1) has been acceptability of casting equations in net submitted. Section 3.5 contains
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volumetric form.
the assessment cases.
1 DSER-Ol 21.6.2.4-3 Westinghouse needs to submit the After the preliminary calculations, i
assessment cases for the Horizontal this model was ng longer used.
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Stratified Flow Model.
The preliminary calculations were i
redone without this model, and i
therefore the model description is not included in WCAP-14807. As a result, the assesments are not i
needed and not performed.
DSEROI 21.6.2.4-4 Westinghouse needs to explain provisions to ensure that options to override the default flow partitioning will i
be used for all AP600 calculations.
DSER-Ol 21.6.2.4-5 Final acceptance of Mixture Overshoot The NOTRUMP FVR (WCAP-Logic must await completion of 14807, Revision 1) has been benchmark and assessment calculations submitted.
to be included in NOTRUMP FVR DSER-OI 21.6.2.4-6 Determination of additional (to G-2 tests)
Section 4 of WCAP-14807, separate effects level swell tests Revision 1 contains GE and necessary for code qualification.
ACHILLES separate effect level swell test simulations in addition to G-2.
DSER-Ol 21.6.2.4-7 Acceptance of modified pump model Benchmark submitted in Section must await submittal of benchmark 3.7 of WCAP-14807, Revision 1 calculations.
DSER-Ol 21.6.2.4-8 Acceptance of implicit treatment of Benchmark submitted in Section gravitational head await staff review of 3.4 of WCAP-14807, Revision 1 the benchmark calculations.
DSER-Ol 21.6.2.4 9 Acceptanc3 of th3 horizontal flow IIv;lizing model must await submittal B:nchmirk submitted in Secti3n and staff review of benchmark 3.3 of WCAP 14807, Revision 1 calculations.
DSER-OI 21.6.2.4-10 The staff cannot determine the adequacy After the preliminary calculations, of the birthing logic until benchmark is submitted and reviewed.
this model was no longer used.
The preliminary calculations were redone without this model for inclusion in WCAP 14807, L
Revision 1. As a result, no benchmark was performed and j
the staff does not need to review the birthing logic.
DSER-Ol 21.6.2.4-11 Acceptance of the Zuber critical heat flux correlation for AP600 SBLOCA analysis The NOTRUMP FVR (WCAP-will be determined after review of the 14807, Revision 1) has been submitted.
NOTRUMP FVR.
DSER-OI 21.6.2.4-12 Acceptance of the smoothing logic between choked and unchoked flow mustThe NOTRUMP FVR (WCAP-14807 Revision 1) has been await submittal and review of the Final submitted.
NOTRUMP Validation Report.
DSER-Ol 21.6.2.4-13 Acceptance of the logic schemes for The NOTRUMP FVR (WCAP-application of fluid node stacking, mixture 14807 Revision 1) has been level overshoot, and bubble' rise changes submitted.
must await the submittal of the assessment cases in the NOTRUMP FVR.
DSER-OI 21.6.2.5 1 The NOTRUMP code tended to The NOTRUMP FVR includes the overpredict the ADS flow rates in the preliminary OSU and SPES-2 OSU and SPES-2 simulations which were redone after the comparisons. The models affecting the preliminary calculations. Included fluid entering the ADS piping, particularly in the report are comparisons for the hot legs and pressurizer, need to be reviewed in the NOTRUMP FVR.
(test data to simulation) of ADS flows.
DSER-OI 21.6.2.5-2 CMT thermal stratification was not Section 6 of the NOTRUMP FVR captured in the CMT tests.
contains the CMT test simulations Westinghouse will further investigate which were redone after the inability to property characterize CMT preliminary calculations.
thermal stratification and these i
assessments will be provided in the NOTRUMP FVR.
DSER OI 21.6.2.5-3 The staff must receive and evaluate the Sections 5 and 6 of the' i
CMT and ADS test simulations that were NOTRUMP FVR contain these identified in Table 21.7 of the SDSER.
test simulations.
l 2
DSER-Of 21.6.2.6-1 Tha st;ff must r:c:iva and evaluite th3 Sectinn 3 of tha NOTRUMP FVR i
benchmark calculations that were contains these benchmarks with i
identified in Table 21.8 of the SDSER.
the exception of the Birthing Logic j
and Horizontal Stratified Flow t
ones which were not performed l
because the coding was not used i
in te NOTRUMP FVR i
calculations and will not be used in AP600 plant calculations.
j DSER-Ol 21.6.2.6-2 The staff must receive, review, and Section 4 of the NOTRUMP FVR evaluate the adequacy of the separate-contains the level swell related -
effects testing relative to level swell and test simulations.
l void fraction distribution.
I DSEROI 21.6.2.6-3 The staff must receive and evaluate the Sections 7 and 8 of the integral test simulations that were NOTRUMP FVR contain the j
identified in Table 21.10 of the SDSER.
integral test simulations.
I DSER-Ol 21.6.2.7-1 Westinghouse needs to address PRHR The comparisons for SPES-2 are primary-side heat transfer comparisons contained in Section 7 of the between NOTRUMP and OSU/SPES-2 NOTRUMP FVR. OSU data in the NOTRUMP FVR.
comparisons were not included because comparable test data q
j was not available.
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DSER-Ol 21.6.2.7-2 Effects of non-condensible gases on PRHR heat transfer should be addressed j
in NOTRUMP FVR.
DSER-OI 21.6.2.7-3 Clarify the use of the COSI condensation j
modelin the AP600 code.
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RAI O DIscription cf it:m Rafiranc2 Whira Answered RAI 440.325 Questions on NOTRUMP CAD (WCAP.
Westinghouse Letter i
14206) related to PIRT, NOTRUMP-NTD-NRC-95-4594; l
modeling of noncondensible gases, and WCAP-14807 Revision 1 NOTRUMP 1-D model.
Section 1.3 contains final l
I RAI 440.326 Should include an AP600 plant Westinghouse Letter nodalization and reference to SAR NTD-NRC 95-4587; I
calculations.
WCAP-14807, Revision 1
[
Section 1.2 contains AP600 plant l
noding diagram.
I g
RAI 440.327 Provide a matrix of tests that will be used Westinghouse Letter i
for assessing each of the PlRT items.
NTD-NRC-95-4610; Also, identify the models that are to be WCAP-14807, Revision 1 i
1 validated for each test.
Section 1.4 contains table of tests j
and parameters selected for l
validation of NOTRUMP for highly ranked PIRT items.
i RAI 440.328 Explain what analyses were performed to Westinghouse Letter determine the limiting failure.
NTD-NRC-95-4587 t
RAl 440.329 Describe the low flow correlations Westinghouse Letter applicable to the prediction of the single NTD-NRC-95-4610 and two-phase frirdon factors in NOTRUMP for AP600 and identify the test data that will be used for the assessment.
RAI 440.330 Describe the enhancements made to the Westinghouse Letter i
NOTRUMP code for AP600.
NTD-NRC-95-4587; WCAP-14807, Revision 1, Section 2 contains the NOTRUMP code changes for AP6M * 'Iculations.
RAI 440.331 Provide the specific inputs for the code Westinghouse Letter extemals used to perform the analyses in NTD NRC-96-4630 the SSAR calculations done in January 1994.
RAI 440.332 Provide a document describing the Westinghouse Letter methods and models comprising the long NSD-NRC-96-4780 term cooling code and describe how the -
code is initialized from the NOTRUMP code.
RAI 440.333 Justify the use of a fixed containment Westinghouse Letter pressure boundary condition since the NSD-NRC-96-4780 response of the safety systems depend on containment pressure.
.- ~ _ - - -._ --.
RAI 440.334 Provide a tIst matrix showing the Wcstinghous3 Letter i
s:parite effects and integrd tasts to be NTD-NRC-95-4610;
- y i
used in the validation of NOTRUMP for WCAP-14807, Revision 1, AP600.
Section 1.4 contains table of tests and parameters selected for s
validation of NOTRUMP.
RAI 440.335 Justification for using constant friction i
factors, particularly at low flow, flow pressure conditions are needed.
i RAI 440.336 Describe if momentum flux is included in 4
AP600 analyses and justify its omission if it is not used.
F RAI 440.337 Demonstrate that the Macbeth correlation is adequate for the low flow and pressure
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conditions expected for AP600.
RAI 440.338 Demonstrate that the NOTRUMP pump j
model can predict the AP600 pump coastdown. Describe and justify the use of the two-phase pump degradation curves for AP600 analyses.
RAI 440.339 Provide time step and nodalization j
studies to justify the AP600 nodalization.
RAI 440.340
}
Discuss the potential for boric acid build-Westinghouse Letter up and precipitation during long NSD-NRC-96-4780 j
transients for AP600.
RAI 440.341 Describe in detail the IRWST model Westinghouse Letter i
i' including how the sparger and plumes NTD-NRO-95-4587 are handled as well as their influence on i
IRWST injection and PRHR heat removal.
RAI 440.342 Provide documentation for a) NOTRUMP
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coding changes along with model j
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benchmarks, b) a description of the containment modeling approach with 1
calculations justifying model, c) a desciv' ei of the "Long Term Cooling 4
Code", d) a section presenting i
calculative methods including sensitivity studies and the full break spectrum -
.i analysis, and e) a test matrix listing the pertinent separate and integral tests used to benchmark the AP600 small break LOCA code package.
RAI 440.432 Identify where choking occurs in the ADS Westinghouse Letter tests and discuss why the asymetric NTD-NRC-95-4610 effects can be ignored in modeling the three ADS valves as a single flow path.
ll RAI 440.433 Expidn the effect of not mod: ling air in Wtstinghouse Letter v
the ADS lines on the ADS system NTD-NRC 95-4610 pressure, flow, and quality responses.
RAI 440.434 Demonstrate the ability of the NOTRUMP code to accomodate single phase steam critical flow since the ADS system is expected to transition to high quality steam flow discharge.
RAI 440.435 Questions related to ADS modeling Westinghouse Letter including explain how NOTRUMP treats NTD-NRC 95-4594 l
the void distribution and release of steam from the two-phase regions in the ADS j
lines.
RAI 440.436 Explain how NOTRUMP uses equation Westinghouse Letter 4-1 of RCS-GSR-003 in computations of NTD-NRC 95-4598 fluid quality.
i RAI 440.437 Questions on ADS test simulation Westinghouse Letter depressurization rates and length of test NTD-NRC-95-4594 simulations.
RAI 440.438 Explain the inconsistency in the Westinghouse Letter I
discussion of the effect of tank pressure NTD-NRC-95-4587 on quality in the ADS Preliminary Validation Report.
RAI 440.439 Has the NOTRUMP code been assessed Westinghouse Letter against single-phase and two-phase NTD NRC 95-4610 pressure drop test data in piping systems with expansions and contractions present?
RAI 440.440 Provide the results of a noding study Westinghouse Letter used to justify the CMT noding in the NTD-NRC-96-4622 CMT Preliminary Validation Report.
Also, provide the plots of the fluid driving heads calculated by NOTRUMP for each side of the loop.
RAI 440.441 Were wall temperatures measured in the facility in the CMT and piping? If so, provide comparisons with the NOTRUMP l
code and discuss the results.
RAl 440.442 Were wall heat structures modeled in the piping and reservoir? If not, justify the omission;if so describe the model.
RAI 440.443 Justify the reservoir nodalization and explain the effects of thermal stratification and mixing, or lack thereof, in the S/W reservoir on the NOTRUMP results.
- i t
RAI 440.444 Was a tims step study performed f:r tha CMT tests? Discuss and show that the 9
time steps used do not contribute to the error in the NOTRUMP predictions. Are the time steps consistent with those used
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j in the plant model?
l RAI 440.445 The early CMT flow rates appear to be Westinghouse Letter i
overpredicted even though the time NTD-NRC-96-4626 j
averaged flows show good comparisons.
j Discuss the NOTRUMP behavior given that the early overprediction of flow may
{
affect the RCS loop temperatures and l
system behavior later in the event.
RAI 440.446 Explain why the CMT inlet flow l
uncertainty is higher than the outlet flow uncertainty measurement for the test.
Explain this uncertainty in light of the l
NOTRUMP inlet flow rate predictions.
i RAI 440.463 Justify use of single node for,SG -
Westinghouse' Letter secondary side.
NTD-NRC-95-4587 RAI 440.464 Perform two-phase level swell WCAP-14807, Revision 1 simulations to justify core noding.
Section 4 for level swell, Sections 4.2.5 and 4.3.4 for noding RAI 440.465 Justify omission of wall heat transler Westinghouse Letter from loop piping and secondary NTD-NRC-95-4594 components.
RAI 440.466 For SIMARC drift flux model... Please WCAP-14807, Revision 1 describe how the void fraction is Section 2.2 computed for countercurrent flow conditions.
RAI 440.467 Two drift flux models were added to Westinghouse Letter NOTRUMP. Which modelis to be used NTD-NRC-95-4587; i
for AP600 cales? Explain models.
WCAP 14807, Revision 1 Section 2.3 RAI 440.468 Provide benchmark cales for level swell WCAP-14807, Revision 1 and counter current flow data to evaluate Section 4 for level swell, flooding.
Sections 3.2 & 3.3 for flooding RAI 440.469 Provide volumetric flow based WCAP-14807, Revision 1 momentum equations and code Section 2.4 for equations, benchmarks for this model change.
Section 3.5 for benchmark RAI 440.470 Questions on Horizontal Stratified Flow After the preliminary calculations, Modelin preliminary NOTRUMP report this model was no longer used.
LTCT-GSR 001 The preliminary calculations were redone without this model, and therefore the model description is not included in WCAP-14807. As a result, the RAI no longer applies.
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- ~..
RAI 440.471.
Discuss the use of partitioning models for Wzstinghouse Letter I
,9 AP600 calculiti:ns and show that there NTD-NRC-95 4598 use would not adversely affect the level 1
swell results.
RAI 440.472 Please explain the liquid reflux flow links Westinghouse Letter 4
and how their use affects level swell, NTD-NRC 95-4594 j
bubble rise, steam production, and fuel cooling.
RAI 440.473 Please explain how the mixture level Westinghouse Letter i
overshoot logic does not introduce errors NTD NRC-95-4587; F
into the NOTRUMP solution that could WCAP-14807, Revision 1 I
change the results or conclusions of an Section 2.8 l
AP600 analysis.
j RAI 440.474 Provide the derivations and the WCAP-14807, Revision 1 i
expressions for the equations comprising Section 2.9 for equations, l
the implicit bubble rise model. Provide Section 3.6 for benchmark, level sw6!! calculations verifying this Section 4 for level swell i
model.
a RAI 440.475 Provide a mathematical description of WCAP-14807, Revision 1 2
modified pump model and comparison Section 2.10 for equations, j
of the old to new model.
Section 3.7 for comparison a
RAl 440.476 Describe mathematically the implicit WCAP-14807, Revision 1 treatment of gravitational head and Section 2.11 for equations, lt provide verification analysis.
Section 3.4 for verification j.
benchmark RAI 440.477 Provide new levelizing drift velocity WCAP-14807, Revision 1 l
correlation and provide a benchmark for Section 2.12 for correlation, model.
Section 3.3 for benchmark l
i RAI 440.478 Provide a sample calculation showing After the preliminary calculations, j
how the birthing region works, this model was no longer used.
j l
The preliminary calculations were i
redone without this model, and therefore the model description is not included in WCAP-14807. As a result, the RAI no longer j.
applies.
j RAI 440.479 Provide a comparison of the NOTRUMP Westinghouse Letter Shah condensation model prediction to NTD NRC-96-4626 condensation test data demonstrating i
applicability of the model to the range of i
conditions expected in AP600.
1 RAI 440.480 Provide a comparison of the results of Westinghouse Letter j
the as implemented Zuber critical heat NTD NRC-96-4626 4
flux correlation to test data over the range of conditions expected for AP600 small break LOCAs.
1 1
~RAI 4' 40.481 Provide c:mparisons of the new i,
NOTRUMP two-phas3 fricti2n multipilgr Wutingh:uss Lett2r NTD NRC 95-4598; j
to separate effects and/or integral test WCAP-14807, Revision 1, data below 250 psia to justify the new Section 2.16 models extrapolation formulation.
RAI 440.482 Provide benchmark of the new critical flow model versus : critical flow tests to Westinghouse Letter verify the model. Describe how the NTD-NRC-96-4630; i
model treats the transition from choked WCAP-14807, Revision 1, to unchoked conditions.
Section 2.13 describes the transition from choked to RAI 440.483 unchoked conditions.
Provide results of a sample fill and drain L
calculation to demonstrate the Fluid WCAP-14807, Revision 1 i
Node Stacking Logic and provide a Section 2.18 for description, mathematical description of the logic.
Section 3.8 for demonstration t
RAI 440.484 Show the effect of the changes to the Westinghouse Letter l
transition boiling correlation on peak clad NTD-NRC-95-4594 temperature.
