ML17279A917

Washington Nuclear Plant 2 LOCA Analysis for Single-Loop Operation

ML17279A917
Person / Time
Site:	Columbia
Issue date:	09/30/1988
From:	Krajicek J SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17279A913	List: GO2-88-053, Application for Amend to License NPF-21,changing Tech Specs to Allow Operation of Plant Up to Power Level of 75% W/One Recirculation Loop Operating.Repts on Analyses of Single Loop Operation Also Encl.Fee Paid ML17279A916 ML17279A917 ML17279A918 ML17279A919
References
ANF-87-118, NUDOCS 8803300294
Download: ML17279A917 (70)

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ADVANCEDNUCLEARFUELS CORPORATION ANF-87-118 Issue Date:

9/1 1/87 WNP-2 LOCA ANALYSIS FOR SINGLE LOOP OPERATION Prepared By:

J.

E. Krajicek BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services Prepared By:

T. Tahvili PWR Safety Analysis Licensing and Safety, Engineering Fuel Engineering and Technical Services AIIAFFIUATEOF KAAFTWEAKIJIIIOII Q~KWV

CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLy Advanced Nuclear Fuels Corporation's warranties and representations con-cemlng the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document is Issued. Accordingly, except as otherwise expressly pro-vided in such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy', completeness, or usefulness of the infor-mation contained in this document. or that the use of any information. apparatus, method or process disclosed In this document will not infringe privately owned rights; or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this docum;nt.

The information contained herein is for the sole use of Customer.

In order to avoid Impairment of rights of Advanced huclear Fuels Corporation in patents or inventions which may be included In the '".lormatlon contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use gn the patent use of the term) of such information until so authorized in writing by Advanced Nuclear Fuels Corporation or until atter six (6) months followingtermination or expiration of the aforesaid Agreement and any extension thereof. unless othermse expressly provided In the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this docu-ment.

XN.NF.F00.765 (

ANF-87-118 TABLE OF CONTENTS Section Pacae

1.0 INTRODUCTION

1 2.0 S UMMARY~

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2 3.0 JET PUMP BWR ECCS EVALUATIONMODEL.................................

5 3.1 3.2 4.0 LOCA During Single Loop Operation..

EXEM/BWR Application To WNP-2 Units ANALYSIS RESULTS AND CONCLUSIONS...

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1 1 4.1 reak Spectrum...............................;.....................

11 B

~ 4.2 MAPLHGR Results.........

12 5.0 EFERENCES..............."..........................................

40 R

ANF-87-118 LIST OF TABLES Table Pacae 3.1 4.1 4.2 WNP-2 Reactor System Data........................

WNP-2 Break Spectrum Results For Normal Operation and SLO.

LOCA Event Times - Single Loop Operation Suction Breaks

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14 15 4.3 MAPLHGR Results'For ANF Fuel

- Single Loop Operation............,..

16 LIST OF FIGURES

~ure 2.1 3.1 4.1 4.2 MAPLHGR vs. Assembly Average Burnup For ANF Fuel WNP-2 System Blowdown Nodalization For BWR/5....

Blowdown System Pressure.....'...................

Blowdown Total Break Flow..

Pacae In WNP-2;.........

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17 18 4.3 4.4 4.5 4.6 4.7 Blowdown Average Core Inlet Flow......,.........

Blowdown Average Core Outlet Flow..

Blowdown Hot Channel Inlet Flow.........,,...

Blowdown Hot Channel Midplane Flow.........

Blowdown Hot Channel Outlet Flow...............

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22 23 4.8 4.9 4.10 4.11 4.12 4.13 Blowdown Blowdown Blowdown Blowdown Blowdown Blowdown Blowdown Blowdown Intact Loop Jet Pump Suction Flow......

Intact Loop Jet Pump Exit Flow.

Broken Loop Jet Pump Drive Flow........

Broken Loop Jet Pump Suction Flow.

Broken Loop Jet Pump Exit Flow.........

Upper Downcomer Mixture.

Middle Downcomer Mixture Level.........

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29 30 31 Lower Downcomer Mixture Level.............

ill ANF-8 LIST OF FIGURES (Continued)

~Fi ure

4. 16 Blowdown Lower Plenum Liquid Mass............................

4. 17 Refill/Reflood System Pressure........................,......

4. 18 Refill/Reflood Lower Plenum Mixture Level....................

4. 19 Refill/Reflood Relative Core Nidplane Entrainment............

4.20 Blowdown Hot Channel Heat Transfer Coefficient...........,...

4.21 Blowdown Hot Channel Center Volume guality.

4.22 Blowdown Hot Channel Center Volume Coolant Temperature.......

4,23 Typical Hot Assembly Heatup Results For ANF Fuel (20 GWd/HTU)

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iv ANF-87-118 ACKNOWLEDGMENT Advanced Nuclear Fuels Corporation appreciates the major contribution to the WNP-2 LOCA analysis for single loop operation made by C.

