ML20236S442

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Package Consisting of Requests for Addl Info Re Gpu Topical Repts TR-033, Methods for Generation of Core Kinetics Data for Retran-02 & TR-040, Steady-State & Quasi-Steady-State Methods Used in Analysis.... SPDS Meeting Summary Encl
ML20236S442
Person / Time
Site: Millstone, Oyster Creek, 05000000
Issue date: 11/20/1987
From:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
To:
Shared Package
ML20236S438 List:
References
FOIA-87-725 NUDOCS 8711250165
Download: ML20236S442 (12)
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gnclosure'l REQUEST FOR ADDITIONAL INFORMAT10N REGARDING GPU NUCLEAR TOPICAL REPORT TR-033

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" METHODS FOR THE GENERATI0M 0F CORE KINETICS DATA FOR RETRAN-02" ' f/

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Please provide a list of accidents and transients GPU Nuclear intends to'Ldb

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analyze using RETRAN-02 1-D methods and RETRAN-02 0-0 methods. Justd t'y the application of the appropriate methodology to each class of ascidents 7'

and transients chosen.

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' 2.. Provide estimates of uncertainties in physics parameters such as the Dop-pier reactivity coef ficient, the void reactivity coefficient, the scch,n reactivity, the effective delayed neutron fraction, the prompt nutron 9eneration time, and the collapsed cross-section parameters listed in V;

4 Table 3-1.

For each parameter., or class of parameters, cite refer 6 ces q provide analyses to justify the estimated uncertainties. Briefly describh how the RETRAN-02 analyses will be performed to conservatively account for the uncertainties.

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Has the computer code BELLER0 PHON been reviewed and accepted for licensing applications? Equation (2-14) assumes a uniform fission cross-section across the reactor core. Estimate the error introduced in Sj through this assumption and explain how it is accounted for. Explain how the in-portance factor is determined.

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4 Estimate the errors introduced by using a temperature independent Doppler

,j coefficient in 0-D applications of RETRAN-02, and negl(cting the moderator

,,e Gl6 temperature dependence of cross-sections in 1-D applications,

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Equation (2-21) has a typographical error. Please provide the correct equation.

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What is the value of radial buckling introduced into a partially con-trolled axial node in the radial leakage correction procedure?

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Table 4-1 and Figures 4-6 through 4-19 provide 3-D/1-D comparisons of-eigenvalues and axial power shapes for a set of dependent cases (control fraction varied while other independent variables held constant). Provide similar comparisons for a set of perturbed cases (e.g., void fractions and fuel temperatures varied to values approaching their transient 1M'a) other than the ones used in collapsing 3-D'to 1-D cross-sections;(

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ENCLOSURE 2 3

D/mb 3 REQUEST FOR ADDITIONAL INFORMATION

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" STEADY-STATE /NO QUASI-STEADY-STATE METHODS USED IN THE

,.3 ANALYSIS OF, ACCIDENTS AND TRANSIENTS" j

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Fuel Assembly Misloading f

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11, > $upply' data supportir4 your conclusic, that for a disoriented assembly the t

l'c ACPR is more severe at'180* than at 90*.

Since CPM is unable to calculate

'I the 90* cases, have the 90* and 180* cases been evaluated with the same 1

code? Show comparisons of ACPR's f or 90* and 180* assembly misorf entation

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t[as a function of exposure and other pertinent data.

b l.2 ' Describe,ta detail the f our-assembly CPM-PDQ calculation.

Is the PDQ cal-evlation,;.erformed with a pin-by-pin geometry representation? How cany

. cross sectin groups are used in tNs calculation?

Have these calculations

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1.3 What margins are applied to the' ACPR cdculated for the misloaded assembly and tM disoriented assembly te account for uncertainties in the calcula-tNn of the bundle power?

1.4 Demonstrate that the 3.2.1 increase in t.he bundle power of a disoriented as-sembly<f.s an ' upper bound for all current and 1,rojected assemblies.

q Ms 1.5 'lWhat is the' ei'iect of fuel /loadin'g on the R-f actor, bundle power and ACPR

/s tend,how igthis:'ef f ect accounted for in the analysis?

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1.6 (Show r at rerdacenent of an exposed assembly by a f resh assembly in a con-l l \\"i)t

'ttg1 cell results ip a naximus ACPR for the range of fuel types used

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in 6yster Creek.

L 17 Demustrate that tu procedure leading to the selection of the highest cell

,[t delta-exppure and highest cell-averge exposure determines the fuel load-loads.yor having the highest ACPR f or present and future Oyster Creek re-ing er f},

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Control Rod Withdrawal Error 2.1 Desc2We the ore condit/lons assumed in the static calculations of the rod uitMrawaledor.