RAI 440.485 Describe the coding and model changes Westinghouse. Letter included in the preliminary ADS test i
simulations and CMT test simulations NTD-NRC-96-4630; These simulations were redone i
and included in WCAP 14807, Revision 1 j
RAI 440.486 Explain why in the preliminary OSU simulations the upper head drains Westinghouse Letter
[
prematurely in the tests.
NTD-NRC-95-4598; These simulations were redone and included in WCAP-14807, Revision 1 RAI 440.487 For the analyses in the OSU Preliminary i
Validation Report (PVR), provide comparisons of the NOTRUMP liquid levels in the core and upper plenum versus test data.
RAI 440.488
(
Discuss the NOTRUMP overprediction of the integrated break flow for the OSU l
PVR calculation.
RAI 440.489 j:
Explain why the NOTRUMP code i
underpredicts the PRHR heat transfer in 3
the OSU PVR and justify how this model i
f results in conservative AP600 SBLOCA ECCS performance predictions.
RAI 440.490 Explain why the NOTRUMP code verpredicts the downcomer liquid level o
during this OSU PVR calculation and justify the model result for AP600 plant calculations.
1
2 RAI 440.491 Provide the core inlit and cora bypass i
miss flow rat 3 predicti:ns f r tha NOTRUMP code.
RAI 440.492 Provide the core inlet and bypass mass flow rate predictions for the blind two inch cold leg balance line break in the OSU PVR. Also provide the liquid level plots for the upper plenum and core region and the void distribution in the j
core region.
RAI 440.493 Discuss the NOTRUMP fast depressurization rate for OSU PVR I
calculations including the break flow j
discharge coefficient and the steam generator heat transfer.
RAI 440.494 Discuss the impact of the delayed CMT-2 4
drainage on the core / upper plenum level response for the OSU PVR calculation.
RAI 440.495 Provide the upper plenum and core liquid i
level plots for this test along with the void distribution in the core.
i RAI 440.496 For this OSU PVR calculation explain why the code overpredicts the liquid inventory in the downcomer and justify that this will not lead to non-conservative predictions of the liquid level in the -
vessel for AP600 plant calculations.
RAI 440.497 Explain the statement that the NOTRUMP code allows a "short spurt of flow at the break" in reference to Figure 5.3-22 of the OSU PVR.
RAI 440.498 For this OSU PVR case, explain the reasons for the highly oscillatory behavior in the PRHR inlet flow calculated by NOTRUMP and why the code predicts a much higher PRHR flow rate.
RAI 440.499 Can the NOTRUMP code model nitrogen entering the RCS? If not, justify the omission of nitrogen effects on AP600
)
resp)nse following small break LOCAs.
RAI 440.502 RAI 440.503 RAI 440.504
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RAI 440.506 RAI 440.507 RAI 440.508 RAI 440.509 RAI 440.510 RAI 440.511 RAI 440.512 RAI 440.513 RAI 440.514 RAI 440.515 RAI 440.516 RAI 440.517 RAI 440.518 RAI 440.519 RAI 440.520 RAI 440.541 RAI 440.542 RAI 440.543 RAI 440.544 RAI 440.545 RAI 440.546 RAI 440.547 RAI 440.548 RAI 440.549 RAI 440.550 RAI 440.551 RAI 440.552 RAI 440.553
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h Westinghouse FAX COVER SHEET D
RECIPIENT INFORMATION SENDER INFORMATION 4
DATE:
3 - i c
'; 7 NAME:
(2 ;ad Hancy TO:
LOCATION:
ENERdY CENTER -
Tec b b et. ik y EAST PHONE:
FACSIMILE:
PHONE:
Office:va - 3 : y.y2 p COMPANY:
Facsimile:
win:
284 4887 l
US MRC outside: (412)374-4887
)
LOCATION:
i i
)
Cover + Pages 1+3 l
The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:
WIN: 284 5125 (Janice) or Outside: (412)374 5125.
COMMENTS:
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4
- 59. PRA Results and Insight
(
J l
Sequence 14:
Steam Generator Tube Rupture Accident (SGTR-07) l Sequence Frequency: 2.4E-09/ year Contribution to Core Damage: 1.4 percent i
Initiating Event Frequency: 5.2E-03/ year 1
Conditional Core Damage Probability: 4.6E-07 l
Description of Seauence N
0.6-iod j j
l A steam generator tube rupture initiating event (break sizeQ/0 u4 equivalent diameter)
I occurs.
Due to failures in nonsafety systems, such as startup feedwater or chemical and i
i volume control systems, or failure to identify and isolate the faulted steam generator, the event I
continues as a challenge to passive core cooling systems, similar to that of a small loss-of-l coolant accident event. One or more core makeup tanks inject into the reactor coolant system, I
and passive residual heat removal is successful. The automatic depressurization system fails I
and the pressurized reactor coolant system loses inventory through the break into the i
secondary side. The reactor coolant system inventory loss cannot be made up after the core I
makeup tanks inject, although the decay heat is being removed by passive residual heat I
removal. Core damage is postulated due to the inability to provide long-term reactor coolant i
system inventory makeup. This sequence is assigned to accident class 6E. In this accident I
class, the reactor coolant system is postulated to be at high pressure and a containment bypass 1
path through the faulted steam generator exists.
I Important Modeling Assumptions l
'Ihe success criteria for this sequence are very conservative. This sequence may not be a core i
damage sequence since the decay heat is being removed by the passive residual heat removal J
l and reactor coolant system inventory loss is made up by the core makeup tanks for a I
considerable time period. The loss of reactor. coolant is expected to be stopped when passive I
residual heat removal cooling lowers the reactor coolant system pressure, thus terminating the I
loss-of-coolant accident and the need for automatic depressurization system and gravity l
injection.
I Risk Important Failures l
Table 59-17 lists the dominant cutsets for this sequence. The dominant risk-important failure I
is the common cause failure of protection and safety monitoring system engineered safety I
feature output logic software and manual diverse actuation system actuation (51 percent).
1 This is followed by various protection and safety monitoring system actuation common cause i
failures.
I Credit is taken for the proceduralized operator action to manually actuate safety-related core I
cooling systems by using the diverse actuation system, if protection and safety monitoring I
system actuation fails.
Revision: 8 T Westinghouse
[h_
September 30,1996 59-37 maap6ceprawvn59.wpf:lb-9/2s/96
J.
k
- 59. PRA Results and Insights r
i Sequence 15:
Steam Generator Tube Rupture Accident (SGTR 23) l Sequence Frequency: 2.3E-09/ year I
Contribution to Core Damage: 1.4 percent I
Initiating Event Frequency: 5.2E-03/ year i
l Conditional Core Damage Probability: 4.4E-07 l-Description of Seauence rx
( c.co l
A steam generator tube rupture initiating event (break si'zQ.4/S bd equivalent diameter)
Due to failures in nonsafety systems, such as startupTeMwater or chemical and l
occurs.
l volume control systems, or failure to identify and isolate the faulted steam generator, the event I
continues as a challenge to passive core cooling systems, similar to that of a small loss-of-I coolant accident event. One or more core makeup tanks are actuated to inject into the reactor I
coolant system, and passive residual heat removal is successful. However, reactor coolant I
pumps fail to trip and this is assumed to prevent core makeup tanks from injecting. He i
automatic depressurization system fails and the pressurized reactor coolant system loses l
inventory through the break into the secondary side. The reactor coolant system inventory i
I loss cannot be made up, although the decay heat is being removed by passive residual heat I
removal. Core damage is postulated due to the inability to provide long-term reactor coolant I
system inventory makeup. This sequence is assigned to accident class 6E. In this accident I
class, the reactor coolant system is postulated to be at high pressure and a containment bypass i
path through the faulted steam generator exists.
(
l Important Modeline Assumptions 3
i Re success criteria for this sequence are very conservative. This sequence may not be a core I
damage sequence since the decay heat is being removed by passive residual heat removal and i
reactor coolant pumps would be tripped eventually to allow core makeup tank injection.
I Then, this sequence would behave like the previously discussed SGTR-07 sequence.
I Risk-Important Failures l
Table 59-18 lists the dominant cutsets for this sequence, he dominant risk-important failure I
is the common cause failure of reactor coolant pump breakers to open and operator to I
manually actuate the automatic depressurization system via the protection and safety I
monitoring system or diverse actuation system (over 95-percent contribution). This is I
followed by various operator actions associated with failed nonsafety systems.
I Credit is taken for the proceduralized operator action to manually actuate safety-related core I
cooling systems by using the diverse actuation system, if protection and safety monitoring I
system actuation fails.
Revision: 8 ENEL September 30,1996 wxt::6 Westilighouse m:\\np600pavev8\\59.wpf:Ib-9CU96 59-38
\\
- 59. PRA Results and Insights 1
Sequence 18:
Consequential Steam Generator Tube Rupture Accident (SGTRC-03) i Sequence Frequency: 2.lE-09/ year l
Contribution to Core Damage: 1.0 percent Initiating Event Frequency: 6.8E-05/ year l
Conditional Core Damage Probability: 3.lE-05 l
Description of Sequence
- 0. tim A l
A consequential steam generator tube rupture initiating event (break si W6-meh equivalent I
diameter) occurs. The starting point of this event may be a transient or a sMiine break I
(or a stuck-open secondary-side valve). One or more core makeup tanks inject into the reactor I
coolant system, and passive residual heat removal and the automatic depressurization system i
are successful. Normal residual heat removal fails and reactor coolant system inventory I
makeup by in-containment refueling water storage tank gravity injection is successful. Sump I
recirculation fails. Core damage is postulated due to the inability to provide long-term resctor i
coolant system inventory makeup and core cooling following failures of the normal residual I
heat removal system and sump recirculation. This sequence is assigned to accident class 6L.
I In this accident class, the reactor coolant system is fully.depressurized but a potential I
containment bypass path through the faulted steam generator exists.
1 Important Modeline Assumptions l
This sequence may not be a core damage sequence since the decay heat is being removed by I
passive residual heat removal and reactor coolant system inventory loss is made up. The loss I
of reactor coolant is expected to be stopped, thus terminating the loss-of-coolant accident and i
the need for sump recirculation, due to low reactor coolant system pressure terminating the I
break flow.
I Risk-Importart Failures I
Table 59-21 lists the dominant cutsets for this sequence. The dominant risk-important failure 1
is the common cause failure to open of explosive valves on recirculation lines (83-percent I
contribution). This is followed by common cause failure of in-containment refueling water I
storage tank level transmitters and operator action to open sump recirculation valves I
(15-percent contribution).
I Credit is taken for the proceduralized operator action to actuate normal residual heat removal, I
and sump recirculation (if automatic actuation fails). Credit is also taken for operator action I
to actuate core rnakeup tanks and the automatic depressurization system as a backup to I
automatic actuation. These actions are not risk-important in this sequence.
Revision: 8 T Westirighouse hh_
September 30,1996 59-41 mW6MprWev8\\59.wptib-9/28/96
f FAX to DINO SCALETTI March 10,1997 CC:
Sharon or Dino, please make copies for:
Bill Huffman Ted Quay Robin Nydes Chip Suggs Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #169 (M5.2.5-26)
In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am researching open items from the smallest item number on. The relevant entry from OITS related to Open Item #169 (M5.2.5-26) is attached. This item is a good example of why we need to maintain some discipline in-statusing our open items. This item as ORIGINALLY asked was satisfied in the August TechSpec revision. However, as a result of that revision, another question was asked on the same section. The new question is related in that it addresses ISLOCA, however, the sinrple request to correct the entry for a reference in TS 3.4.8 was resolved in August. This item (#169) should be closed on two counts. First, because we corrected the reference in August of 1996 and second, l
because a more relevant question has been asked by NRC as Q24 (OITS #4970). It seems a l
reasonable request that NRC at least acknowledge receipt of the change associated with item #169.
We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed". Thank you.
1 l
l Jim Winters 412-374-5290 l
l l
l
k.%
AP600 Open Itm Tracking System Datzbase: Exec tivaSummary Datn 3/10/97 Selecties:
litem no] between 169 And 169 Sorted by item #
i licm DSER Section/
Titic/ Description Resp (W)
NRC No Branch Question T)pe Detail Status Engineer Status Status Letter No. /
Date 169 NRR/SPLB
' 5.2.5 MTG.OI TECllSPEC/Suggs C.
Closed Action W M5.2.5-26 (REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE) 13 3 4.8 bases section " Applicable Safety Analyses" refers to reference 4 instead of relerence 3.
appropnate reference will be referenced. rkn 3/29 i
Closed - With issuance of the Tech Specs in SSAR Rev. 9.
Action W - We need to exphcitly tell NRC where in TechSpec we cover ISIMA, especially in light of carher TechSpec RCS pressure boundary valve isolation 3 4 8 per Chapter 5 telecon with NRC on 12/2/96.
'Ihe original review conurent was resolved in the August 1996 Tech Spec revision. The new question regarding where ISLOCA is addressed in the Tech Specs is logged as Q24 of OITS item 4970. In general, that response (being issued today) states that ISI OCA is a risk-based issue and therefore does not meet the NRC NUREG-1431 criteria for inclusion in the Tech Specs. Because this issue is a duplicate of 4970 Q24 and the original question for item 169 was addressed, this item is W-status = Closed rin 3/10/97.
b
-b N
i I
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i Page: 1 Total Records: I
y FAX to DINO SCALETTI March 10,1997 CC:
Sharon or Dino, please make copies for:
Diane Jackson Ted Quay Don Lindgren i
Richard Orr Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEMS FOR SSAR Chapter 2 This is a background package for the remaining open item for SSAR Chapter 2 for your action.
SSAR Chapter 2 is of interest because by our joint NRC/W schedule, the FSER for this section should be turned into Projects by the end of April. There are 3 Open Items (547,556 and 4997) with NRC Status of Action W. These items (OITS report attached) have been discussed repeatedly with NRC and the technical description is included in the item's " Status Detail." Items 547 and 4997 were discussed as part of the Senior Management Review on March 3,1997. We will be providing a markup of Chapter 2 to address items 547 and 4997 by March 14, 1997. This markup and any subsequent discussions should be used for preparation of the FSER. The agreed to changes to the SSAR for Chapter 2 will be included in Revision 12. Item 556 was resolved by changes included in Revision 10 (December,1996, over 2 months ago) of the SSAR. We believe that no further Westinghouse action is required for item 556. It seems a reasonable request that NRC acknowledge receipt of this information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N" or " Closed".
Thank you.
Jim Winters 412-374-5290
l e..,
1
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AP600 Open item Tracking System Database: Executive Summary Date: 3/1967 l
Selection:
[nrc si code l=' Action W' And [DSER Sectionl hke '2.** Sorted by item #
Item DSER Settson/
Tetle/tksenpanon Resp (W)
NRC No Branch Question Type Detail Staus Engurer Status Staus letter No /
Date 547 NRRECUB 2.5432 DSER-OI Orr /NRCBM Aaion W Action W NSD-NRC-96-4738
- Westmghouse should consuler and document in the SSAR, the effect[s of differential settlement _
]
~
Closed Letter NSD-NRC-96-4738 provided a respume for tlus issue UkhessAferential settienent between buddings ' In'pricular, the rnain s'eam U
line leavmg the aus. buddmg and einenng the turbine buildmg is considered NRC Status Update provuled m September 5,1996 letter:
) As agreed dunng the telephone conferena on May I,19%, Westinghouse should clanfy in SSAR section 3.8.5 tha the nuclear island inasemat will be desagned and constructed giving due consideranon to the effects of constnsction sequennal loadmg.11as wdl be treated as Confirmatory item 2.5 43-2.
! Action Westinghouse j
ilhas will be mairessed with the response to item 768 IDSER OI 3 8 5-10).
_ j 556 NRR/ECGB 2.548-1 DSER-OI Orr/Landgren/BIC Closed Action W NTD-NRC-95-4433 4/3/95 Westmghouse should add C(X. Actum item 2.5 4 8-I to the SSAR, requinng that the COL applicant discuss and evalume site specific static and dynanuc f
I'*I c.arth ressures and_hydrustatic groundwater pressures acting on plant safety-related facahties.
Closed - Comtuned lacense nem included in SSAR Rev 2, secum 2.5.4 I 8.
'NRC Status - Review tespome to request fa specific interface regiarenent for lucral earth pressure. Response was al as other site parameters provide for a
[
jsite with accegnable laseral earth pressures.
Sept 5,1996 letter inconectly provided NRC posnion on item #556 and status was clianged to Action - W. The comment was moved to stem # 547 Ole 2.5 4 3-2. NRC to confirm that item # $56 is resolved.