E.

Hendrix of Intermountain Technologies, Inc.

ANF-87-118

1.0 INTRODUCTION

The results of a LOCA-ECCS analysis for the Supply System Nuclear Project No.

2 (WNP-2) to support single loop operation (SLO) are reported in this document.

These calculations were performed with the generically approved Advanced Nuclear Fuels Corporation (ANF)

EXEM/BWR Evaluation Model(1~

)

in accordance with Appendix K of 10 CFR 50( ),

and the results comply with the U.S.

NRC 10 CFR 50.46 criteria.

The initial SLO condition selected for this analysis was at 75 percent power and 50 percent flow.

The objective of the analysis was to demonstrate that the Maximum Planar Linear Heat Generation Rate (MAPLHGR) vs.

exposure curve calculated for ANF fuel in two loop operation is applicable for SLO.

ANF-87-118 2.0

SUMMARY

The results of the ECCS analysis presented herein support the use of the two loop MAPLHGR's for ANF fuel when the WNP-2 reactor is operating in the SLO mode.

Single loop operation of WNP-2 with the two loop ANF fuel MAPLHGR's assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S.

NRC acceptance criteria of 10 CFR 50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe.

That is:

The calculated peak fuel element clad temperature does not exceed the 2200 F limit.

2.

The calculated total oxidation of the cladding nowhere exceeds 17%

times the total cladding thickness before oxidation.

3.

The calculated maximum hydrogen generation does not exceed 1% of the zircaloy associated with the active fuel cladding in the reactor.

4.

The LOCA cladding temperature transient is calculated to be terminated at a time when the core is still amenable to cooling.

5.

The system long-term cooling capabilities provided for the initial core and subsequent reloads remain applicable to ANF fuel.

The MAPLHGR vs.

exposure curve applicable to ANF fuel in WNP-2 which supports CPR consistent with the flow dependent MCPR curve (1.35 at 50 percent of

ANF-87 rated flow) for two loop and SLO is shown in Figure

2. 1.

The APLHGR limit (MAPLHGR) of Figure

2. 1 is constant at 13.0 kw/ft for assembly average exposures from 0 to 20 GWd/MTU and the limit decreases approximately linearly from 13.0 to 7.9 kw/ft for assembly average exposures from 20 to 35 GWd/MTU.

15 14 13 12 11

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10 CC 8

10 15 20 25 30 ASSEMBLY AVERAGE BURNUP, GWD/NTU 35 40 Figure 2.1 HAPLHGR vs. Assembly Average Burnup For ANF Fuel In WNP-2

ANF-87-118 3.0 JET PUMP BWR ECCS EVALUATION MODEL 3.l LOCA Durin Sin le Loo 0 eration The loss-of-coolant accident (LOCA) break spectrum analysis for a

BWR/5 for two loop operation conditions is described in Reference 4,

and the WNP-2 plant specific MAPLHGR analysis for two loop operation is described in Reference 5.

This document describes LOCA-ECCS analysis and MAPLHGR justification for WNP-2 SLO operation.

The same single failure assumption was made in both the two loop and the SLO LOCA-ECCS analyses.

During SLO the recirculation pump in the inactive loop is not in operation and the rotor is effectively locked while the recirculation flow control valve is at the minimum position( ).

Both intake. and discharge block valves in the

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e recirculation loop remain open during SLO.

A significant resistance to e loop recirculation flow does exist, but a small amount of flow can pass through the idle loop during LOCA conditions.

A break can be hypothesized to occur in either loop.

However, a

break in the inactive loop would behave essentially like a

break during two loop operation except that substantial break flow would come from only one side of the break (because of the fixed pump rotor and recirculation flow control valve at its minimum position).

System performance would then be like that resulting from a somewhat smaller break during two loop operation.

The scenario of breaks smaller than the limiting break has already been covered by the BWR/5 LOCA break spectrum analysis(").

Further consideration in this report will be given only to the case where a break occurs in the active loop.

This case differs from the two loop case in one important respect:

there is no flow coastdown in the intact (idle) recirculation loop.

Previous SLO analysis(

i

)

assumed that the consequence of a

lack of recirculation loop coastdown flow (which continues to supply liquid from the wncomer to the lower plenum, during two loop.operation) would be an almost ediate flow stagnation in the core and a very early CHF (O. 1 sec);

this

6 ANF-87 resulted in degraded heat transfer very early in the transient and required

'that a

MAPLHGR reduction factor of 0.84 be imposed for SLO conditions on the MAPLHGR for NSSS vendor fuel for WNP-2.

3.2 EXEM BWR A lication To WNP-2 The ANF EXEM/BWR ECCS Evaluation model codes were used for this SLO LOCA-ECCS calculation.