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i 2.i' Specify ihe core power, exposure, f uel loading, control rod pattern and flow h

in the rod withdrawal error analysis depicted in Figures 3-1 through 3-3 and Table 3-1.

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{bhh (lle j - < ty'. ,L 23 ' What sirgins are assigned to the ACPR's in order to account for uncertain- ?4 ties in the calculation? 24 In view of the large ACPR reduction that occurs during the RVE, demon-strate that the procedures outlined in Section 3 uniquely identify the largest undetected ACPR, when variations in core loadings, fuel types, 9 f failed LPRM detectors and APRM channels are taken into account. That is show that the two' calculations performed in the RWE analysis (strongest rod and largest combination of f ailed LPRM's end APRM channels) are bounding. 3. Loss of Feedwater Heating 3.1 Specify in' detail the plant operating conditions and the assumptions used in the three-dimensional evaluation the loss of feedwater heating (LFVH) event. t 3.2 In the topical report, loss of feedwater heating is attributed to the clo-sure of the steam extraction line causing a gradual cooling of the feed-water. Loss of feedwater heating, however, can also result when feedwater is bypassed around the heaters, causing a f aster cooling of the feedwater. How is this second feedwater heater transient accounted for 1n the analy- ~ sis? 3.3 What benchmarking of the CPU analysis methods has been performed for events of this type? Rave comparisons of NODE-P with Oyster Creek startup mea-l surements been made? What uncertainty allowance has been included in the ACPR calculation to insure that the MCPR safety limit is not violated? 3.4 Describe the statepoints and initial conditions (exposure, rod pattern. flow, void fraction, pressure, inlet subcooling) selected and demonstrate that this selection bounds the maximus ACPR. How many NODE-P calculations are performed in a typical LFWH analysis? 3.5 Describe the procedures used to analyze the LFWH event including the heat balance used to relate the feedwater temperature reduction to core inlet subcooling. 3.6 How is criticality maintained when the power and inlet subcooling are var-j ied independently? If criticality is not maintained between the initial and final states what is the effect of this approximation? 3.7 How is the calculated ACPR adjusted to the initial CPR (IL/R) and what is the basis of this adjus ment? Is this adjustment conservative? 3.8 Are the systas vartsbles such as pressure, feedwater flow, steam flow, etc., assumed constar.: during this transient? If so, provide the basis for this assumption. Can chac,*es in these variables result in a limiting ACPR during the transient making tnc final-state ACPR calculation not bounding? 1 1 a SAFETY PARAW.ETER DISPAY SYSTEM MEETING AIItNDENCE LIST Name Representing R. L. Ferguson NRR/PDI-4 .g Joel Kramer NRR/DLPQE/HFAB Mike Goodman NRR/DLPQE/HFAB Richard J. Eckenrode NRR/DLPQE/HFAB S. H. Weiss NRR/DLPQE/HFAB P. A. Blasioli NU Licensing P. H. Blanch !