. Action W - Westmghouse will identify SSAR reference to AP600 safety-related facihues. NRC letter of 12/W96 lSSAR Revtsson 10 included infunnanon in second parag;raph of Secuan 1.2 on sne-specific structures that addresses this p 4997 NRR/ECGB 25 RAIOl Orr Actum W Action W lRAI 23134 '- Sne Design Paraneters
' ~
~
Westmghouse uses the term
- interface requirements
M i
g This temunology is unacceptable. Westinghouse should use " site parameter" or " site design parameter
- in accordance with 10 Cf-lt 52 47(aWiii) and past prutur on the Advanced Boshng Water Reactor ( ABWR) and System 80+ designs Interface requireneras are different than sne parameters as distmguished in 52 47(a) Westinghouse is sequested to review its SSAR, particularly those sections cited above, and revise it accordingly.
l i
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Page: 1 Total Records: 3
VV ridMiligllUUS$
evsvvvcnoncci 8
ame 4
RECIPIENT INFORMATION SENDER INFORMATION DATE:
Aqag,./ /O /997 NAME:
=l < al Dh Te>t.r TO:
LOCATION:
ENERGY CENTER -
b/tt / OFFMW EAST PHONE:
FACSIMil2:
PHONE:
Office:t/n -3 7y-ya9o COMPANY:
Facsimile:
win:
284 4887
() 5 jg/#_C outside: (412)374 4887 LOCATION:
s Cover + Pages 1+3 The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:
WIN: 284 5125 (Janice) or Outside: (412)374 5125.
COMMENTS:
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- 5. Heretor Cool =t System cd Conected Systems normal level signify a possible increase in unidentified leakage rates and alert the plant operators that corrective action may be required. Similarly, increases in contairunent sump level signify an increase in unidentified leakage. The following sections outline the methods used to collect and monitor unidentified leakage.
5.2.5J.1 Containment Sump Level Monitor Leakage from the reactor coolant pressure boundary and other components not otherwise identified inside the contairunent will condense and flow by gravity via the floor drains and other drains to the contairunent sump.
A leak in the primary system would result in reactor coolant flowing into the contairunent sump. Leakage is indicated by an increase in the sump level. The contairunent sump level is monitored by two seismic Category I level sensors. The level sensors are powered fnnn a safety-related Class IE electrical source. These sensors remain functional when subjected ta a safe shutdown carthquake in confonnance with the guidance in Regulatory Guide 1.45.
The containment sump level and sump total flow sensors located on the discharge of the sump pump are part of the liquid radwaste system.
Failure of one of the level sensors will still allow the calculation of a 0.5 gpm in-leakage rate within I hour. 'lle data display and processing system (DDS) computes the leakage rate and the plant control system (PLS) pmvides an alarm in the main control room if the average change in leak rate for any given measurement period exceeds 0.5 gpm for unidentified leakage.2 Unidentified leakage is the total leakage minus the identitied IeilagCTheTeaEge rate algorithm subtracts the identified leakage directed to the sump.
The measurement interval must be long enough to permit the measurement loop to adequately detect the increase in level tha' would correspond to 0.5 gpm leak rate. and yet shon enough to ensure that such a leak rate is detected within an hour. The measurement interval is less than or equal to I hour.
When the sump level increases to the high level serpoint, one of the sump pumps automatically stans to pump the accumulated liquid to the waste holdup tanks in the liquid radwaste system. The sump discharge flow is integrated and available for display in the control tuom.
l Procedures to identify the leakage source upon a change in the unidentified leakage rate I
into the sump include the following:
i Check for changes in contairunent atmosphere radiation monitor indications.
=
Check for changes in containment humidity, pressure, and temperature,
=
l Check makeup rate to the reactor coolant system for abnormal increases.
lL
< ! ? ' < '
- D i <- h>L i; O OL GYH.
n,,.
lievision: 10 December 20,1996 5.2 24 W Westinghouse l
I l
g
- 5. R: actor Cool:nt System end Courected Systems I
l l
Check for changes in water levels and odier parameters in systems which could leak l
water into the contaimnent, and l
Review records for maintenance operations which may have discharged water into the
=
contairunent.
5.2.5.3.2 Reactor Coolant System Inventory Halance Reactor coolant system inventory monitoring provides an indication of system leakage.
Net level change in the pressurizer is indicative of system leakage. Monitoring net makeup from the chemical and volume control system and net collected leakage provides an important method of obtaining infonnation to establish a water inventory balance. An abnormal increase in makeup water requirements or a significant change in the water inventory balance can indicate increased system leakage.
The reactor coolant system inventory balance is a qu:.ntitative inventory or mass balance
)
calculation. This approach allows determination of both the type and magnitude of r
leakage. Steady-state operation is required to perform a proper inventory balance l
calculation. Steady state is defined as stable reactor coolant system pressure, temperature, power level, pressurizer level, and reactor coolant drain tank and in-contairunent refueling water storage tank levels. The reactor coolant inventory balance is done on a periodic basis arkt when other indication and detection methods indicate a change in the leak rate.
The mass balance involves isolating the reactor coolant system to the extent possible and observing the change in inventory which occurs over a known time period. This involves isolating the systems connected to the reactor coolant system.
System inventory is determined by observing the level in the pressurizer.
Compensation is provided for changes in plant conditions which affect water density. The change in the inventory determines the total reactor coolant system leak rate. Identified leakages are monitored (using the reactor coolant drain tank) to calculate a leakage rate and by monitoring the intersystem leakage. The unidentified leakage rate if then calculated by subtracting the identified leakage rate from the total reactor coolant system leakage rate.
TC.1 ME 6 c.13 SPN, Since the pressurizer inventory is controlled during normal plant operation through the level control system, the level in the pressurizer will be reasonably constant even if leakage exists. The mass contained in the pressurizer may fluctuate sufficiently, however, i
to have a significant effect on the calculated leak rate. The pressurizer mass calculation
)
includes both the steam and water mass contributions.
Changes in the reactor coolant system mass inventory are a result of changes in liquid density. Liquid density is a strong function of temperature and a lesser function of pressure. A range of temperatures exists throughout the reactor coolant system all of which may vary over time. A simplified, but acceptably accurate, model for determining mass changes is to assume all of the reactor coolant system is at T
)
l 1
Revision: 10 3 Westinghouse 5.2-25 December 20,1996 I
I
r-n
(
{
- 5. Re ctor Coolint System and Connected Systems L
4 The inventory balance calculation is done by the data display and processing system with additional input from sensors in the protection and safety monitoring system, chemical and volume control system, and liquid radwaste system. The use of components and sensors in systems required for plant operation provides conformance with the regulatory guidance in Regulatory Guide 1.45 that leak detection should be provided following seismic events that do not require plant shutdown.
5.2.5.3.3 Containment Atmosphere Radioactivity Monitor Leakage from the reactor coolant pressure boundary will result in an increase in the radioactivity levels inside containmeru.
The containment atmosphere is continuously monitored for airbome gaseous radioactivity. Air flow through the monitor is provided by the suction created by a vacuum pump. Gaseous and NdFn concentration monitors indicate radiation concentrations in tbc containment atmosphere.
/ arL.
- ' d % * ' '-
S The gas channel can resporyd) rapidly tArcacitreookm: pref = h=dag kige. Nn v
j sa neutron activation product'which hp proportional to power levels. Atiditionally - Nn-hos-a relatively short half life and consequently wi!! rerh ajunibrium wpidiy. An increase in activity inside contairunent would therefore indicate a leakage from the reactor coolant pressure boundary. Based on the concentration of NdFn and the power level, reactor coolant pressure boundary leakage can be estimated.
,CN uti sat.-r.c.5 W i d
-4N/k o
The N dFn monitoring system h= a high r~tivity when the reactor is operating at a power range higher than 20 percent. The Q rnonitor is seismic Category 1. Confonnance with the guidance that leak detection shoufd be provided following seismic events that do not require plant shutdown is pmvided by the seismic Category I classification. Safety-related Class IE power is not required since loss of power to the radiation monitor is not consistent with continuing operation following an eanhquake. jd level.~in nue-hour-a-leak-lewhf 0.5 hpm caiilWdemdcd._
& 20 pan pawer-Operating experience has
, -indicated the average long-tenn leakage (from sampling losses, collected leakoffs, and
[
unidentified leakage to the containment) from the reactor coolant system ranges between 0.1 and 0.3 gpm. The No concentration will increase by at least 25 percent above an i
existing 0.1 gpm leakage backgmund and almost 10 percent for an existing 0.3 gpm
\\ leakage. Both increases are well within the sensitivity of the NdFn monitor capabilities.,
~
Radioactivity concentration indication and alarms for loss of sample flow, high radiation.
and loss of indication are provided.
Sample collection connections permit sample collection for laboratory analysis. The radiation monitor can be calibrated during power operation.
5.2.5.3.4 Containment Pressure, Temperature and Humidity Monitors 1
Reactor coolant pressure boundary leakage increases containment pressure, temperature, and humidity. values available to the operator through the plant control system.
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Revision: 10 h.[.' ([%~
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r FAX to DINO SCALETTI March 10,1997 I
CC:
Sharon or Dino, please make copies for:
Bill Huffman Don Lindgren Robin Nydes Ed Carlin l
Earl Novendstern Ed Cummins l
Bob Vijuk Brian McIntyre OPEN ITEMS FOR LOFTRAN This is a background package for the remaining open items for LOFTRAN for your information.
LOFTRAN is of interest because by our joint NRC/W schedule, the FSER for this section should be turned into Projects by the middle of April. There are 2 Open Items with NRC Status of Action W.
Both of these items still require some Westinghouse action. Thank you.
I D
Jim Winters 412-374-5290 i
i l
k, I
f
g av AP609 Open Item Tracking System Database: Executive Sununary Date: 3/10/97 I
Selection:
[nre sa cc:lel=' Action %" And [ resp engl like loft *' Soned by item #
hem DSER Sec.mn/
Tule/Descnption M*5P (W)
NRC No.
Branch Questum Type-Detal Status Engueer Status Status Lener No /
Dane
'3134 NRR/SRXB 21 6.1.7-2 DSER G LOITRAN/R.----
Action W Action W
!21.6.8.7-2 i
'Wesunghouse needs to idenefy the infonnrion provided in RAI sesponse thas wdl be incorporned into the LOFIRAN find vanficanon and wahdaron iVa V) dtunnens (WCAP-14M7) or the code apphcalniny document (WCAP-14234).
..... _ _... _ _ __ _ _ _ _ _ _ _ _ _ _ _ _. _. _ _ _ ~... _... _. _
Westinghouse lenes NSD-NRC-96-4814. dased 9/5/96, explans the W-action so sevise WCAPs 14307 and I4234 to incorporne infonnanon from RAl. _. _ _. _
l yesponses. Expected _Qum is 12/20f96. skn ll/I5/96 3222 NRIUSCSB 21 6.5-27 DSER G Ldtran/Carlin.Ed Actum W Actum W
^
~
~ ~ ~ ~ ~
- 21.6.5-27 1Wesanghouse needs to demons *rane the acceptalnisy of IDFTRAN for the calctdarson of the MSLB mass and energy selease for the AP600 IMBA
[evahsation enodel.
'[
~
['
.tespume k@ wnsen and is eth_by the,end of Nov. rin 11/15/96 '
~ l TN t
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1 Page: 1 Total Records: 2
A vQ W
Westinghouse FAX COVER SHEET emum pje I f /7 RECIPlE*dT INFORM ATION SENDER INFORM ATION DATE:
3//g'/f>7 NAME:
Q, [ p //
R kg,,,yg LOCATION:
g,g,,;j, Q TO:
Jg(f kg jpyf
.sfjz -3 rll c; jgy PHONE:
PHONE:
COMPANY:
Q(
LOCATION:
[f 4,,//g, ////,
FAX:
(412) 374 5099 1
I Cover + Pages 1 + /b
/7 44/
Y j
REMOVE ALL STAPLES i
PENCIL WILL NOT TRANSMIT USE BLACK PEN 4
PLEASE MAKE COPIES OF TWO SIDED PAGES j
Comments:
- Jim, Attached are the following documents: (1) Human Factors Engineering design description and ITAAC, pages 3.2-1 thmugh 3.2-8; (2) markup of the Minimum inventory design description and ITAAC for the main control room,4 pages; and (3) markup of the minimum inventory design description and ITAAC for the remote shutdo;m mom,4 pages. These have been reviewed and approved by my management and are forwarded in advance of the fomial copies. We have not yet decided where to place the remote shutdown room ITAAC. We may place it with the Data Display and Processing System design description and ITAAC. You will Mso notice that a few minor changes have been made to the ITAAC on task analysis as compared to the draft that I faxed you on December 19, 1996. If you have any questions or comments following your review, please call me at 412-374-5104.
Thank You, Steve Kerch Phone Number of Roceiving Ecuipment:
. T/,n 4,,m,,j Jo / - #/f-2.2.2 2 Y
1 4
Certified De:ign Material th f.2
'/7 HUMAN FACTORS ENGINEERING
=
Revision: 33 Effective: 4444AMi2/28/97 3.2 Human Factors Engineering Design Description j
%: main ecm:ct reem !CP1 p:ctid= a f=i'i:y =d =:cu== for de 2f: cent:ct and epen:icn of 6: p!=:.The AP600 human-system interface (HSI) will be developed and evaluated based upon a human factors engineering (HFE) program. The HSI scope includes the main control room (MCR) and the remote shutdown room (RSR). The HSI scope provides the displays, controls, procedures, and alarms required for normal, abnormal and emergency plant operations. Implementation of the HFE program involves the completion of the following human factors engineering analyses and plans.
1.2-3: '1C R :=lude: : ::==:= cpen:= wcd::atic=, = :=!===:= cp:=:= n ci::::icn.
Wey-related-thsplay. =d ufe:y =l ::d ec=c!;.The integration of human reliability with human factors engineering design is performed in accordance with the implementation plan.
Critical human actions (if any) and risk important tasks are identified and used as an input to the task analysis activities.
- 2. b F MCR p=vids : =:d!: cd:p=: ="=nm=
fr =: by MCR cp=::=. Task analysis is performed in accordance with the task analysis implementation plan. Task analysis identifies 4
the information and control requirements for the operators to execute the tasks allocated to them.
3.
3: h =
y:::r :.:d=: (HS!) =:c== =:i!d': :: 1: MCR cp=:=: !=! d: $: :!=-
y:::r, p!=: :nfer.c:!= :y:::r,==p;::-!::d p c=d = :y;::r, :d:y =!:::4.*Sp!:y:,.va!!
p=:! : '--.::i: sy:::r, =d =:.:=!: (;;f: =d d:d!=M).The HSI design is performed in accordance with the HSI design implementation plan. He HSI design includes the functional design of the operation and control centers and the HSI resources, the specification of design guidelines, the detailed HS! resource design specifications, and the man-in-the-loop concept testing.
4.
3: MCR =d 6: =:i!d': HS! pm.i:===:ica of MCR :=h: hy MCR cp=::= :: Op=:e de p!=: =d m:in.!: p!=: Meef-An HFE program verification and validation implementation plan is developed. The plan establishes methods for conducting evaluations of the HSI design.
5.
The HFE program verification and validation is performed in accordance with the HFE verification and validation plan and includes implementation of the following activities:
a.
Task support verification b.
HFE design verification c.
Integrated system validation d.
Issue resolution serification e.
Plant HFE/HSI verification 3.2 1 W95tiflgh0058 mwmocNTAAcsvev:hto302 w#1t>-030697
(
C:rtified De:ign M teri:1 y'*) e 3 cf17 Ht'AAN FACTORS ENGINEERING Revision: 23 Effective: 44WM962/28/97 Inspection, Test, Analyses, and Acceptance Criteria Table 3 2-1 specifies the inspections, tests, analyses, and associated acceptance enteria for the MCR.
4 e
I i
3.2-2
[ W8Silfigt10US8 m wTAAcsvevntm2 wot:to 030697
Certified Design Material rfg 1 /7 HUMAN FACTORS ENGINEERING Revision: (3 Effective: 4 IWW962/28/97
)
Table 3.21 Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment inspections Test. Analyses Acceptance Criteria
- 1. 2-E: '!CP nid:,
^
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cp:=:e i,tu ; n:
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ep:=:c
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af::y -i::d of the documentation associated wk:y -6::d hp!::.1 =d di,;t;. =J uf::y 6::d with the integration of human uf :y :&=d =nt:6.
wawoh-The integration of human reliability analysis with human i
r-liability analysis with human factors engineering design will be A report exists and concludes that i
factors engineenng design is performed.
cntical human actions (if any) j performed in accordance with the and risk important tasks were implementa6on plan.
identified and examined by task analysis, and used as input to the HS1 design, and procedure development.
- 2. 4, 7.: 'tCR p:: :dn :
> S= C:r E:d D=ign 't=nd.