The codes which comprise EXEM consist of RODEX2( ),

RELAX(

FLEX( I) and HUXY/BULGEX(

~

).

The latest versions of these codes were used for this SLO

analysis, and they are referenced in the generic two loop analyses("~

).

'In the unlikely event that a

LOCA would occur during

SLO, a rapid drop in core flow would be expected to occur during the early phase of the event because the idle loop pump is not operating.

The core flow transient during the e

phase (0 to 5

seconds) of a single loop LOCA is the principal event wh could distinguish such an accident from a

LOCA occurring during normal two loop operation.

Generally, the magnitude of the initial drop in core flow

'ill 'increase as the break size increases, and the critical heat flux (CHF) may occur earlier in time due to the decreased core flow.

Therefore, the effect of break area on PCT was considered in this analysis to confirm this trend for SLO condition.

The system behavior during a

LOCA is determined primarily by the LOCA break parameters:

break location, break size and break configuration together with the ECCS systems and plant geometry.

Variation in core geometric parameters produce only secondary effects on the system behavior.

Thus, by using bounding core neutronic parameters, the LOCA-ECCS results established by this analysis will apply for future cycles unless substantial changes are made in the plant operating conditions, plant hardware or core design such that the analysis no longer bounds the plant conditions.

ANF-87-118 The system blowdown calculations for the WNP-2 SLO differ from those for two loop operation in several ways.

The back flow through the idle loop jet pump was modeled consistent with expected SLO steady state conditons.

The core,

steam, feedwater and control rod drive coolant flows and reactor power for SLO were taken from Reference 6.

The calculation was performed for a

core composed of all ANF 8x8 fuel.

Other than the changes mentioned, the system blowdown model remains the same as for the two loop model.

The initial conditions are shown in Table

3. 1, and the system blowdown nodalization diagram is shown in Figure 3. 1.

The system blowdown calculations are followed by HOT CHANNEL calculations.

The core geometry is identical with that-used in the two loop analysis.

Th.

initial conditions were adjusted to correspond to those for the single lonp

.. operating point( ).

The

power, axial peaking and flow of the hot assembly determined by an XCOBRA thermal.hydraulics calculation to support a

NCPR sistent with the flow dependent HCPR curve (1.35 at 50 percent flow).

The nodalization and geometry used in the reflood calculation are identical to those of the two loop analysis.

In the FLEX code the intact loop is not modeled in detail because intact loop flows are insignificant at the time of rated low pressure core spray.

Thus, no changes were required in the FLEX nodalization or geometry.

The initial conditions for the reflood calculation are entirely determined by the system blowdown calculation.

The HUXY/BULGEX heatup calculation of the hot plane (center one foot node) was done identically with previous two loop analyses:

fuel stored

energy, gap thermal conductivity and dimensions from RODEX2 as a function of power and exposure; time of rated

spray, decay

power, heat transfer coefficients and coolant conditions from RELAX; and time of hot-node-reflood from FLEX.

Bounding fission and actinide product decay heat obtained with end-of-cycle neutronics in the system blowdown calculation assure that the power input to the HUXY/BULGEX heatup calculation is conservative.

Appendix K spray heat ansfer coefficients are used for the spray cooling period: for heated fuel s,

the values of 1.5, 3.0 and 3.5 Btu/hr-ft2-F are used in the interior,

ANF-87

corner, and peripheral rods respectively; for the unheated fuel channel and water rod surfaces, the values of 5.0 and 1.5 Btu/hr-ft2-F are used.

For the reflood period, the Appendix K reflood heat transfer coefficient value of 25 Btu/hr-ft2-F is used.

Peak cladding temperature (PCT) and the cladding oxidation percentage are specifically determined for the ANF fuel geometry.

ANF-87-118 TABLE 3. 1 WNP-2 REACTOR SYSTEM DATA FOR SINGLE LOOP OPERATION Primary Heat Output, NW (102% of rated)

Total Reactor Flow Rate, lb/hr (49.8% of rated)

Active Core Flow Rate, lb/hr Steam Dame Pressure, psia Reactor Inlet Enthalpy, Btu/lb Recirculation Loop Flow Rate, lb/hr Steam Flow Rate, lb/sec dwater Flow Rate, lb/sec ated Recirculation Pump Head, ft Rated Recirculation Pump Speed, rpm Moment of Inertia, ibm-ft2/rad Recirculation Suction Pipe I.D., in Recirculation Discharge Pipe I,D., in (0.75 x 3389)

= 2542 54.0 x 106 48.26 x 106 1020.

511.2 13.62,x 106 10.62 x 106, 10.59 x 106 800.

1782.