&C Engineering M. S. Kai NU Safety Analysis l David McDaniel NU - Millstone 3 Engrg i e b AGENDA f 4 O LICENSING STATUS 9 0 SPDS STATUS AND DESIGN CONCEPT OVERVIEW O PARAMETER SELECTION 0 DISCUSS PARTICULAR PARAMETERS CONTAINMENT ISOLATION HOT LEG TEMPERATURE k RHR FLOW CONTAINMENT HYDROGEN CONCENTRATION 0 CONCLUSIONS I i 4 LICENSING STATUS ~ ORIGINAL SPDS SAR SUBMITTED ON APRIL 5,1984 0 0 SEPTEMBER,1984 MEETING TO DISCUSS SAR i - 0 NRC STAFF LETTER DATED MARCH 18, 1985 FORMALLY IDENTIFIED FOUR (4) PARAETERS AT ISSUE FOUR (4) OPTIONS PROVIDED BY NRC STAFF RESPONSE TO NRC STAFF CONCERNS AND REVISED SAR O MAY 24, 1985 SPDSAUDITONJULY29-3b,1985 0 ~ POST-AUDIT DISCUSSIONS 0 AUGUST 2, 1985 AUGUST 13, 1985 SEPTEMBER 18, 1985 0 DRAFT LICENSE CONDITION ON OCTOBER 28, 1985 O COMMENTS SUBMITTED ON NOVEMBER 5,1985 FINAL LICENSE CONDITION ISSUED ON NOVEMBER 25, 1985 0 SUPPLEENT NO, 4 TO MP-3 SER ISSUED ON DECEMBER 6,1985 0 0 EETING WITH STAFF ON JANUARY 8,1986 O JANUARY 10, 1986 DISCUSSION SUPPLEENT NO. 5 TO MP-3 SER ISSUED ON FEBRUARY 14,198S O m____-. SPDS STATUS AND DESIGN CONCEPT OVERVIEW f SPDS DECLARED OPERATIONAL IN NOVEMBER,1985 0 1 INTEGRAL PART OF PLANT PROCESS COMPUTER 0 DESIGNED TO BE USED BY CONTROL ROOM PERSONNEL O 0 FULLY CONSISTENT WITH E0PS O PRIMARY DISPLAYS 0 SECONDARY DISPLAYS 0 ICC DISPLAYS 1 l ~ - ~ - - - - - -.. _ _ _ _ _ _ _ _, _ _ _ _ _ _ l PARAMETER SELECTION l i SPDS - Af D TO CONTROL ROOM PERSONNEL T l 0 STATUS OF THE PLANT AND TO ASSESS 1F COND CORRECTIVE ACTIONS TO AVOID A DEGRADED CORE SPDS FUNCTION IDENTlCAL TO THE FUNCTION OF T O SAFETY FUNCTION STATUS TREES SPDS FUNCTION COMPATIBLE WITH A SYM? TOM 0 ORIENTED APPROACH STATUS TREES PROVIDE AN IDEAL LINK BETWEE 0 EMERGENCY OPERATING PROCEDURES THis PHILOSOPHY PROVIDES A CLEAR, DEFINABLE WAY OF O DETERMINING THE BOUNDS OF SPDS SPDS WILL BE MODIFIED WHEN REVISION ARE 0 STATUS TREES, IF THEY APPLY TO MILLSTONE UNIT NO. 3 ALL VARIABLES INPUT TO THE PROCESS COMPU 0 FROM THE SPDS CONSOLE i i CONTAINMENT ISOLATION 0 SYSTEM STATUS; HDI SYMPT 0M-0RIENTED i' VERIFICATION OF CONTAINMENT ISOLATION IS AN AU 0 OPERATOR ACTION BECAUSE OF THE MILLSTONE UNfT N0, 3 CONTAINMENT DESIGN, 0 CONTAINMENT ISOLATION MONITORING CAN BE ACC MONITORING RADIATION INSIDE AND OUTSIDE CONTA MONITORING RADIATION RELEASES PROVIDES BRO O IDENTIFYING FAILURES IN ISOLATION CONTAINMENT VALVE POSITION IS MONITORED BY 0 WHICH IS CLASS lE ESF STATUS PANEL REVIEWED AS PART OF CRDR A 0 RELIABLE THAN SPDS I _.-.mm_.. n 9 HDT LEG TEMPERATURE i 0 HIGHEST HOT LEG TEMPERATURE IS AVAILABLE FROM THE ICC DISPLAY O PROVIDING ALL HOT LEG TEMPERATURES IS CONSISTENT WITH THE SPDS DESIGN PHlLOSOPHY i i l 0 . ALL FOUR HOT LEG-TEMPERATURES COULD BE ADDED AS SPDS VARIABLES 9 ---__.m_____.__. l POST-LOCA RECIRCULATION FLOW 0 SYSTEM STATUS; RQI SYMPTOM-0RIENTED 0 INCOMPLETE INFORMAT10N ABOUT CORE COOLING 1 0 CORE COOLING STATUS TREE WILL MONITOR POST-LOCA REClRCULAT10N PHASE . 0.. POST-L.0CA RECIRCULATION F!.0WS ARE AVAILABLE FROM THE SPDS CONSOLE ON DEMAND FROM THE OPERATOR G i HYDROGEN CONCENTRATION t 0 STATUS TREE IS APPROPRIATE SINCE HlGH CONTAINMENT RADIATION ~ AND HIGH CONTAINMENT PRESSURE ARE PRECURSORS TO HYDROGEN GENERATION 0 CONTINU0US HYDROGEN MONITOR REQUIRES REMOTE OPERATOR ACTION TO INITIATE . 0.. ONE OF.THE MEANS OF MONITORING HYDROGEN CONCENTRATION IS POST ACCIDENT SAMPLING SYSTEM WHICH IS NOT COMPATIBLE WITH CONTINUOUS MON 1TORING G 0 HYDROGEN CONTROL IS A LONG-TERM CONTAINMENT RECOVERY ITEM THAT DOES NOT RELY UPON IMMEDIATE OPERATOR ACTIONS O HYDROGEN CONCENTRATION IS AVAILABLE FROM THE SPDS CONSOLE ON DEMAND FROM THE OPERATOR [ _

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