- S
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=:tH: " =b:p=: = --- =-
- .b=:ic-2.? ', '!u&r E!=d
- b=:ic: 2.? ' 'I;& r':!=d f= =: by h: '1CR S =3!:=:ir: V=t!=ic:
'I:n =d!:=tr: Y= 6:ic:
- pn::=. Task analysis is
- SyAem, SyAem-A report exists and performed in accordance with the concludes that function based task analysis implementation plan.
") S:: C:r E:d S:!;; 't.:-id, task analyses were conducted in
=bma= 2 2.5, 'iCR E=r;=:y conformance with the task
"-bitbi!!:y Sy:=--
analysis implementation plan and include the following functions:
": E= C:-uE:d D=ig ' hand,
- _b=::!:: 24 ?. C!= !S & rd Control reactivity; control RCS U"S Sy:==.An inspection of the boron concentration; control fuel task analysis documentation will and clad temperature; control be performed RCS coolant temperature, pressure, and inventory; provide RCS flow; control main steam pressure; control SG inventory; control containment pressure and temperature; provide control of main turbine.
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HUMAN FACTORS ENGINEERING Revision: 33 Effective: 4444462/28/97 Table 3.21 (cont)
Inspections. Tests. Analyses, and Acceptance Criteria Design Commitment inspections. Test. Analyses Acceptance Criteria c:: C:::-t D:.;;n I
W:: :a!. =b=:., 2 2 5 '!CR E ::;;n:y 4h:b', S;. :m A report exists and concludes that operational sequence analyses (OSAs) were conducted m conformance with the task analysis implementation plan.
OSAs performed include the following:
- plant heatup and startup from post refueling to 100% power;
- reactor :np. turbine tnp. and safety injection;
- natural circulauon cooldown (startup feedwater with SG);
- loss of reactor or secondary coolant:
post LOCA cooldown and depressunration;
- loss of RCS inventory during shutdown; loss of ILNS during shutdown; manual ADS actuation;
- manual reactor tnp via PMS, via DAS;
- ADS valve testing dunng mode 1 3.2-4
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l Effective: 44AW962/28/97 Table 3.21 (cont)
Inspections, Tests Analyses, and Acceptance Criteria Design Commitment inspections Test, Analyses Acceptance Criteria
{
- 3. R "" ren: :::: cu:!;ti :c
-pee - ': 45!
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-, p ! =
r -- : = =
- h: 'ICP are:=. " 5:
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-==pu::"::d p:=:Jur; ped ed An inspection of the
-a f:.
ed A p!;>
- .;f::y :!=:d disp!:y:
HS1 design documentation will be
^ ' ! p = :' : ^'-- : = ^ -, :-
ea!! p=:' Hf :===, :-
performed.
=d : = -^'; 3: =d
=d ; ::c!: D-': =d d:di;;::dt ded::=:dt A report esists and The HSI design is performed in concludes that the HS! design accordance with the HSI design was conducted in conformance implementation plan.
with the implementation plan and includes the following documents:
- Operation and Control Centers System Specification Document i
Functional requirements and design basis documents for the alarm system, plant informadon system, computenzed procedure system, wall panel information system, soft controls, and the qualified data processing system.
- Design guideline documents for the alarm system, plant information system displays, computerized procedure system.
and soft control displays.
Design specifications for the alarm system, plant information system displays, computenzed e
procedure system, qualified data processing system displays, wall panel information system displays, and controls (soft and dedicated).
Man-in-the-loop concept test
- reports, l
1 l
I I
1 3.2 5 I
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l HUMAN FACTORS ENGINEERING Revision: 23
=
Effective: 4444442/28/97 1
1 Table 3.21 (cont)
Inspections Tests. Analyses, and Acceptance Criteria Design Commitment Inspections. Test. Analyses Acceptance Criteria A R: '!CR =d 'h: =2:!d!:
7: = = d = :!y =: cf'h:
T': : ; =d =:!;
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.--5;;n:;'
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u_..._a,,,_..._,u,_u_,
serification and validation
- p ::== me stC R u e!
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.b E'E " C.. 2. ~ * "' ' '
implementation plan is developed.
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4
-) Sf ag S: p!= ': : =f:.
2 ) :=:.= ===:=. rip rd
- d'; :'-:: f !!^"
- i
- din ei;
- p=iEd =:!d=tA report exists and concludes that the HFE verification and validation plan
- ) ^ =id==:
was developed and includes plans for the following activities:
,_ _n <___i.,_,_ _r _.,__.
aseedom
!= ;; t a d 'e n of = c!= :
Task support verification see,dem
- HFE design venfication
...x ':= t'no
. Integrated system validation f=d:::: ':= tad
. Issue resolution verification
. :n ;: :=:= :d:
- Plant HFE/HS! verification swpwo. An inspection of the HFE venfication and validation plan will be performed.
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HUMAN FACTORS ENGINEERING Revision: 33 L
Effective: 4WS4#962/28/97 i
Table 3.21 (cont)
Inspections, Tests, Analyses, and Acceptance Criteria Design Commitment inspections, Test. Analyses Acceptance Criteria
- 5. The HFE program ventication A report exists and concludes and validation is performed in that accordance with the HFE verification and validation plan
- a. An mspection of the
- a. Task support venfication was and includes implementation of documentation for the task support conducted in conformance with the following activities:
venfication will be performed.
the implementation plan and
)
includes venfication that the
- a. Task support venfication information and controls provided
- b. HFE design venficauon by the HS! matches the display
- c. Integrated system validauon and control requirements
- d. Issue resolution venficauon generated by the function based
- e. Plant HFE/HS! venfication task analyses and the operational sequence analyses.
i
- b. An inspection of the
- b. HFE design verification was documentation for the HFE design conducted in conformance with venfication will be performed.
the implementation plan and includes venfication that the HSI design is consistent with the AP600 specific design guidelines developed for each HSI resource.
- c. Tests and analyses of the
- c. 'Ihe test and analysis results following plant evolutions and demonstrate that the MCR transients, using a facility that operators can perform the physically represents the MCR following:
configurauon and dynamically represents the MCR HSI and the i) Heat up and start up the plant operating charactenstics and to 100% power responses of the AP600 design, will be performed:
ii) Shut down and cool down the plant to cold shutdown i) Normal plant heatup and startup to 100% power iii) Bnng the plant to safe shutdown following the specified ii) Nonnal plant shutdown and transients cooldown to cold shutdown iv) Bring the plant to a safe, iii) Transients: reactor trip and stable state following the turbine trip specified accidents 3.2-7 WW$0m m:wAAcsvevntcao2 wpt 1b.030697
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Revision: 33 Effective: 44WS4/962/28/97 Table 3.21 (cont)
Inspections, Tests. Analyses, and Acceptance Criteria Design Commitment inspections Test Analyses Acceptance Criteria IV) Accidents:
)
small-break loss-of-coolant accide:..
large break loss-of coolant accident steam Ime break
- feedwater line break steam generator tube rupture
- d. An inspection of the
issue resolution venfication will conformance with the be performed.
implementation plan and includes venfication that human factors a
i issues documented in the design issues tracking system have been addressed in the final design.
- c. An inspection of the plant
- e. The plant HFE/HSIis HFE/HSI design venfication consistent with the HFE/HSI documentation will be performed.
verified in Sa. through 5d.
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{
1 Tableg Minimum Inventory of Displays andW " y
_ _ Controls Description Control Display M gp Yes YO
\\
p'eutron Flux 1MechQ ^
~
Yes Yt5 Qg Reactor Coolant System (RCS) Pressure Wide Range Hot Leg Temperature Yes
]q Wide Range Cold Leg Temperature Yes ny i
L 3, rt, Containment Water Level Yes yg b;
'd Containment Pressure Yes
%g
[
Yes Pressunzer Water Level Lurf Pressunzer Reference Leg Temperature Yes n
Yes
/
Pressunzer Pressure I4edh Core Exit Temperature Yes
/eS I 4
Yes
/e5 RCS Subcooling
],19,g. @ ~
In Containment Refueling Water Storage Tank (IRWST) Water Level Yes
/s Passive Residual Heat Removal (PRHR) Flow Yes
/gf Yes
/c.5 PRHR Outlet Temperature Yes Passive Conta2nment Cooling System (PCS) Storage Tank Water Level Yes PCS Cooling Flow Yes IRWST to Normal Residual Heat Removal System (RNS) Sucuon Valve
/g Status Thntainment Isolation Valve StstusN Yes gNemM Yes
/c$
l
.J Containment Area High-Range Radiauon Level rent Yes S
Containment Pressure (Extended Range)
Yes Containment Hydrogen Concentration
. TManual Reactor Trip Yes Lie ei Manual Safeguards Actuation Yes
_, Manual Core Makeup Tank Actuation Yes jk' Note: Dash (.) andacmas not apphcath.
I 1
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ys yes
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Certified Design Motorial PROTECTION AND SAFETY MONITORING SYSTEM Revision: 2 Effective: 10/31/96 1
Table 2.5.2 5 (cont)
Minimum Inventory of Displays and Fixed Position Controls
(, i A" d A
)
Description Control Display 4lce kAutomauc Depressurization System (ADS) Stages 1,2. and 3 Iniuadon Yes h usl-
\\
$ ADS Stage 4 Initiation Yes Manual PRHR Actuation Yes Manual Containment Cocling Actuation Yes Manual IRWST Injecuon Actuation Yes Manual Containment Recirculauon Actuation Yes Manual Containment Isolation '" '- "' " -
Yes l
Manual Main Steam Line Isolauon Yes
)
Manual Feedwater Isolation Yes 4
Manual Containment Hydrogen Igniter (Nonsafety Related)
Yes Note. Dasa H in&cstes nos apphcable.
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s PFIOTECTION AND SAFETY MONITORING SYSTEM m-Hevision: 2 Effective: 10/31/96 The PMS. in conjunction with the operator workstations, provid 7.
functions:
wmal a a) The PMS provides for the minimum inventory of display d fixed position controls, as identified m Table 2.5.2 5. in the main control room (M
).
bi The PMS provides for the transfer of control capability from the MCR to the remote shutdown room (RSR).
- n. m q 1: 9.-.L.
- :,,,,,,,,,,,,_,, 77 e g ; = 3 7. -.r:3, =
,, y,,; -
R
- Tdh if : f.... d~ R n 1 :r r2 - 2: RSR i.~.. ::d r '- '-d pr:..x 8.
a) The PMS automatically removes blocks of reactor trip and engineered safety features actuation when the plant approaches conditions for which the associated function is designed to provide
)
protection. These blocks are identified in Table 2.5.2-6.
b) The PMS automatically produces a reactor trip or engineered safety feature initiation upon an attempt to bypass more than two channels of a func. on that uses two-out-of four initiation logic.
3 c) The PMS provides the interlock functions identified in Table 2.5.2-7.
9.
Setpoints are determined using a methodology which accounts for loop inaccuracies, response testing and maintenance or replacement of instrumentation.
- 10. The PMS hardware and software are venfied and validated through a program that provides confirmation that system functional requirements are properly and correctly implemented in the delivered hardware and software.
Inspections, Tests. Analyses, and Acceptance Criteria Table 2.5.2 8 specifies the inspections, tests, analyses, and associated acceptance enteria for the PMS, 2.5.2-2 W85fillgfl00$8 mw4owTAACsvev2.newwo20502.wof:1ts110696
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Revision: 2 Effective: 10/31/96 2
Table 2.5.2 5 Minimum Inventory of Displays andD' ".f. Controls Description Control Display
/)lgry
,Muk,
' y'eutron Flux Yes Yd5
~
sgdg Reactor Coolant System (RCS) Pressure Yes Ye.6 Yes Wide Range Hot Leg Temperature
]
Wide. Range Cold Leg Temperature Yes Yes K3 L 3 g e-l-Conwnment Water Level khdaI 6 Containment Pressure Yes
/d5 Yes
%3 Pressunzer Water Level L5 +
Pressunzer Reference Leg Temperature Yes Pressunzer Pressure Yes Yes
/cS Im r4 Core Exit Temperature
[e5 Yes RCS Subcooling
[Q Yes
$pr3 In-Conwnment Refueling Water Storage Tank (IRWST) Water Level Yes pj Passive Residual Heat Removal (PRHR) Flow Yes J/c.5 l PRHR Outlet Temperature Yes Passive Conwnment Cooling System (PCS) Storage Tank Water Level PCS Coolmg Flow Yes Yes IRWST to Normal Residual Heat Removal System (RNS) Sucuon Valve v.
/g" Status Yuntainment !solauon Valve StstusN Yes g
i hmk Yes
/d 3 Containment Area High-Range Radiauon Level
'op6 Yes Contaanment Pressure (Extended Range)
Containment Hydrogen Concentrauon Yes 7
Manual Reactor Trip Yes 1 e4 Manual Safeguards Actuauon Yes Yes Janual Core Makeup Tank Actuauon Note: Dun a inceses not apphcable.
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PROTECTION AND SAFETY MONITORING SYSTEM Revision: 2 Effectivs: 10/31/96 Table 2.5.2 5 (cont)
Minimum Inventory of Displays and Fixed Position Controls
[g Ad d4
)
Description Control Display Algf kAutomaue Depressunzabon System (ADS) Stages 1. 2. and 3 Inidauon Yes pthua l -
- ADS Stage 4 Inidauon Yes Manual PRHR Actuanon Yes Manual Containment Cooling Actuauon Yes Manual IRWST Injecuon Actuation Yes Manual Contaanment Rectreuladon Actuation Yes Manual Containment Isoladon Am.m, Yes Manual Main Steam Line Isoladon Yes 4
a Manual Feedwater Isolauon Yes Manual Containment Hydrogin !gniter (Nonsafety Related)
Yes Note. Dash t l ind. cates not appbcable.
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- TX COtFIRMATION REPORT **
AS OF MAR 10 '97 09:03 PAGE.01 lJETSO/RM 468 EC EAST DATE TIME TO/FROM MODE MIN /SEC PGS CMDu STATUS 02 03/10 08:53 301 504 2222 G3--S 09'28" 017 OK I
e i
i l
o:N FAX to DINO SCALETFI March 11,1997 CC:
Sharon or Dino, please make copies for:
Bill Huffman Diane Jackson Ted Quay Robin Nydes Chip Suggs Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #172 (M5.2.5-29)
In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am researching open items from the smallest item number on. The relevant information from OITS related to Open Item #172 (MS.2.5-29) is attached. We provided the original comparison to STS with NSD-NRC-96-4833 on October 11, 1996. We then provided probability risk assessment information related to the differences from STS with NSD-NRC-97-4939 on January 14,1997. This was reiterated in the RAI responses provided by NSD-NRC-97-4972 of February 6,1997. This item
(#172) was asked by a technical branch other than the Tech Spec branch and requests justification specific to a single TechSpec section. The letters identified above were in response to questions asked by the Tech Spec branch and provide generaljustification for Action Times. Included in the general justifications are specific entries for TS 3.4.9, the subject of this item #172. Please help us provide the branch to branch coordination required to obtain proper review of this information. We believe that the letters identified above resolve the concerns of item #172. We requested your action to change the NRC Status of this item on February 14, 1997. Since NRC should be responsible to review information submitted by Westinghouse, it seems a reasonable request that NRC acknowledge receipt of the information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N". Thank you.
G Jim Winters 412-374-5290
(
.. _...+...
AP600 Open item Tracking System Database: Executive Summary Date: 3/11/97 Selection:
litem no] hetween 172 And 172 Sorted by item 0 liern DSER Sectmn/
Tule/thnptmn Resp (W)
NRC No Hranch Questmn Type Detad Status Engurer Samus Status letter No. /
Dese 172 NRR/SitB 52.5 MTG-Ol TECllSPEC/Suggs, C.
Oosed Actson W
.M5.2.5-29 (nEACTOR GX)LANT PRESSURE BOUNDARY LEAdGE) STS 3.415 states that,should the contaannwnt air' cooler condensme flow ^
~
~
rate rnmutor bemne inoperable, a channel check should te performed on the - -- - ~
atmosphere radmacuvity morutoronce per B hours. The AP600 lTS 3 4.9 simes that a grab sample should be performed once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Wesunghouse should provide jushficanon segardmg the acceptabshty of the
!ahemane acton l Action: submit T.S. 3 4 9 wah June % rev. rka 3/28 i
'r Ckmed - With issuance of the Tech Specs in SSAR Rev. 9.
- Action W - Need an explananon of Acton Times as tLey relate to STS.
l Closed - Apphcable information provided in NSIANRC-%4833 of 10/11/96. NSD-NRC-97-4939 of 1/14/97 and NSD-NRC-974972 of 2N97.