22,700 21.56 21.56

10 ANF-87 AOS 32 STCAM 37 ADS 32 Qe Steam Dome HPCS 3d 75 Qs Upper Downcomer 5s Q4 Upper Plenum 13s 4s E

8 OC st 34 LPCS 7s 07 Feeowater LP'I 17s 08 Qs Bs Oe 28 28s E

3 w

ct Cd 0

14 K

E 0

e c5 Ot 2s Average Core Ct Cda E

o do 18 Containment 31 25 Otd dl Cd Ld tc

~e Cl CC Ci Cdi Cd O

C

.9 V

5o 38 Qii 10s Guioe TubeS Lower Plenum (Lower)

Os 12s 1s Lower P tenum (Upper)

Qte O57 19 05$

20 c

Cd 9s

~g to Qts i 24 23 vn e

E C'd

\\P

~C 16 Qtd Ots Reclrculatlon Pump Flow Control Valve O VOLUME JUNCTION

~ VALVE 21 22

~tt Roc lrculatlon Flow C Val Figure 3. 1 System 8lowdown Nodalization For WNP-2

ANF-87-118 4.0 ANALYSIS RESULTS AND CONCLUSIONS 4.1 Break S ectrum The existing break spectrum for MNP-2, performed for two loop operation, showed the most limiting break to be a split break in the recirculation suction piping with an area of six-tenths of the double-ended pipe area (0.6 DES/RS)( ).

A larger break occurring during SLO might cause an earlier CHF to occur near the core mid-plane resulting in higher PCT's than those calculated for the 0.6 DES/RS break.

Therefore, analysis of a LOCA during SLO was made for two suction break

sizes, at beginning of life (BOL) fuel

exposure, to determine if phenomena unique to SLO initial conditions would cause a change in the previous calculated limiting break.

The two breaks analyzed were a

double-ended guillotine break on the recirculation pump suction side (1.0 y /RS) and the limiting 0.6 DES/RS.

A third analysis for a

double-ended illotine break on the recirculation pump discharge side (1.0 DEG/RD) was performed through the time of rated LPCS to confirm the limiting break remained in the suction line.

Table

4. 1 shows the calculated PCT's for LOCA's initiated from both normal operation and SLO for the 0.6 DES/RS and the 1.0 DEG/RS, breaks.

The calculated PCT for the 0.6 DES/RS break is higher than the 1.0 DEG/RS break for both SLO and normal two loop operation by approximately the same amount.

These results demonstrate that the break spectrum results presented in Reference 4 also apply for SLO.

In addition, the peak fuel rod stored energy and surface temperatures at the time of rated LPCS were compared for the 0.6 DES/RS and the 1.0 DEG/RD breaks.

This comparison showed these temperatures for the 0.6 DES/RS break were higher than those for the pump discharge side break by approximately 100'F at the time of rated low pressure core spray (LPCS).

Also at the time of

LPCS, the blowdown core decay power in the limiting break (0.6 DES/RS) is twelve percent higher than for the discharge eak.

Therefore, it was not necessary to continue the discharge break culation through core reflood.,

12 ANF-8 4.2 MAPLHGR Results The MAPLHGR results were obtained by analysis of the 0.6 DES/RS break at the most limiting fuel exposure( ).

At 20 GWd/MTU the fuel stored energy used to initialize the RELAX/HOT CHANNEL calculation is the maximum calculated to occur in the cycle as determined by the RODEX2 code.

This means that the fuel temperature in the hot channel is at its maximum calculated value regardless of the time in the cycle that the pipe rupture is assumed to occur.

The PCT at 20 GWd/MTU was also observed to the highest value over the exposure range for WNP-2.

With no changes in the fuel design and using bounding neutronic input data to the system blowdown calculation, the MAPLHGR result at 20 GWd/MTU is the only calculation required to show t,'.at the two loop MAPLHGR results are applicable to SLO.

iW Table 4.2 lists the major event times for this analysis.

System bio results.

are given in Figures

4. 1 through

4. 16.

System refill and ref results are shown in Figures

4. 17 through

4. 19.

Results from a

RELAX/HOT CHANNEL calculation are given in Figures 4.20 through 4.22.

The results are applied as boundary conditions. for the HUXY/BULGEX heat up calculation at the most limiting exposure.

The resulting clad temperatures as calculated by HUXY/BULGEX are given in Figure 4.23.

An examination of these plots reveals the following information:

1.

The sudden loss of drive fluid in the jet pumps allowed a

sudden drop in lower plenum pressure of sufficient magnitude to allow flow through the inactive jet pumps to "reverse" from their initial negative flow to a

positive flow in the earlier part of the blowdown.

2.

The lack of a pump coastdown in the intact loop allowed flow through the suction and exit junctions qf the operating (broken loop) pumps to remain in the positive direction (the drive, of cou

13 ANF-87-118 reversed to supply fluid to the break) during the earlier part of the blowdown.

3.

Because of l.

and 2.

above, the initial drop in core flow was not sufficient to cause an early CHF at the core mid-plane.

CHF is delayed until approximately 8 seconds after the tim" of break.