I S
oF i
Page. 1 Total Records: I
k
- )
FAX to DINO SCALETTI March 11,1997 CC:
Sharon or Dino, please make copies for:
Bill Huffman Ted Quay Don Lindgren Mike Corletti Ed Cununins Bob Vijuk Brian McIntyre OPEN ITEM #164 (M5.2.5-20)
This item should now be becoming an embarrassment In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am researching open items from the smallest item number on. The relevant documentation related to Open item #164 (MS.2.5-20) is attached. We provided this FAX response on January 10, 1997. We resent the FAX with a request for NRC Status change on February 12, 1997. We believed that this list of references resolved the concerns of item
- 164 and subsequent telephone conversations. We believe that it is an NRC responsibility to review Westinghouse submittals and it seems a reasonable request that NRC acknowledge receipt of the information provided on references. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N".
Thank you.
C>
Jim Winters 412-374-5290 i
- >y AP600 Open Item Tracking System Database
- Executive Summary Date: 2/12M7 Selection-Dtem no] between 164 And 164 Sorted by liem #
leem DSER Section/
Title /Descnpeson Resp (W)
NRC No Branch Quession Type Detal Status Engwieer Status Starus im No /
Dese 164 NRR/SPLB 52.5 MTG01 Corletti.M Closed Actum W M5 2120 (KEAC'IOlt Ct MM ANT PRESSURE llOUNDARY II.AKaGE) Identify each system cunnected to the seactor coulant system (RCS) that I
[
g could capenence innersystem ledage and provide a discussaan of the leak desectsua method, anchading proseceve features to ensise shas the syseem dues ame.
- cverpressunze.
Cloned. Wewsaghannw has coengueted nearswy submissals to supg=wt staff review See the response for RAI 440132 for a discussion ef the's isnue.
l Action W - per 12/2 selecon. Weimnghouse no provide explien sefercam to where we covered the systems connected to the RCS un the SSAR or other dalarnent.
[ Action N - FAX to Huffman on II10N7 provuled emphen references.
v i
L i
N
&u l
t Page: 1 Total Records: I 6
.m..
.m.
m
l1e
>t*
i 'g FAX to BILL HUFFMAN January 10, 1997 CC:
Don Lindgren Mike Corletti Brian McIntyre ADDITIONAL INFORMATION FOR OPEN ITEM 164
~
I This is in response to the 12/2/% request to provide, explicitly, where we covered leakage from each system connected to RCS in the SSAR or other document. We explicitly cover intersystem leakage from the RCS in WCAP-14425, the ISLOCA report. This WCAP is referenced in the SSAR in a number of places. The most relevant are in section B-63 of SSAR section 1.9, and in st:bsection 1.9.5.1.7. We believe this completes Westinghouse actions required for Open item 164 and request NRC direction to change its "NRC Status". We recommend " Action N".
I l
Jim Winters 412-374-5290 i
l l
I j
1
m.
t
')j y FAX to DINO SCALETTI i
March 11,1997 CC:
Sharon or Dino, please make copies for:
Bill Huffman Ted Qacy Don Lindgren Bruce Rarig Bob Osterrieder Earl Novendstern Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEMS FOR NOTRUMP This is a background package for the remaining open items for NOTRUMP for your information.
NOTRUMP is of interest because by our joint NRC/W schedule, the FSER for this section should be turned into Projects by the middle of April. There are 37 Open items with NRC Status of Action W.
Twenty nine (29) of these items still require some Westinghouse action. The remaining 8 (3144, 3147,3148,3149,3150,3157,3158 and 3159) have been answered by information in either NSD-NRC-96-4851 of 10/18/% or NSD-NRC-96-4863 of 10/28/% (over 4 months ago). It seems a reasonable request that NRC acknowledge receipt of this information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of these items. We recommend " Action N". Thank you.Thank you.
Jim Winters 412-374-5290
[
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AP600 Open item Tracking System Database: Executin Summ:ry Date: 3/lW97 Selection:
[nrc st codeliAction W' And [ resp eng] like **trum* Sorted by item #
hem DSER Sectionf Title /Descnptum Resp (W)
NRC No.
Branch Question Type Detail Status Engineer Status Staus leuerNo /
Dase r
2608 NRR/SRXB 15 RAIOl TRUMP /Novermissem/O Actum W Actam W
[WCAP-14206 (NOIRUMP CAD)
~
440.335 Page 411. nem no. 2. "Fnctsunal wessure drops",jusuracanon for uung constant inctum factors, pamcularly at kiw flow, low pessure t
condmons a:e needed Please demonstrae ihm the use of constant frictaan fxtes are adequae for simulaxm of AlWX) at low psessure, low flow
.condamms.
2609 NRR/SRXB 15.
RAlot TRUMP /Novendstem/O Actum W Actum W
~
I WCAP-14206 (NOTRUMP CAD) 440 336 Page 4 11, item no. 3
- Momentum Equarum," momentum flux has tren shown for convenrmmal plants so be a second order effect and has twa (excluded in many small break LOCA analyses. Please descnbe of momentum flux is included in the APMU analyses. If not, justificanon for its onunum is
!alw needed
)
2610 NRR/SRXB 15.
RAlol TRUMP /Novendssem/O Actam W Actum W
- WCAP-14206 (NOTRUMP CAD) i'
{440 337 Page 4-12, item de,"CilF correlanon,"the use of the Macbeth conclamn needs to le justified for low pressure, low pressure condmons Please hm.=*== that the Macbeth conclation is adequase for the low flow and pressure condamms expected for APM10.
26ll NRR/SRXB 15.
RAlot TRUMP /Novendstern/O Action W AummW jWCAP-14206(NOTRUMP CAD) 440.338 Page 4-13 nem 6," Pump numielmg," please demonstrare that the NOTRUMP pump model can predict the AIW10 pump coastdown. Also
- desenbe and jusefy the use of the twcwphase pump degradation curves for use in AP600 analyses.
2612 NRR/SRXB I 5.
RAl-OI TRUMP /Novendseern/O Action W Actum W l
.WCAP-14206 (NOTRUMP CAD) 440 339 On Page 4-16 nem 2, it is stated the sina no change to the numencal scheme has been made to NOTRUMP that no samtmg nor tune step studies are needed. The INEL disagrees with this stasement. Since the successful performana of the passive safety systems depend on the anurate nmmlelmg of the small pressure differences that charactenze AP600 phenomenological behavior, node and ume step size can affect the magnitude of these small pessure differences dnv:ng the flow in the system. Please provide time step and nodalizanon studies tojushfy the AP600 smulahzatum
)
2615 NRR/SRXB 15.
RAIOl TRUMP /Novendseem/O Action W Actum W i
WCAP 14206 (NOTRUMPCAD) 440.342 It is customary to provide a calculanve methods document separase from the documents &scnbmg the code phyucal nunki changes, tenchmarks, and twealt spectnen analyses. Sina many codes are used to perform a single analysis and in view of the fact that the long term coolmg code and contamment models have yet to be descnted, it would be useful to provide a calculative methods report detalmg how all of the various codes are meerfaced
- and used to produce the break spectrum analyses. This document should desenbe the innial condamms and provide sensitnity studies akmg wnh
'justificanon for the nodahzatum, models, and
,' " pertainmg to AP-600. It should also contan the small break liUA spectrum analysis. In summary, the NOTRUMP Apphcabahty Docurient should contain the following informanon:
9 o A NOTRUMP code sectum desenbing all changes to the code to accomnumlate AP-600 akeg wnh nuntel tenchmarks, i
o A descnption of the -
-- nanlehng approach wth calculations justifying the numiel, I
o A descnption of the "Long Term Cooling Code" descnbing the nethods, use, and code benchmarks Also present results of the analyses and descnbe 1
how the code is interfaced wah NOTRUMP, o A section presenung the calculmise methods includmg sensitivity calculatums jusufymg each of the ctmles compnsing the AP-600 small tweak LOCA i
, analysis package and the full tweak spectrum analysis, and o_A test mains hsung the pemnent separate and integral tests used to tenchmark the AP400 small tweak LOCA ctmle package.
i Page: 1 Total Records: 37 t
m--
w
e AP600 Open item Tracking System Database: EXecurve Summary Date: 3/11/97 Selection:
[nre st codel=' Action W' And l resp eng] like *trum' Sorted by item C ltem DSER SectKmf Titic/Descnptam Resp (W)
Ngc No Branch Questum Type Detad Status Engineer Staus Status Letter No. /
Date 2922 NRIUSRXB 15 RAl 01 TRUMP /NovemisteswO Actum W Actum W NOTRUMP CMT PVR (MT01-GSR4)II) 440 441 Were wall temperatures measaned in the facility in the CMT and pigung? If so, how does the NOTRUMP code compare to these daa Please provide the compartsons and discuss the results.
2923
' NRR/SRXB 15.
RAlol TRUMP /Novendsrerrv0 Actnm W Actum W NOTRUMP CMT PVR (MT01-GSR4)ll) 440 442 Were wall hem stnactures modeled in the ppmg and reservoir? If not, please justify the onussaan? If so, please desenhe the model and mesh structure in the all malls where wall heat was simulated;
)
2924 NRR/SRXB 15 RAlot TRUMP /NovendsternM)
Actum W Actum W NOTRUMP CMT PVR (MT01-GSR-Cll)
{
440.443 Please confirrn and justify the reservoir nodaluation? Fig. 3-1 indicates that a smgle rumle was nukleled Please justify the nodalizarson and emplain the effects of thermal stranfication and mining, or lacht thereof. in the S/W reservost on the NOTRUMP results.
j 2925 NRR/SRXB 15.
RAIDI TRUMP /Novendsiern/O AammW Action W
~
~NOTRUMP CMT PVR (MTDI-GSR 4)ll)
I 440 444 Was a tune step study performed for these nests? What tirne steps were used to sirnulate these tests? Please discuss and show ths the tine steps l used_do not contnhute to the error in the NOTRUM_P predictams. Are the tirn_e steps consi_ stent with those used in the plant nunlet?
2927 1A; NRR/SRXB 15 RAICI TRUMP /NovendssernA)
Action W Action W b
NOTRUMP CMT PVR (MTUI-GSR4)ll) p 440 446 As snenuoned in Sectum 5 0, please summanze the referenced seport and triefly explain why the inlet flow imceitainty is lugher than the outlet flow uncertainty prasurement for the tests. Please explain this uncertainty in light of the NOTRUMP inlet flow raec predictavis.
3140 NRR/SRXB 21.62.21 DSERol TRUMP /Novendseem/O ActKm W Action W 2162.2-1 Westinghouse needs to identify which information from the NOTRUMP4 elated RAI response wdl be formally incorpursed into NOTRUMP-relsed documentarum such as the final venficanon and vahdation report, the code apphcainhty document (WCAP-14206), or the SSAR.
314i NRR/SRXB 21622-2 DSERol TRUMP /Novendstem/O Actum W Actkm W i
21.6 2.2-2 Westmghouse needs to submit the final ven_facs_non_ and valaision report.
3142 NRR/SRXB 21.6.2.4 1 DSERCI TRUMP /Novendseerrv0 Acsum W Action W yesunghouse needs to emplain what provision will be uel to ensure that volumetric-based nunnentum equaines udt te used for all AP600 calculatnms.
3143 NRR/SRXB 21.6 2.4-2 DSER-OI TRUMP /Novendstern/O Actum W Actum W 21 6.2.4-2 Wesunghouse needs to subnut the NOTRUMP assessnent cases to demonstrate the adequacy of the re<astmg of the nuwnentum equzum armi dnft flun
[equaLums in net volunrtne form Page-2 Total Records: 37
AP600 Open Item Tracking System Database: Executive Summary Date: 3/11M7 Selectiott:
[nre st code l=' Action W* And [ resp engl like **trum' Sc:ted by item #
liem DSER Section/
Tale /Desenpum Resp (W)
NRC No Branch Question Type Detail Status Engmeer Status Staus Letter No. /
Date 3144 NMR/SRXB 216243 DSER-OI TRUMP /Novendstern/M Gosed Actum W 21 6.2.4-3 Westmghouse needs to sutmut the assessnent cases to demonstrate the acceptahday of modificatums to the transient terns in the momentum equaram of NOTRUMP.
~
~
~
Closed - Response provided via Westinghouse letter NSD-NRC-96485l daed October I8,1996.
}
3145 NRR/SRXB 28624-4 DSER-OI TRUMP /Novendstern/O Action W Action W 2162.4-4 Westinghouse needs to explan what provision will be used m NOTRUMP to ensure that options to overnde the defauh flow putsteomng will be used for all AP600calculsions.
3146 NRR/SRXB 21624-5 DSER4)I TRUMP /Novendstem/O Action W Actum W
.21.6 2.4-5 Westinghouse needs to ccunplete all benchma;i and assessment calculmions (to be included in the FV&V reptxt) to demonst: ate the acc@ lay of the
,kyic mod (ications for apphcation of the NOTRU_MP code to the AIV10 S_BL(X'A.
3647 NRR/SRXB 21624-6 DSER-OI TRUMP /Novendstern/O Closed Actam W
,216.24-6 Westinglese needs to deternune wtiether to use additional separase effects level swell tests to support the Quahficatum of the NOTRUMP code to address
- the code's capahday to ppredict tw& phase level swell and system mass inventory (see Open item 216 2.6-2)
Closed Respume provi&d via Westinghouse lettNNSD'NRb'96-4860' dated October 25. '1996.
i I
3148 NRR/SRXB 28624-7 DSER4.)I TitUMP/Novendstern/O Closed Action W k
21.624-7 k
Westmghouse needs to submit benchmari calculanms to demonstrate that the modified pump numiel is reasonable for apphcatum of NOTRUMP to the Q
AIV10 SBLOCA.
j Closed - Response presided sia Westmghouse letter NSD-NRC-964851 dated October is,1996 3149 NRR/SRXB 2862.4-8 DSER4)I TRUMP /Novendstem/O Closed Action W 7
2162.48 Westmghouse needs to sutmut benrhmark calculatums to demonstrae the acceptabahty of the changes made to the NOTRUMP gravitational head term, j
[
and applicatnhty to the AIV10 SBLtEA.
Closed - Response provided via Westmghouse letter NSD-NRC-964851 dated October 18,1996.
3150 NRR/SRXB 21.6.2.4-9 DSER-Ol 11tUMP/Novendstern/O Closed Actam W I
21.6.2A9 Westmghouse needs to subnut benchmari calculmions (to be mcluded in the FVAV) to dempnstrate the acceptatwhty of the namiel changes and additions. j Closed - Response provided via Westmghause letter NSD-NRC-96485l' daed Ochter 18.1996.
l I
3151 NRR/SRXB 21 6.2.4-10 DSER-OI TRUMP /Novendstern/M Action W Actam W 21624-10 Westinginarse needs to submn tenchmxL calculanons to demonstrate the aaeptabihty of the adequacy of the NOI' RUMP turttung logic, and as j
[
i apphcatnhty to the AP600 SBLOCA.
i f
i Page: 3 Total Records: 37
l w
w AP600 Open Item Tracking System Database: Executive Summary Date: 3/11/97 Selection:
lnre st etxlel=' Action %" And l resp eng] like **trum** Sorted by item #
Item DSER Section/
1stlelDescnpum Resp (W)
NRC No Branch Question Type Detail Status Engineer Staus Staus tener No. /
Dze 3152 NRR/SRXB 21624-11 DSER4)I TRUMP /Novendstern.H Action W Actum W
- 2162.4-11
- 11
- e NOTRUMP FV&V report and assessnrnt calculations need to demonstrate the acuptabihty of the Zuber cruscal heat flus conclaion for APtdK)
SBLOCA analysis.
?
3153 NRR/SRXB 2162412 DSER-Ol TRUMP /Novendstern/M Actam W Actum W
'21.6 2.4-12
{
,The NOTRUMP FV&V report needs to de==or* the aaeptainhty of the snuxnhing logic.
l 3154 NRR/SRXB 2162.4-13 DSER4)I TRUMP /Novendsterrd)
Actam W Actum W 21.6.2 4-13 4Westinglumise needs to submit the assessnrnt calculaimns to demonstrate acceptable logic opersnm and logic interactions dunng the FV&V of the Al'ta)
- NOTRUMPcode.
3155 NRR/SRXB 216.25-1 DSER4)I TRUMP /Novendsterno Aamm W Actum W I
'21 A.2 5-1 l
]
't
- Westmghouse needs to address the models affecting the fluid entenng the ADS pegung, particularly for the tan legs and pressuruct in the FV&V report.
3156 NRR/SRXB 21.6.2 5-2 DSER4)I TRUMP /Novendseem/Il Actum W Action W i
{216.23-2
~
l jWestmghouse needs to investigate the NOTRUMP code's inabdity to properly characterne CMT thennal stratification and to better explan some of the l
differences in CMT discharge flow compan_ sons. _
l 3157 NRR/SRXB 2162.5-3 DSER4)I TRUMP /NovendsternM)
Gud Action W I
21625-3 Westinghouse needs to submn its reanalysis of peviously anMyzed component seperase<ffects tests thas are hsted in Table 217 to denumstrate the
[
acceptabihty of these tests.
{
i Closed - Analyses providel via Westingluwase letter NSD-NRC-96-4863, dmed October 28.1996.
j 3158 NRR/SRXB 216.26-I DSER-Ol TRUMP /NovendstemK)
Omd Action W 21.6.2 6-1 l
Westmghouse needs to sulmut benchmark calculatums to denumstrate the acceptainhty of the NOTRUMP nudel changes and addities for which these I
benchmark calculatums are to be peiformed.
Closed - Response provided via Westinghouse letter NSD-NRC-96-4851_ dated Octaher 18,1996.
l 3159 NRR/SRXB 21626-2 DSER-OI TRUMP /Novendstern/O Gmed Actum W 21626-2 l
Westinghouse needs to demonstrate the oserall adequacy of the seperate effects testing relating to level swell and vomi fractum distnbutum in the t
NOTRU MP code (see Open Item 216 2.44).
Closed - Response provided via Westinghouse leuer NSD-NRC-96-4860 dated October 25.1996.
l 3160 NRR/SRXB 21 6.2.6-3 DSER-OI TRUMP /NovendssernM)
Actum W Adam W 21 6.2.6-3 Wesunglansse needs to submit reanalyses of the insegral sysaems tests hsted in Table 21.10 Page: 4 Total Records: 37 i
~--,
AP600 Opesi item Tracking System Datathese: Executive Summary Date: 3/1157 i
Selection:
[nre st codel=' Action W' And [ resp eng] like irum** Sorted by item #
v hem DSER Section/
Title /Descnptum Re5P (W)
NRC No Branth Questum Type Detail Staus Engpneer Samus Staus letter No. I Dme 3161 NRR/SRXB 2162.7-1 DSER-OI TRUMP /Novendstern/K Actum W Actum W
,21 6.2.7-I
{Wesonghouw needs to suturut PRHR pnmary side heat transfer conpansons between NOTRUMP and OSU/ SPEC-2 daa in the FV&V reptwt; 3162 NRIUSRXB 2162.7-2 DSER4)I TRUMP /Novendssern/II Actam W Actum W
]
~216 2.7-2 The NOTRUMP FV&V report needs to address the etTects of noncondensible gases on PRHR hem transfer.
3229 NRR/SRXB 21624-3 DSER-CN Trump Actum W Actam W
.2162.4-3
(
l Westinghouse needs to venfy the the NOTRUMP code does not use the_Bjornard and Gnffith unhfication i
3230 NRR/SRXB 2162.4-4 DSER-CN TRUMP /NovendsternK)
AaneW Actum W 121.6 2.4-4 l
Westinghouse needs to venfy that heat hnk nrtimmlology for transitum bothng is ruit used in APW) NOTRUMP cakutations.
_ j t
3232 NRR/SRXB 21 6.2.7-1 DSER-CN TRUMP /Novendssern/O Actum W Actum W i21 6.2.7-1
,Compansons of the NOTRUMP code simulations to the OSU and SPES-2 sest daa in the FV&V report shouki canfinn the apphcabihty or insensitivity of the NOTRUMP flow regimes nmxlels to the Ley system response parameters b
1 hes i
i i
i Page: 5 Total Records: 37 1
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and Ih 5 04 it in a ieHm, onct Orh W%
G C omfnhvfd - o iac ock F in tte Md SSAt cWcm G E rmg ayrmMs').
k N p A 2/n 7.6.2.3 Interlocks for the Accumulator Isolation Valve and IRWST Discharge Valve The accumulator isolation and IRWST injection isolation valve operators are nonsafety-related since the valves are not required to change position to mitigate an accident. The SSAR Chapter 15 safety analyses assume that these valves are not subject to valve mispositigning (prior to an accident) or spurious closure (during an accident). Valve mispositioning and spurious closure are prevented by the following:
(dot) The AP600 Technical Specifications, SSAR Section 16.1, require these valves to be open and power locked out whenever these injection paths are required to be available. The accumulators are required to be available when the RCS pressure is above 1000 psia. Both IRWST injection lines are required to be l
available in Modes 1, 2, 3. One IRWST injection line is required to be available in Mode 4, 5, and in Mode 6 with the reactor upper internals not removed and the refueling cavity not filled.
(dot) The AP600 Technical Specifications, SSAR Section 16.1, require verification that the MOVs are open every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. They also require verification that power is removed every 31 days.
(dot) With power locked out, redundant (nonsafety-related) valve position indication is provided in the main control room and remote shutdown workstation. Valve position indication and alarm are provided to alert the operator if these valves mispositioned. These indications are powered by different nonsafety-related power supplies.
In addition, the valves have a confirmatory open signal during an accident (S-signal for accumulator MOVs i
and ADS stage 4 signal for IRWST MOVs). The valves also have an automatic open signal when their close permissive clears during plant startup. The confirmatory open and the automatic open control signals are provided to the valve operator by the nonsafety-related plant control system.
,8 19 I\\
3 FAX to DINO SCALETTI March 6,1997 t
l CC:
Sharon or Dino, please make copies for:
Bill Huffman Ted Quay Don Lindgren Mike Corletti 4
Ed Cummins l
Bob Vijuk j
- Brian McIntyre l,
OPEN ITEM #157 (M5.2.5-13) i j
In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am researching open items from the smallest item number on. The relevant documentation related to Open Item #157 (MS.2.5-13) is attached. We provided the attached FAX response on January 9, j
1997 (two months ago). We believed that this list of references resolved the concerns of item #157 and subsequent telephone conversations. It seems a reasonable request that NRC acknowledge receipt of the change. We requested this acknowledgement on February 12,1997 (almost a month ago). We
{
understand that NRC must determine if the information provided is adequate, but this determination 1
itself is an NRC action. We request that NRC provide a definitive action for Westinghouse or j
provide direction to change the status of this item. We recommend " Action N". Thank you.
i Jim Winters l
412-374-5290 i
i I~
l a
d i
4
AP600 Open Item Tracking System Database: Project Manage:nent Summary Date: 3/6/97 Selection:
[ item nol between 157 And 157 Sorted by item #
Cotwd/ Resp Engmect Tale /Descnotion 3"
DSER Sectent issue Clouwe l*mh (W)
NRC f
No.
Branch Questmn Type Staus Detal Res Est (hn)
Staus Staus ICP Draft Review Transnut l
157 NRR/SPLB 52.5 MTGot Lindgren. D.
/ Corletti.M I
. Closed Acton W 1/20N5 A 5/24/95 S
'f M515-13 (REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE) Identify each syseem ihm's susceptible to irwenystem leakage, discuss the mettumi ofleak I
e i
Ldesecton, and protecove fear es.
l See the ' response for RAI 440132 fw a discussion of tius issue i
DISCUSSED AT 1/25/95 MEETING BETWEEN WESTINGilOUSE AND NRC PLANT SYSTEMS ERANCil l
NRC to renew rat 440132,210.61. Sectum 5 4.7. and Section I.9.5
- ...=
Ckned - W-Pw has completed neassary subnustals to support staff review.
! Action N - NRC to review RAI 440132. 210 61, Section 5.4.7, and Secton 1.9.5.
i l Action W - per 12/2 eclecon, Westeghouse to provide emphcat references to where we covered the CVS ptwinm of ISLOCA in the SSAR or (ther document.
l
[A_cuan Njg to @_ffnun on 1/9/97 provided emphcit referen s.
i
.b u
i.
l I
t L
Page-1 Total Records: I
o,
- f i
FAX to BILL HUFFMAN January 9,1997 CC:
Don Lindgren Mike Corletti Brian McIntyre ADDITIONAL INFORMATION FOR OPEN ITEM 157 This is in response to the 12/2/96 request to provide, explicitly, where we covered the CVS pnrtion of ISLOCA in the SSAR or other document. We explicitly cover the CVS por ion of ISLOCA in WCAPel4425, the ISLOCA report.' This WCAP is referenced in the SSAR in a number of places.
The most relevant to the CVS potion of ISLOCA are in section B-63 of SSAR section 1.9, and in subsection 1.9.5.1.7. We believe this completes Westinghouse actions required for Open item 157 and request NRC direction to change its "NRC Status".We -recommend " Action N" Jim Winters 412-374-5290 9
4 9
343 4
e
_.s,._.
(
h Westinghouse FAX COVER SHEET W
i RECIPIENT INFORMATION SENDER INFORMATION i
1 l
OATE:
3 9 7 NAME:
('f mL 6 c, i
TO:
LOCATION:
ENERdY CENhR -
)
L.'Axr*y/6.11Hdi,vem EAST l
PHONE:
FACSIMILE:
PHONE:
Office: 412 -3 74 -C7 7 COMPANY:
Facsimile:
win:
284 4887 I
U$ l\\3RC -
outside: (412)374-4887 LOCATION:
4 l
Cww+Pgm 1+d The following pages are being sent from the Westinghouse Energy Center, East Tower, Monroeville, PA. If any problems occur during this transmission, please call:
4 WIN: 284 5125 (Janice) or Outside: (412)374 5125.
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COMMENTS:
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- 6. Ergineered Saf1ty Features Although gas accumulation is not expected. there is a vertical pipe stub on the top of the inlet piping high point that serves as a gas collection chamber. Level detectors indicate when gases have collected in this area. There are provisions to allow the operators to open manual valves to locally vent these gases to the in-containment refueling water storage tank.
l The passive residual heat removal heat exchanger, in conjunction with the passive containment cooling system can provide core cooling for an indefinite period of time. After the in-I containment refueling water storage tank water reaches its saturation temperature (in about I
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), the process of steaming to the containment initiates.
Condensation occurs on the steel containment vessel, which is cooled by the passive containment cooling system. The condensate is collected in a nonsafety-related gutter I
anangement located at the operating deck level which retums the condensate to the in-containment refueling water storage tank. The gutter normally drains to the containment sump, but when the passive residual heat removal heat exchanger actuates, a nonsafety-related isolation valve in each of the two gutter drain lines shut and the gutter overflow retums oirectly to the in-containment refueling. water storage tank.
Recovery of the condensate maintains the passive residual heat removal heat exchanger heat I
sink for a very long time. Without recovery of the condensate, the in-containment refueling water storage tank inventory is sufficient to provide passive residual heat removal heat exchanger operation for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
ne passive residual heat removal heat exchanger is also used to maintain a safe shutdown
~
l condition. It removes decay heat and sensible heat from the reactor coolant system to the in-containment refueling water storage tank, the containment atmosphere, the containment vessel, and finally to the ultimate heat sirk - the atmosphere outside of containment. This occurs after in-contair, ment refueling water storage tank saturation is reached and steaming to containment initiates.
6.3.2.1.2 Reactor Coolant System Emergency Makeup and Boration he core makeup tanks provide reactor coolant system makeup and boration during events not i
involving loss of coolant when the normal makeup system is unavailable or insufficient.
here are two core makeup tanks located inside the containment at an elevation slightly above the reactor coolant loops. Durmg normal operation, the core makeup tanks are completely full of cold, borated water. De boration capability of these tanks provides adequate core shutdown margin following a steam line break.
The core makeup tanks are connected to the reactor coolant system through a discharge injection line and an inlet pressure balance line connected to a cold leg. De discharge line is blocked by two normally clored, parallel air operated isolation valves that open on a loss of air pressure or electrical power, or on control signal actuation. L v<t(S K m..
c.-
Revision: 5 February 29,1996 y Westinghouse
/
- 6. E gineered Safity Features
~
The gravity injection line flow paths from the in-containment refueling water storage tank The contam.nent recirculation lines that connect to the gravity injection lines The check valves selected for these applications incorporate a simple swing-check design with a stainless steel body and hardened valve seats. The passive core cooling system check valves are safety-related, designed with their operating parts contained within the body, and with a f
I low pressure drop across each valve. The valve intemals are exposed to low temperature reactor coolant or borated refueling water.
During normal plant operation, these check valves are closed, with essentially no differential pressure across them. Confidence in the check valve operability is provided by operation at I
no differential pressure clean / cold fluid environment, the simple valve design, and the specified seat materials.
The check valves normally remain closed, except for testing or when called upon to open following an event to initiate passive core cooling system operation. The valves are not subject to the degradation from ficw operation or impact loads caused by sudden flow reversal and seating, and they do not experience significant wear of the moving parts.
I These check valves are periodically tested during shutdown conditions to demonstrate valve 1
operation. Rese check valves are equipped with nonintrusive position sensors to indicate when the valves are open or closed.
In current plants, there are many applications of simple swing-check valves that have similar operating conditions to those in the passive core cooling system. The extensive operational history and experience derived from similar check valves used in the safety injection systems of current pressurtzed water reactors indicate that the design is reliable. Check valve failure to open and common mode failures have not been significant problems.
I 6.3.2.2.7.7 Accumulator Check Valves ne accumulator check valve design is similar to the accumulator check valves in current pressurtzed water reactor applications. It is also similar to the low differential pressure I
openmg check valve design described in subsection 6.3.2.2.7.6.
Lwt (@) hm.
Q-i s
1 Durmg normal operation, the check valves are in the closed position with a nominal j
differential pressure across the disc of about 1550 psid. He valves remain in this position, j
except for testing or when called upon to open following an event. They are not subject to j
the degradation from flow operation or impact loads caused by sudden flow reversal and l
seating. ney do not experience significant wear of the moving parts and they are expected l
to function with minimal backleakage.
i i
Revision: 5 3 Westinghouse February 29,1996 6.3-21 I
l
h
- 6. Eagineered Saf;ty Festures In the incontainment refueling water storage tank injection lines, the squib valves are in series with normally closed check valves, in the containment recirculation lines, the squib valves l
are in series with normally closed check valves in two lines and with normally closed motor operated valves in the other two lines. As a result. inadvertent opening of these squib valves will not result in loss of reactor coolant or in draining of the incontainment refueling water i
storage tank.
1 The type of squib valve used in these applications provides zero leakage in both directions.
It also allows flow in both directions. A valve open position sensor is provided for these valves. Lwt @ %
W, Squib valves are also used to isolate the fourth stage automatic depressurization system lines.
l These squib valves are in series with normally open motor operated gate valves. Redundant-l
. series controllers are provided to prevent spuriously opening of these squib valves. The type of squib valve used in this application provides zero leakage of reactor coolant out of the I
reactor coolant system. The reactor coolant pressure acts to open the valve. A valve open position sensor is provided for these valves.
6.3.2.3 Applicable Codes and Classifications Sections 5.2 and 3.2 list the equipment ASME Code and seismic classification for the passive core cooling system. Most of the piping and components of the passive core cooling system within containment are AP600 Equipment Class A, B, or C and are designed to meet seismic Category I requirements. Some system piping and components that do not perform safety-related functions are nonsafety related.
I The requirements for the control, actuation, and Class IE devices are presented in Chapters 7 and 8.
6.3.2.4 Material Specifications and Compatibility Materials used for engineered safety feature components are given in Section 6.1. Materials for passive core cooling system components are selected to the meet the applicable material requirements of the codes in Section 5.2, as well as the following additional requirements:
Parts of components in contact with borated water are fabricated of, or clad with, austenitic stainless steel or an equivalent corrosion resistant material.
Internal parts of components in contact with containment emergency sump solution during recirculation are fabricated of austenitic stainless steel or an equivalent corrosion resistant material.
Valve seatmg surfaces are hard-faced to prevent failure and to reduce wear.
)
Revision: 5 yg February 29,1996 6.3-23
. A.
)
i 9
I
?
l INSERT A i
j j
The core makeup tank discharge isolation valves are diverse from the passive residual heat removal j
heat exchanger outlet isolation valves because they use different globe valve body styles and different j
air operator types.
1 i
1 i
INSERT B i
The accumulator check valves are diverse from the core makeup tank valves because they use different j
check valve types.
4 i
i 1
2 INSERT C l
The IRWST injection squib valves are diverse from the containment recirculation squib valves because 2
they are designed to different design pressures.
i 1
1 l
1:
l-i
0 i
1 4
4 FAX to DINO SCALETTI March 12,1997 CC:
Sharon or Dino, please make copies for:
Bill Huffman Ted Quay Robin Nydes Chip Suggs i
Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #173 (M5.2.5-30)
This item is similar to item #172 (yesterday's FAX). The relevant documentation related to Open Item #173 (MS.2.5-30) is attached. We provided the original comparison to STS with NSD-NRC-%-
4833 on October 11,1996. We then provided probability risk assessment information related to the differences from STS with NSD-NRC-97-4939 on January 14, 1997. This was reiterated in the RAI responses provided by NSD-NRC-97-4972 of February 6,1997. We then asked for a new NRC Status, with a package like this one, on February 17, 1997. This item (#173) was asked by a l
technical branch other than the Tech Spec branch. The letters identified above were in response to questions asked by the Tech Spec branch and provide general justification for Action Times. Included in the general justifications are specific entries for TS 3.4.5, the subject of this item #173. Please 4
help us provide the branch to branch coordination required to obtain proper review of this information. We believe that the letters identified above resolve the concerns of item #173. Since NRC should be responsible to review information submitted by Westinghouse, it seems a reasonable request thht NRC acknowledge receipt of the information related to Open Item #173. We request that NRC provide a defm' itive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N". Thank you.