The MAPLHGR results are presented in Tables 4.2 and 4.3.

The PCT values for the two full SLO breaks are approximately 100'F higher than the corresponding breaks in the two loop analysis primarily because the recirculation flow control valve is not in the full open position.

This added resistance does

~cnot affect the system blowdown calculation time, because the

,'low at the break is choked during the system blowdown.

During the latter part of the blowdown, the break is unchoked and the additional resistance from ihe recirculation w control valve decreases break flow and delays the reflood calculation e

by approximately 10 seconds.

Thus, the SLO PCT values are higher than the two loop PCT values primarily due to this system effect, but there is

. still more than 300'F of margin to the PCT limit of 2200'F.

The results presented here for the limiting break (0.6 DES/RS) and highest fuel exposure with maximum stored energy show that the calculated PCT is 1884'F for a

MAPLGHR of 13 kw/ft.

Since the PCT and metal-water reaction percent are well within the Appendix K of 10 CFR 50(

) limits, it is concluded that the MAPLHGR's for ANF fuel as calculated for two loop operation apply for SLO.

A MAPLHGR reduction factor for ANF fuel in WNP-2 is not required for SLO..

ANF-87, TABLE 4. 1 WNP-2 BREAK SPECTRUM RESULTS FOR NORHAL OPERATION AND SLO Break 0.6 DES/RS 1.0 DEG/RS PCT('F)

(Two Loop Operation) 1698 1650 PCT('F)

(SLO) 1786 1752 All.at beginning of life exposure.

f 15 ANF-87-118 TABLE 4.2 LOCA EVENT TIMES - SINGLE LOOP OPERATION SUCTION BREAKS Event Time sec 0.6 DES/RS 1;0 DEG/RS

~304 ft2

~507 ft2 Start Initiate Break Feedwater Flow Stops Steam Flow Stops Low Mixture Level t Pumps Uncover irculation Suction Nozzle Uncovers Lower Plenum Flashes (equality

> 0)

HPCS Flow Starts LPCI Flow Injection Starts Rated Spray Calculated Depressurization Ends (vessel pressure reaches 1 atm)

Start of Reflood (high density fluid enters core)

Peak Clad Temperature Reached 0.00 0.05 0.55 5.05 9.2 11.4 16.8 16.3 17.7 62.2 75.3 126.3 156.9 182.3 0.00 0.05 0.55 5.05 9.0 11.3 16.4 18.4 17.5 62.1 75.0 119.0 150.2 177.0

16 ANF-8 TABLE 4.3 MAPLHGR RESULTS FOR ANF FUEL - SINGLE LOOP OPERATION Result Assembly Average Burnup, GWd/MTU MAPLHGR, kw/ft Peak Cladding Temperature,

'F Peak Temperature Axial Location, ft Local Zr/Water Reaction (Max),

Total Hydrogen Generated,

% of Total Zr Generated 0.6 DES RS Break 20.

13.0 1884 6.25 1.26

HAP-2 0.6 DES/RS BDN QO HO CD e Q.

ill

~O CD CD Lll CZ Q.

~ lO LU Q.

Q O LLJ m Q.

Q.

O CV 10 20 30 40 50 Trt~E.

(SEc) 60 70 80 Figure 4. 1 Blowdown System Pressure ll I

CO I

CO

CI CI CU F)

+O CI Ol NNP-2 O.S OES/AS SON CDM Cflo

~ tO og Mns MKO CXl o CI CI CI i0 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.2 Blowdown Total Break Flow

Cl Cl AO

~

CU WNP-2 0.6 DES/RS BDN Cl Cl Q %

hJ CO

+O QO

~O Ol Cl

<O CI O

~ Cl O<

o Cl TO 10 20 30 40 50 TIME (SEC) 70 80 Figure 4.3 Blowdown Average Core Inlet Flow ll I

CO I

Co

CS

~ O Q tl NHP"2 0.6 DES/RS BDN Cl ED W

VJ QhQO

~O Cl IPI o

0

~ O Wm C)

W~

~O ID CJ 70 10 20 30 40 50 TIVe (SEC) 60 70 80 Figure 4.4 Blowdown Average Core Outlet Flow

WNP-2 0.6 OES/RS EXP HC (3

UJ U)~ IA

<o I

UJ H IIl UJ

~ O 10 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.5 Blowdown Hot Channel Inlet Flow ll I

CO I

CO

HNP-2 0.6 DES/RS SL EXP HC 10 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.6 Blowdown Hot Channel Hidplane Flow I

CO

WNP-2 0.6 DES/RS EXP HC 10 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.7 Blowdown Hot Channel Outlet Flow

>o CU o Q o WNP-2 0.6 DES/RS BDN Oo QJ O Cn '4 (0

OO

~ lO ED CL.