Jim Winters 412 374-5290 e.
J 1
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AP600 Open Item Tracking System Database: Executive Summary -
Date: 3/12/97 Seledion:
[ item noj between 173 And 173 Sorted by Item C liem DSER Sectmnf Title / Description Resp (W)
NRC No Branch Questmn Type Detaal Staus Engineer Staus Status im No. f Due 173 NRR/SPLB 525 MTG4M TECHSPEC/Suggs, C.
Closed Actum W M5 2.5-30 (REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE) STS 3.4.15 includes SR 3.4.15.2, which stmes that a channel operational f 3test (COT) should be performed on the c.-
= sWaa radioactivity morutor every 31 days. APbOO TS 3 4 5 includes SR 3 4.9.2 mluch state thz the COT slumand be performed every 92 days. Westmghouse shouW provide ficarna for the deviation inwn STSs 1
Action W: pautfacatum of diffenences between AP600 TS and STS will be provided with TS. rka 3/29 Closed - With issuance of the Tech Specs in SSAR Rev. 9.
{ Action W-Need an explananon of Actam Tunes as they relase to STS.
Closed - Wuh issuance of letters NSIKNRC-96-4833 (10-11-%) which explans differences between STS and AP600 TS, NSD-NRC-97-4939 (I-14-97)
[which provides the response to R Al 630.10 for PRA support of TS, and NSD-NRC-97-4972 (24-97) wluch responds to RAls 630 II-I4 segarding the basis for completion times and surveil, lance _ frequencies. rka 2/_24N7 D
M 4
Page: 1 Total Records: I
i Westinghouse Energy Systems ha ass Sectric Corporation
"** P****a mmosse l
NSD-NRC 96-4833 DCP/NRC0616 j
Docket No.: STN 52 003 October 11, 1996 i
Document Control Desk
}
U.S. Nuclear Regulatory Cocimission j-Washington, D.C. 20555 l*
' ATTENTION:
T.R. QUAY i
SUBJECT:
CLOSING THE LAST DSER OPEN ITEM FOR AP600 SSAR SECTION j
16.1, TECHNICAL SPECIFICATIONS (TS)
Dear Mr. Quay:
i
)
This tener is written to close the last DSER open item for AP600 SSAR Section 16.1 Technical l
Specifications (TS). Westlagham comruitted to provide written explanation of technical differences l
between the AP600 TS and those presented in NUREG-1431, the Standard TS (STS). Attached are:
l 1.
A roadmap which identifies the sections comprising the STS versus those included in the i
AP600 TS. For any TS that are included in the STS but not in the AP600 TS, an explanation is provided. For any TS that are included in the AP600 TS but not in the STS, those sections are shaded in the roadmap and explained. Explanations are also provided for other content
- differences between the STS and AP600 TS.
.J 1
)
2.
A description of general or overall changes whose explanations apply to multiple TS.
1 3.
A list of technical differences between the STS and AP600 TS. De TS and BASES are i
grouped by section and an explanation of each difference is provided.
4.
A table of and expInnarian for those LCOs whose endpoint is defined as MODE 4 for the AP600, rather than MODE 5 or "Go to LCO 3.0.3' per the STS.
i l
Discussions regarding ties between the AP600 PRA and the Technical Specifications will be provided j
in the response to RA1630.10.
1 4
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=
(
4SD NRC.96-4833 DCP/NRC0616 2
October 11, 1996
\\
l "this submittal closes Open Item Tracking System (OITS) hem 2353, which la the final open item for the AP600 Technical Specifications. If you have any questions regarding this transmittal, please contact Robin K. Nydes at (412) 374 4125.
A Brian A. McIntyre, ger j
Advanced Plant Safety and Licensing Ada Anachment i
cc:
W. Humnan, NRC I
A. Chu, NRC C. Grimes, NRC l
N. Liparulo, Westinghouse (w/o Attachments) 1 i
2
_ ~ ~ ~
4 I
l l
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Westinghouse Energy Systems smass sectric Corporation renwe % istmosas NSD-NRC-97-4939 DCP/NRC0705 l
Docket No.: STN 52-003 4
January 14, 1997 s
l Document Control Desk U. S. Nuclear Regulatory Commission j
Washington, DC 20555 i
ATTENTION:
T.R. QUAY
SUBJECT:
WESTINGHOUSE RESPONSE TO RAI 630.10
Dear Mr. Quay:
Enclosed we three copies of the Westinghouse response to RA1630.10 regarding AP600 Technical l
Specification deviations from NUREG-1431 based on probability risk assessment. 'the NRC technical staff should review this response as part of their review of the AP600 Technical Specifications. This i
closes DSER open item tracking system item #3054. If there are any questions regarding this i
transmittal, please contact Robin K. Nydes at (412) 3744125.
h/
Brian A. McIntyre, Manager
~
Advanced Plant Safety and Licensing Sml
)
enclosure cc:
Angela Chu, NRC -(w/ enclosure)
W. C. Huffman, NRC -(w/ enclosure)
Nicholas Liparulo, Weia: ham - (w/o enclosure) f9 o
e e
"o NRC REQUEST F@R ADDITIONAL INFORMATION g
4 Question 630.10.
Provide a list of proposed AP600 Technical Specification requirements that deviate i
from NUREG 1431 based either totally or partially on probabilistic risk assessment (PRA) or PRA insights.
I p.-a-w Tbc deviations from NUREG 1431 are explained in Reference 1. There are no AP600 Technical Specifications which deviate from NUREG-1431 with the PRA as the basis, i
However, selection of a standardized Completion Time or Surveillance Frequency ena tA.~
" M. PRA results as described in Reference 2. Per NRC request, j
amis a list com g the NUREG 1431 Standardized Technical Specification (STS) completion tim d surveillance frequencies to the AP600 TSs. Deviations
- f... E n F_".2
.on are less restrictive than STS times are highlighted and any PRA
]
relationshi ts given la the comment column.
ggfn U
^
SSAR Revision: NONE
References:
1.
NSD NRC 96-4833, Closing the Last DSER Open Item for AP600 SSAR Section 16.1, Technical Specifications (TS), 10/11/96.
2.
NSD-NRC-96-4699, Westinghouse AP600 Technical Specifications Approach,5/3/96.
630.10 1 T W850ngh0008 d1 8
s a
weninghouse enere snten.
i Bectric Corporation Pmawehmmma mwom j
NSD NRC 97 4972 j
DCP/NRC0732 4
Docket No.: STN 52 003 l
February 6,1997 i
Document Control Desk l
U. S. Nuclear Regulatory Commission Washington, DC 20555
{
TO:
T.R. QUAY
SUBJECT:
RESPONSE TO RAIs 630.11 THROUGH 63Q.14 i
REFERENCE:
LETTER FROM NRC TO WESTINGHOUSE (HUFFMAN TO LtPARULO),
l
" REQUEST FOR ADDITIONAL INFORMATION ON WESTINGHOUSE AP600 i
TECHNICAL SPECIFICATIONS OPTIMlZATION METHODOLOGY", DATED j
DECEMBER 12, 1996.
i Enclosed for NRC review are the Westinghouse responses to the following Technical Specification j
RAls, provided by the above Reference, i.
4 1
630.11 Completion Time Anchor Point i
630.12 Surveillance Frequency Baseline 630.13 Request for Response to RAI 630.10 l
630.14 Differences Between the Proposed Tech Specs Approach and Tech Specs Rev. 2 i
i This completes WeiWe activity for Open item Tracking System item 4224 through 4227, a report for which is attached. Please advise as to the NRC status for l'oese items, if you have any 2
questions regarding this transmittal, please contact Robin K. Nydes (412) 374-4125.
f i
Brian A. McIntyre, Manager Advanced Plant Safety and Licensing det enclosure anachment ec:
W. Hufhnan, NRC (w/ enclosure /anachment)
A. Chu, NRC (w/ enclosure /anachn=m) q e
e
je, 4
FAX to DINO SCALETTI March 12,1997 CC:
Sharon or Dino, please make copies for:
Ted Quay Bill Iluffman Diane Jackson Tom Kenyon Joe Sebrosky Cindy Haag Don Lindgren Robin Nydes Brian McIntyre Ed Cummins Bob Vijuk This is a reminder list of the Open Items where we have recently provided background documentation showing the difference between "W Status" and "NRC Status". In all cases, we believe the next action is with NRC and await your definitization of a Westinghouse action or your direction to change the "NRC Status" to something other than " Action W" Note that we have received no information from NRC on items on this list for over a week. Note that submittal dates over a year old and s
request dates over a month old are in bold type.
Open item Number Westinghouse Submittal Request for Status Change 1
142 (M3.Il-9) 2/29/%
2/3/97 3/4/97 157 (MS.2.5-13) 1/9/97 2/12/97 3/6/97 164 (M5.2.5-20) 1/10/97 2/12/97 3/11/97 172 (MS.2.5-29) 1/14/97 2/14/97 3/11/97 173 (M5.2.5-30) 1/14/97 2/17/97 3/12/97 182 (MS.4.11-5) 1/10/97 2/20/97 184 (MS.4.11-7) 1/13/97 2/20/97 405 7/8/%
2/11/97 556 (DSER 2.5.4.8-1) 12/20/96 3/10/97 681 (DSER 3.8.2.4-3) 2/11/97 2/17/97 7% (DSER 3.8.2.4-28) 2/11/97 2/17/97 1 of 5 I
'9 Open Item Number Westinghouse Submittal Request for Status Change i
710 (DSER 3 8.3.1-1) 1/16/97 2/18/97 716 (DSER 3.8.3.2-5) 1/16/97 2/18/97 717 (DSER 3.8.3.3-1) 1/16/97 2/18/97 718 (DSER 3.8.3.3-2) 1/16/97 2/18/97 722 (DSER 3.8.3.4-3) 1/16/97 2/18/97 724 (DSER 3.8.3.4-5) 1/16/97 2/18/97 4
1 729 (DSER 3.8.3.4-10) 1/16/97 2/18/97 731 (DSER 3.8.3.4-12) 1/16/97 2/18/97 l
740 (DSER 3.8.4.1-3) 1/16/97 2/20/97 754 (DSER 3.8.4.4-6) 1/16/97 2/20/97 2
757 (DSER 3.8.4.5-1) 1/16/97 2/20/97 758 (DSER 3.8.4.5-2) 1/16/97 2/20/97 782 (DSER 3.9.2.3-1) 6/30/95 2/28/97 783 (DSER 3.9.2.3-2) 2/19/97 2/28/97 786 (DSER 3.9.3.1-1) 2/29/%
2/28/97 l
793 (DSER 3.9.3.3-2) 2/19/97 2/28/97 l
801 (DSER 3.9.6.2-4) 2/19/97 2/28/97 i
802 (DSER 3.9.6.2-5) 2/19/97 2/28/97 1
805 (DSER 3.9.6.2-8) 2/19/97 2/28i97 d
807 (DSER 3.9.6.3-1) 2/19/97 2/28/97 809 (DSER 3.9.6.4-2) 2/19/97 2/28/97 854 (DSER 4.2.8-1) 4/12/96 3/4/97 1
1172 (DSER 11.2-6) 2/21/97 3/4/97 i
1210 (DSER 12.4.2-2) 4/30/%
2/6/97
)
i 1227 7/8/96 2/11/97 1228 7/8/%
2/11/97 1231 7/8/96 2/11/97 1232 7/8/%
2/11/97 1716 12/17/96 1/28/97 4
2 of 5 1
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Open Item Number Westinghouse Submittal Request for Status Change 1727 12/17/96 2/28/97 1730 2/19/97 2/28/97 1731 2/19/97 2/28/97 1736 2/19/97 2/28/97 1740 10/11/96 2/28/97 j
1742 12/17/96 2/28/97 1745 12/17/96 2/28/97 1747 12/17/ %
2/28/97 1753 12/17/96 2/28/97 1760 12/17/96 2/28/97 1792 (DSER-CN 3.9.2.1-4) 10/23/96 2/28/97 i
1793 (DSER-CN 3.9.2.3-1) 10/23/96 2/28/97 1797 (DSER-CN 3.9.2.4-4) 10/14/96 2/28/97 1802 (DSER-CN 3.9.3.3-3) 9/5/96 2/28/97 1803 (DSER-CN 3.9.3.3-4) 9/5/96 2/28/97 l
1807 (DSER-CN 3.9.7-1) 6/19/96 2/28/97 I888 (DSER-COL 3.8.2.4-1) 2/11/97 2/17/97 j
2034 (DSER-015013.)
7/8/96 2/11/97 2066 12/17/96 2/28/97 2347 1/16/97 2/18/97
~
2348 1/16/97 2/18/97 2349 1/16/97 2/18/97 3057 5/30/%
2/18/97 3144 (DSER 21.6.2.4-3) 10/18/ %
3/11/97 i
3147 (DSER 21.6.2.4-6) 10/25/ %
3/11/97 3148 (DSER 21.6.2.4-7) 10/18/ %
3/11/97 3149 (DSER 21.6.2.4-8) 10/18/ %
3/11/97 l
3150 (DSER 21.6.2.4-9) 10/18/96 3/11/97 3157 (DSER 21.6.2.5-3) 10/25/96 3/11/97 3 of 5 l
l I
o, Open Item Number Westinghouse Submittal Request for Status Change 3158 (DSER 21.6.2.6-1) 10/18/96 3/11/97 j
3159 (DSER 21.6.2.6-2) 10/25/96 3/11/97 3247 (RAI 230.98) 4/30/96 2/18/97 3372 (RAI 210.213) 1/8/97 2/28/97 4123 (RAI 480.440) 12/13/96 3/10/97 4124 (RAI 480.441) 12/13/96 3/10/97 4125 (RAI 480.442) 12/13/0^
3/10/97 l
4126 (RAI 480.443) 12/13/96 3/10/97 4127 (RAI 480.444) 12/13/96 3/10/97 4128 (RAI 480.445) 12/13/ %
3/10/97 j
4129 (RAI 480.446) 12/13/ %
3/10/97 l
4130 (RAI 480.447) 12/13/ %
3/10/97 4131 (RAI 480.448) 12/13/96 3/10/97 4132 (RAI 480.449) 12/13/96 3/10/97 4133 (RAI 480.450) 12/13/96 3/10/97 4134 (RAI 480.451) 12/13/ %
3/10/97 4135 (RAI 480.452) 12/13/96 3/10/97 4136 (RAI 480.453) 12/13/96 3/10/97 4137 (RAI 480.454) 12/13/ %
3/10/97 4138 (RAI 480.455) 12/13/96 3/10/97 4139 (RAI 480.456) 12/13/ %
3/10/97 4140 (RAI 480.457) 12/13/ %
3/10/97 4141 (RAI 480.458) 12/13/ %
3/10/97 4142 (RAI 480.459) 12/13/96 3/10/97 4143 (RAI 480.460) 12/13/96 3/10/97 4998 2/19/97 2/28/97 4999 2/19/97 2/28/97 5001 2/19/97 2/28/97 5002 2/19/97 2/28/97 4 of 5
ee Open item Number Westinghouse Submittal Request for Status Change Note that the_ status was changed for a large number of items so they have been removed from the table.
Thanks for your help.
O Jim Winters I
l 5 of 5
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,5 RECIPIENT INFORMATION SENDER INFORMATION DATE:
mn <u,-i /3 /997 NAME:
j,1 Q,,
TO:
LOCATION:
ENERGY CENTER -
%m l(an/ced EAST PHONE:
FACSIMILE:
PHONE:
Omce: W2-3 'N-rz 9 o COMPANY:
Facsimile:
Win:
284 4887 WR outside: (412)374 4887 LOCATION:
J s
Cover + Pages l[
4 The following pages are being sent from the Westinghouse Energy Center, East Tower,
)
Mortroeville, PA. If any problems occur during this transmission, please call:
WIN: 284 5125 (Janice) or Outside: (412)374 5125.
COMMENTS:
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6
t' NRC REQUEST FOR ADDITIONAL INFORMATION
- =
Question 471.23 Revision 1 Verify that the airborne radiation monitors described in Section 11.5.2.3.2 of Chapter 11 of the SSAR will be sensitive enough to detect 10 DAC-hrs in any area of the plant that can be accessed by plant personnel.
Response
Five (5) airborne radiation monitors are described in Subsection 11.5.2.3.2. An additional six (6) monitors for areas than can be accessed by plant personnel are described in Subsection 11.5.2.3.1. These radiation monitors are part of the permanently installed AP600 radiation monitoring system and provide ;;eneral area monitoring.
These radiation monitors are supplemented by local portable continuous air monitors (CAMS). CAM use is directed by the Health Physics staff during maintenance operations with a high potential for airborne radioactivity levels.
The eleven (!!) radiation monitors mentioned above monitor selected areas of the plant that can be accessed by plant personnel. These selected areas are as follows:
- 1) Fuel Handling Area
- 2) Auxiliary Building
- 3) Annex Building
- 4) Main Control Room and Technical Suppon Center
- 5) Containment
- 6) Health Physics and Hot Machine Shop
- 7) Radwaste Building Areas 1,2,3,6, and 7 are monitored by measuring the concentration of radioactive materials in the exhaust air from each area.
Area 4 is monitored by measuring the concentration of radioactive materials in the supply air.
Area 5 is monitored by three separate airborne process monitors:
- 1) The Containment Air Filtration Exhaust Radiation Monitor measures the concentration of radioactive materials in the containment purge exhaust air.
- 2) The Containment Atmosphere Radiation Radiogas Monitor measures the radia' ion from the radioactive gases in the containment atmosphere.
- 3) The Containment Atmosphere Radiation N"/F Monitor measures the concentration of radioactive airborne gaseous contamination inside the containment as an indication of reactor coolant pressure boundary leakage.
471.23-1 3 Westingflouse Rev.I
r NRC REQUEST FOR ADDITIONAL INFORMATION These eleven (11) monitors are sensitive enough to detect 10 DAC-hours as discussed below.
SSAR Table 11.5-1 provides a listing of each detector, the principal isotope (s) it monitors, and the detector's nominal range. The lower value of the detector's nominal range corresponds to the detector's minimum detectable level. These minimum detectable levels are achieved with a 95% confidence level at standard operating conditions. The following table summarizes Table 11.5-1 and includes the DAC occupational values from Table 1. Column 3, of Appendix B (Annual Limits on intake (Alls) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage) of 10 CFR 20.
DAC Occupational Detector Minimum Values - 10 CFR 20, Airborne Process Radiation Monitor Isotope (s)
Detectable Level Appendix B. Table 1 Fuel Handling Area Exhaust Kr-85 1.0E-6 pCi/cc 1.0E-4 pCi/cc Xe-133 1.0E-6 pCi/cc 1.0E-4 pCi/cc Auxiliary Building Exhaust Kr-85 1.0E-6 pCi/cc 1.DE-4 pCi/cc l
Xe-133 1.0E-6 pCi/cc 1.0E-4 pCi/cc Annex Building Exhaust Kr-85 1.0E-6 pCi/cc 1.0E-4 pCi/cc Xe-133 1.0E-6 pCi/cc 1.0E-4 pCi/cc MCR Supply Air Duct Particulate Sr-90 1.0E-12 pCi/cc 8.0E-9 pCi/cc Cs-137 1.0E-12 pCi/cc 6.0E-8 pCi/cc MCR Supply Air Duct Iodine I-131 1.0E-ll pCi/cc 2.0E-8 pCi/cc MCR Supply Air Duct Gas Kr-85 1.0E-7 pCi/cc 1.0E-4 pCi/cc Xe-133 1.0E-7 pCi/cc 1.0E-4 pCi/cc Containment Air Filtration Exhaust Kr-85 1.0E-6 pCi/cc 1.0E-4 pCi/cc Xe-133 1.0E-6 pCi/cc 1.0E-4 pCi/cc Health Physics and Hot Machine Shop Exhaust Sr-90 1.0E-13 pCi/cc 8.0E-9 pCi/cc Cs-137 1.0E-13 pCi/cc 6.0E 8 pCi/cc Radwaste Building Exhaust St-90 1.0E-13 pCi/cc 8.0E-9 pCi/cc Cs-137 1.0E-13 pCi/cc 6.0E-8 pCi/cc Containment Atmosphere N"/F" N-13 1.0E-7 pCi/cc N/A F-18 1.0E-7 pCi/cc 3.0E-5 pCi/cc Containment Atmosphere Gas Kr-85 1.0E-7 pCi/cc 1.0E-4 pCi/cc Xe-133 1.0E-7 pCi/cc 1.0E-4 pCi/cc 471.23-2 Rev.1
s NRC REQUEST FOR ADDITIONAL INFORMATION N
The above table shows that for each principal isotope, the minimum detectable level for each monitors detector (s) is almost two (2) to almost five (5) orders of magnitude below the corresponding 10 CFR 20 DAC occupational value.
These radiation monitors utilize two basic types of detectors, as described in Section 11.5.2.3.2. The particulate (Sr-90/Cs-137) and iodine detectors use shielded fixed filters, located in the sample stream, that are viewed by beta and gamma sensitive scintillators, respectively. The radiogas detectors use beta sensitive scintillators with their sensitive volumes directly exposed to the process or sample stream.
The response time for each fixed filter detector depends upon background radiation levels, airborne radioactivity levels, sample flow rate, and system configuration. When the detectors have achieved statistically accurate operating conditions, the detector response times are as follows:
- 1) Step change in radioactivity levels above the ALERT setpoint - < 4 seconds, not including sample transport time.
- 2) Gradually increasing radioactivity levels above the ALERT setpoint - < 2 seconds, not including sample transport time.
The step change requires a longer response time to assure that the change is not a spurious radioactivity spike.
The time to achieve statistical accuracy (95% confidence level) can vary from ten minutes to one hour, depending upon radioactivity concentrations. The only time the detectors will not be operating under statistically accurate conditioiis will be the time following a filter change or a system shutdown for maintenance.
Sample transport times are minimized by locating the detectors as close as practicable to the process sample point.
The response time for the in-line detectors is less than ten seconds. These detectors are prcvided with dynamic background radiation compensation.
Combining the minimum detectable levels shown in the table above with the detector response tirr.es discussed above, it has been shown that each monitor is sensitive enough to detect 10 DAC-hours.
SSAR Revision:
i in Table 11.5-1, " Radiation Monitor Detector Parameters", add the following:
Add "(Note 5)" in the " Service column for:
Containment Atmosphere Gas Containment Atmosphere N"/F" Fuel Handling Area Exhaust Auxiliary Building Exhaust 471.23 3 3 Westingflouse pay, j
1 1
5 l
NRC REQUEST FOR ADDITIONAL INFORMATION ini
~~
i.
Annex Building Exhaust Main Control Room Supply Air Duct (Particulate) - both entries MCR Supply Air Duct (lodine) - both entries MCR Supply Air Duct (Gas) - both entries Containment Air Filtration Exhaust H.P. & Hot Machine Shop Exhaust Radwaste Building Exhaust In " Notes:" add:
"S. Monitor is sensitive enough to detect 10 Derived Air Concentration (DAC)-hours."
471.23-4 W westinghouse Rev.1
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- Winters, James i
From:
Dino Scaletti To:
Winters, James
Subject:
AP600 OITS Date:
Wednesday, February 26,1997 3:38PM
- Jim, i
Some of the following may have been already addressed as action NRC.
Move items 1809,1810 and 1811 to " Action N" Move items 3264, 3265, 3266, 3267, and 3268 to " Action N" Move items 3269, 3270, and 3271 to " Action N" Move the 15 Items referenced in your 2/14/97 fax to " Action N" Move #5 (RAI 410.263) to " Action N" Move 1883,2430,2431,2432, and 3518 to " Action N" i
Dino 4
i e
d I
Page 5
7 FAX to DINO SCALETTI March 14,1997 CC:
Sharon or Dino, please make copies for:
Bill Huffman Ted Quay Don Lindgren Ron Vijuk Terry Schulz Mike Corletti Ed Cummins Bob Vijuk Brian McIntyre OPEN ITEM #182 (MS.4.11-5)
In my quest to make sure we have provided NRC with everything needed to prepare an FSER, I am researching open items from the smallest item number on. The relevant documentation related to Open item #182 (M5.4.11-5) is attached. We provided the original fax of a markup on January 10, 1997 (over two months ago). We requested an NRC Status change ad resent the markup on February 20,1997 (almost a month ago). We included the changes indicated on the January fax in Revision 11 of the SSAR on February 28,1997 (two weeks ago). Although we understand that NRC may need to review the IRWST design for ADS actuation, we believe that the information provided is sufficient to resolve the concerns of item #182. It seems a reasonable request that NRC acknowledge receipt of the information. We request that NRC provide a definitive action for Westinghouse or provide direction to change the status of this item. We recommend " Action N". Thank you.
De Jim Winters 412-374-5290 k
4 l
AP600 Open hem Tracking System Database: Executive Summary page: 3/14/97 Selection:
litem no] between 182 And 182 Sorted by item #
liem '-
DSER Secnont Title /Descnpaa Resp (W)
NRC No Hranch Questam Type Detail Stasus Engineer Status Stasus Lxtier No. /
Date 182 NRR/Si1H 54l)
MTG 01 Corletti.M Closed Action W MS.4 Il-5 (PRESSURIZER RELIEF DISCilARGE) Secnon 5 4. Ilistates that the IRWST is sized based on the heat kud and sacant volume folloMag I an actuation of the ADS. Does this mclude stemn, water, and -he gases from all three ADS stages Provide the analysis.
?
Closed - See Secuan 6 3 for a discussion of the IRWST during actuation of the automatic depressunzmon system DISCUSSED AT'I'/25S5 MEETING -
BETWEEN WESTINGIM)USE AND NRC f1 ANT SYSTEMS BRANCll I
Clowd - NRC to review follommg Westmghouse ;provahng of specific SSAR referenw Specific reference provided by fax markup of SSAR on January !
10.1997 Speciric reference included in SSAR Revision iI of 2/28N'. jww N
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Page. l Total Records: I
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- 5. Reactor Coolant System cnd Connected Systems 5.4.10.3 Design Evaluation An evaluation verifies the design adequacy and structural integrity of the reactor coolant loop and the primary equipment supports system. This evaluation compares the analytical results with established criteria for acceptability. Structural analyses demonstrate design adequacy for safety and reliability of the plant in case of a seismic disturbance, and/or loss of coolant accident conditions. Loads that the system is expected to encounter during its lifetime (thermal, weight, and pressure) are applied, and stresses are compared to allowable values.
Subsection 3.9.3 discusses the modeling and analysis methods.
5.4.10.4 Tests and Inspections Nondestructive examinations are performed according to the procedures of the ASME Code,Section V, except as modified by the ASME Code,Section III, Subsection NF.
5.4.11 Pressurizer Relief Discharge The AP600 does not have a pressurizer relief discharge system. The AP600 has neither power operated pressurizer relief valves nor a pressurizer relief discharge tank. Some of the functions provided by the pressurizer relief discharge system in previous nuclear power plants are provided by portions of other systems in the AP600.
The safety valves connected to the top of the pressurizer provide for overpressure protection of the reactor coolant system. First, second, and third-stage automatic depressurization system valves provide for depressurization of the reactor coolant system and venting of noncondensable gases in the pressurizer following an accident. These functions are discussed in subsections 5.2.2,5.4.12, and in Section 6.3. He AP600 does not have power operated relief valves connected to the pressurizer.
De discharge of the safety valves is directed through a rupture disk to containment atmosphere.
De discharge of the first, second, and third-stage automatic depressurization system valves is directed to the in-containment refueling water storage tank.
For the automatic depressurization system valves, the following discussion considers culy the gas venting function. Only the first stage automatic depressurization valves are ased to vent non-condensible gases following an accident. The sizing considerations and design basis for the in-containment refueling water storage tank for the depressurization function are discussed I
throughout Section 6.3. The provisions to minimize the differential pressure between the containment atmosphere and the interior of the in-containment refueling water storage tank I
are also discussed in subsection 6.3.2.
He safety valve on the normal residual heat removal system, which provides low temperature overpressure protection, discharges into the in-containment refueling water storage tank. See subsection 5.4.7 for a discussion of the connections to and location of the safety valve la the normal residual heat removal system.
Revision: 11 February 28,1997 5.4-62 T Westinghouse N
p l
- 5. React:r Coolant System cnd Ccnnected Systems 5.4.11.1 Design Bases The containment has the capability to absorb the pressure increase and heat load resulting from the discharge of the safety valves to containment atmosphere. The in-containment refueling water storage tank has the capability to absorb the pressure increase and heat load from the discharge, including the water seal, steam and gases, from a first-stage automatic depressurization system valve when used to vent noncondensable gases from the pressurizer following an accident. The venting of noncondensable gases from the pressurizer following an accident is not a safety-related function.
5.4.11.2
System Description
Each safety valve discharge is directed to a rupture disk at the end of the discharge piping.
A small pipe is connected to the discharge piping to drain away condensed steam leaking past the safety valve. The discharge is directed away from r.ny safety related equipment, structures, or supports that could be damaged te the extent that emergency plant shutdown is prevented by such a discharge.
The discharge from each of two groups of automatic depressurization system valves is connected to a separate sparger below the water level in the in-containment refueling water storage tank. The piping and instrumentation diagram for the connection between the autorruttic depressurization system valves and the in-containment refueling water storag,e tank is shown in Figure 6.3-1. The in-containment refueling water storage tank is a stainless steel lined compartment integrated into the containment interior structure. The discharge of water, steam, and gases from the first-stage automatic depressurization system valves when used to vent noncondensable gases does not result in pressure in excess of the in-containment refueling water storage tank design pressure. Additionally, vents on the top of the tank protect I
the tank from overpressure, as described in subsection 6.3.2.
Overflow provisions prevent overfilling of the tank. The overflow is directed into the refueling cavity. The in-containment refueling water storage tank does not have a cover gas and does not require a connection to the waste gas processing system. The normal residual heat removal system provides nonsafety-related cooling of the in-containment refueling water storage tank.
)
l 5.4.11.3 Safety Evaluation The design of the control for the reactor coolant system and the volume of the pressurizer is such that a discharge from the safety valves is not expected. The containment design i
pressure, which is based on loss of coolant accident considerations, is greatly in excess of the I
pressure that would result from the discharge of a pressurizer safety valve. The heat load resulting from a discharge of a pressurizer safety valve is considerably less than the capacity of the passive containment ecoling system or the fan coolers. See Section 6.2.
Venting of noncondensable gases, including entrained steam and water from the loop seals in l
the lines to the automatic depressurizations system valves, from the pressurizer into spargers l
1 Revision: 11 3 W6Stilighouse 5.4-63 February 28,1997 a
h j
- 5. Rerctor Coolant System and Cennected Systems
'j below the water line in the in-containment refueling water storage tank does not result in a significant increase in the pressure or water temperature. The in-containment refueling water storage tank is not susceptible to vacuum conditions resulting from the cooling of hot water i
l in the tank, as described in subsection 6.3.2. The in-containment refueling water storage tank has capacity in excess of that required for venting of noncondensable gases from the pressurizer following an accident.
1 5.4.11.4 Instrumentation Requirements The instrumentation for the safety valve discharge pipe, containment, and in-containment I
refueling water storage tank are discussed in subsections 5.2.5,5.4.9, and in Sections 6.2 and 6.3, respectively. Separate instrumentation for the monitoring of 'the discharge of noncondensable gases in not required.
5.4.11.5 Inspection and Testing Requirements i
Sections 6.2 and 6.3 discuss the requirements for inspection and testing of the containment and in-containment refueling water storage tank, including operational testing of the spargers.
Separate testing is not required for the noncondensable gas venting function.
5.4.12 Reactor Coolant System High Point Vents The requirements for high point vents are provided for the AP600 by the reactor vessel head vent valves and the automatic depressurization system valves. He primary function of the reactor vessel head vent is for use during plant stanup to properly fill the teactor coolant system and vessel head.
Both reactor vessel head vent valves and the automatic depressurization system valves may be activated and controlled from the main control room.
j The AP600 does not require use of a reactor vessel head vent to provide safety-related core cooling following a postulated accident.
The reactor vessel head vent valves (Figure 5.4-8) can remove noncondensable gases or steam from the reactor vessel head to mitigate a possible condition of inadeg ae core cooling or impaired natural circulation through the steam generators resulting from the accumulation of noncondensable gases in the reactor coolant system. The design of the reactor vessel head vent system is in accordance with the requirements of 10 CFR 50.34 (f)(2)(vi).
The first stage valves of the automatic depressurization system are attached to the pressurizer and provide the capability of removing noncondensable gases from the pressurizer steam space following an accident. Venting of noncondensable gases from the pressurizer steam space is not required to provide safety-related core cooling following a postulated accident. Gas accumulations are removed by remote manual operation of the first stage automatic depressurization system valves.
He discharge of the automatic depressurization system valves is directed to the in-containment refueling water storage tank. Subsection 5.4.6 and Section 6.3 discuss the automatic depressurization system valves and discharge system.
Revision: 11 February 28,1997 5A-64 3 Westinghouse 5
f