Z O M~

O Co

~O

) O OO l0 20 30 40 50 60 TIvE (SEc) 70 80 Figure 4.8 Blowdown Intact Loop Jet Pump Suction Flow Tl I

Co

CI Oo WO WNP-2 0.6 DES/RS BON CI O CI Ill Ol Qlao

~O CI o

<o po H

Lll CL3g Q ol M

Ch 8

II)

CO 20 30 40 50 TAHE (SEC) 60 70 80 Figure 4.9 Slowdown Intact Loop Jet Pump fxit Flow I

CO I

CO

QO CuO K co HHP-2 0.6 DES/RS BDN OO

~O WO M

Q)~O C)

LL QJ O) O CK C3 CL

~O Cl OOO WI O

TO io RO 30 40 50 TIME (SEC) 70 80 Figure 4.10 Blowdown Broken Loop Jet Pump Drive Flow

mo OJ o Q O

~ CV NAP-2 0.6 DES/RS BDN CD CO m~O Cl g CV o

h.

Oo Ho ED CD CO Q

C.

m CI Cl Cl TO XO 20 30 40 50 60 T1VE (SEC) 70 80 Figure 4.11 Blowdown Broken Loop Jet Pump Suction Flow Tl I

CO I

00

NNP-2 0.6 DES/RS BDN OO O~

LIJ Cfl tQQO

~O tD

<O O

X ILJ QO tQ OO I

O 10 20 30 40 50 60 TIHE (SEC) 70 80 Figure 4.12 Blowdown Broken Loop Jet Pump Exit Flow I

CO

WNP-P.

0.6 DES/RS SL BDN l0 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.13 Blowdown Upper Downcomer Mixture Level Il I

CO I

CO

HNP"2 0.6 DES/RS SL BDN 10 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.14 Blowdown Middle Downcomer Mixture Level

WNP-2 0.6 DES/RS SL BON 10 20 30 40 50 TZVE (SEC) 60 70 80 Figure 4.15 Blowdown Lower Oowncomer Mixture Level rl I

CO I

QO

O

~n gO WNP-P.

0.6 DES/AS SL BDN mO

~1 CD CDo

~O

~O CD M

C5M~n

~O UJ g O CLJ~O 3-CD O

O 10 20 30 40 50 TIME (SEC) 60 70 80

Cl Cl Cl Ol NNP-2 0.6 DES/RS S

EFLOOD O

Cl Cl CI Cl WO p) 0)

O=O

<O Q) OO EU CL~ Cl Cn OO COO QJ O

~ Cl CL I-O QO Z O Hn 0

UJ~O Cl Cl Cl Ol 20 40 60 80 100 TIME (SECS.)

120 140 160 Figure 4.17 Ref ill/Ref1ood System Pressure rl I

CO I

00

NNP-2 0.6 DES/RS SL REFLOOO 20 40 60 80 100 TreE (SECS.j l20 i40 160 Figure 4.18 Refill/Reflood Lower Plenum Mixture Level

WNP-2 0.6 DES/RS S

EFLOOD CO QO UJ lLJ CL CV Cl c80 SO 100 110 120 130 TIME (SECS.)

140 150 160 Figure 4.19 Refill/Reflood Core Hidplane Relative Entrainment

HNP-2 0.6 DES/RS SL EXP HC io 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.20 Blowdown Hot Channel Heat Transfer Coefficient

WNP-2 0.6 DES/RS EXP HC 10 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.2I Blowdown Hot Channel Center Volume equality Tl I

CO I

CO

WNP-2 0.6 DES/RS SL EXP HC C)) g)

CL tU hJQo I

~~

CL UJI CD o

)o C))

10 20 30 40 50 TIME (SEC) 60 70 80 Figure 4.22 Blowdown Hot Channel Center Volume Coolant Temperature

0.6 DES/RS SL 2

AABU 2500 u

2000 1

2 3

4 5

6 7

8 2

9 io 1112131415 3 10161718192021 4 11172223242526 5 12182327282930 6 13192428313233 7 14202529323435 8 15212630333536 PIN28 PIN 13 CANISTER 1500 1000 CJ 50 40 80 120 160 200 TIME (SEC}

240 280 320 Il 1

CO I

CO Figure 4.23 Typical Hot Assembly Heatup Results For ANF Fuel (20 GWd/MTU)

40 ANF-87-118

5.0 REFERENCES

2.

3.

7.

8.

9.

10.

12.

13.

"Generic Jet Pump BWR 3

LOCA Analysis Using the ENC EXEM Evaluation Model," XN-NF-81-71(A),

Supplement 1,

Exxon Nuclear

Company, September 1982.

"Generic LOCA Break Spectrum Analysis for BWR 3 and 4 with Modified Low Pressure

. Coolant Injection Logic,"

XN-NF-84-117(P),

Exxon Nuclear

Company, December 1984.

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors,"

10 CFR 50.46 and Appendix K of 10 CFR 50, Federal

Register, Volume 39, Number 3, January 4,

1974.

"lffA B

2 Ep 1

~

A I

I f

BNR B,"

X -

ft., f" Nuclear Company,'ecember 1985.

"WNP-2 LOCA-ECCS Analysis; MAPLHGR Results,"

XN-NF-85-139, Exxon Nuclear

Company, December 1985.

Letter, R.

A. Vopalensky to J.

B. Edgar, "Supply System Data Package No.

~

~

~

~

44," WANF-2B-87-0023, March 2, 1987.

"General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance With 10 CFR 50 Appendix K,

Amendment 2-One Recirculation Loop Out-of-Service,"

NED0-20566-2, Revision 1, Class 1, July 1978,

Letter, R.

A. Vopalensky to J.

B. Edgar, "Single Loop Operation Contract No. 2808-2B," WPEN-2B-85-0075, July 11, 1985.

"RODEX2:

Fuel Rod Thermal-Mechanical

Response

Evaluation Model," XN-NF-

~81-88 A

, Revision 2, Exxon Nuclear

Company, February 1983.

"RELAX:

A RELAP4-Based Computer Code for Calculating Blowdown Pk,n~X-

- -l,fl 2B,R I

I,E I

t 1,

June 1981.

"FLEX:

A Computer Code for the Refill and Reflood Period of a LOCA," XN-

~f-->>,f12B,R 11 I,E II I

p,f lpl.

"HUXY:

A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K

N p

Pp1 U

I n ~A, R

11 I,

E N

Company, November 1,

1975.

"BULGEX:

A Computer Code to Determine the Deformation and the Onset of Bulging of,Zircaloy Fuel Rod Cladding,"

XN-74-21, Revision 2,

and XN-NF-27, Revision 2, Exxon Nuclear

Company, December 31, 1974.

ANF-87-118 Issue Date:

9/11/87 WNP-2 LOCA ANALYSIS FOR, SINGLE LOOP OPERATION Distribution D. A. Adkisson R.

E. Collingham J.

G.

Ingham S.

E. Jensen T.

H. Keheley D.

C, Kilian J.

E. Krajicek J.

L. Haryott G.

L. Ritter H.

E. Williamson J.

B, Edgar/WPPSS (50)

Document Control (5)

4 S

~R REP(p P0 Cy AnO C

0 Cy ~

p +y*y4 UNITEDSTATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20555 April 18, 1988 8805120280 MEMORANDUM TO:

John'W. Craig, Acting Chief Plant Systems Branch

~

Division of Engineering 5 System Technology Conrad E. McCracken, Acting Chief Chemical Engineering Branch Division of Engineering 5 System Technology Faust Rosa, Chief Electr ical Systems Branch Division of Engineering 5 System Technology THROUGH:

FROM:

SUBJECT:

George W. Knighton, Director Project Directorate V

Division of Reactor Projects - III, IV, V and Special Projects Robert Samworth Project Directorate V

Division of Reactor Projects - III, IV, V and Special Projects WNP-2 FIRE INSPECTION Charles Ramsey of Region V has initiated a team inspection of fire protection issues at WNP-2 to take place June 6 through June 10.

The inspection will have the-primary objective of closing open items.

The inspection will follow Inspection Module 64100, titled "Postfire Safe Shutdown, Emergency Lighting

'nd Oil Collection Capability at Operating and Hear-Term Operating Reactor Facilities" and Module 64704, "Fire Protection/Prevention Program."

Where

possible, the inspection should close out issues identified (e.g., in previous inspections) and reviewed by the staff.

I have enclosed the draft inspection plan prepared by Mr. Ramsey, along with the source documents listed in the plan.

I have also enclosed the schedule for critical events in the inspection.

Please review and comment on the inspection plan as appropriate with the objective of focusing the inspection for efficient utilitization of the limited manpower being made available.

We would like to finalize the inspection plan this month to give adequate time for collecting necessary documents for preparation for the inspection.

Therefore, please provide your input by April 25th.

Robert B. Samworth, Senior Project Manager Project Directorate V

Division of Reactor Projects - III, IV, V and Special Projects

a, 1

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i t es.4

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'pril 8th April 18th April 22nd April 25th May 23rd - 27th June 6th - 10th June 10th WNP 2 FIRE INSPECTION SCHEDULE (TENTATIVE)

Staff input due back to Ramsey Final Plan Management Approval (Knighton)

Announce Inspection to licensee, Tell them what documents are needed.

Inspection Preparation (Get info from licensee, Review FSAR, Familiarize self with plant

& issues.

Meet'at site at 8 a.m.

on Monday, 6/6, tour of plant, meet with licensee staff (entrance meeting) 1 p.m. exit meeting

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Ins ection Plan for Re-ins ection of WNP-2 Usin Ins ection Module No. 64100 Team Leader:

Chuck Ramsey, Region V, FTS 463-3767

~Back rovnd Region' Inspection Report Nos. 50-397/86-05, 397/86-25, 397/87-02, and 397/87-19 document 14 open/unresolved items and one violation relating to fire protection and safe shutdown capabili'ty at WNP-2.

In addition, by letter dated November 11, 1987 (G.

W. Knighton, NRC, to G.

C. Sorensen, SS), the NRC forwarded a Safety Evaluation Report addressing fire protection and safe shutdown capability as described in Amendment No. 37 to the WNP-2 FSAR, in which 28 open items were forwarded to the licensee with a requested response within 60 days.

The licensee's resolution to these issues appears to have resulted in a significant change in the safe shutdown methodology described by the licensee in Amendment No.

19 to the WNP-2 FSAR which was reviewed and accepted by the NRC in a. Safety Evaluation Report dated March, 1983 (NUREG 0892).

Furthermore, followup on Part 21 Report Nos.86-14P, 86-14P1; LER Nos.

84-31-L6, 87-29-LO, 87-30-LO; Information Notice Nos. 61-06, 85-09; and Inspection Module No. 64704 regarding safe shutdown and routine fire protection features need to be performed.

~0b ective'he objective of this inspection is to obtain verification of the adequacy and acceptableness of the licensee's post fire safe shutdown physical configuration, methodology, operating procedures, and routine fire protection program implementation.

The scope of the inspection is limited to areas of concern identified in one violation, 14 Region Y open/unresolved

items, 28 NRR

open items, two Part 21 reports, three LERs, one Information Notice, and

~

~

routine Inspection Modu'le NO. 64704.

Tasks Scheduled Com letion Date l.

One week to prepare for inspection by reviewing background material designated or provided by the inspection team leader and provide input into the inspection plan.

1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> Nay 15, 1988 2.

Attend entrance meeting with licensee and participate in site tour.

0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> June 6, 1988 3.

Perform inspection of licensee corrective actions to concerns by assignment as.follows a.

Systems Engineer(s) will verify licensee's corrective actions for hot/cold shutdown concerns from inside and outside of the control room, with and without onsite power.

1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> June 10, 1988 b.

,Electrical Engineer(s) will verify licensee's corrective actions for associated circuits relative to common bus, common enclosure, spurious signal and electrical separation concerns.

1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> June 10, 1988 c.

Fire Protection Engineer(s) will verify licensee's corrective actions

for physical separation, fire barrier/

detection and suppression provided for safe shutdown capability concerns.

In addition, verify the l.icensee's implementation of routine fire protection program activities as required by Inspection Module 64704.

1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> June ll, 1988 4.

Confer with team leader as necessary and confirm basis for determinations made.

1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> Daily 5.

Prepare Summary of inspection findings:

a.

Draft summary of results.

1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />.

June 11, 1988 b.

Discuss findings as needed at exit meeting with licensee 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> June 10, 1988

- c.

Provide handwritten work copy of findings to team leader 5 days after exit meeting d.

Provide final inspection report input to team leader 10 days after exit meeting 6.

Followup on Inspection Findings Assist in the close-out of followup items generated as a result of the inspection.

As needed.

Level of Effort and Period of Performance

~

~

~

The level of effort is estimated at 600 professional staff hours.

The inspection is scheduled for the period June 6-10, 1988.

Inspection team

P

members are required to arrive at the site for the entrance meeting at 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br /> on June 6, 1988.

The entrance meeting will be held at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />.

Onsite inspection effort is scheduled to terminate at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on Friday June,10, 1988.

Meetin s and Travel All inspection team members are required to attend the entrance/exit meetings with the licensee and daily meetings (approximately one hour starting at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />).

Travel arrangements should be made to arrive at motel accommodations near the site in Richland, Washington, by Sunday, June 5, 1988.

A minimum of two vehicles will be used for transportation of inspection team members at the site.

Staff Time Re uirement Inspection preparation Travel to and from site Inspection Weekend days on travel Overtime worked Documentation 5 days 2 days 6 days 2 days 1 day (one day of overtime expected to accumulate during the week and one day on Saturday) 10 days Period of Performance Preparation/in-office review Onsite inspection Documentation (1 week)

May 23-May 27 (5 days)

June 6-June 10 (2 weeks)

June 14-June 28

T~A A.

S stem En ineer s

1.

LER No. 84-31 Lo-L6 - Appendix R Deficiencies 2.

Unresolved Item 86-25 Safe Shutdown Methodology 3.

Unresolved Item 86-25 Safe Shutdown Procedures 4.

Unresolved Item 87-19 Loss of All AC Power Due to Fire 5.

Unresolved Item 87-19 Use of ADS Inhibit Switch 6.

Unresolved Item 86-25 Emergency Lighting Support for S,S.

7.

Unresolved Item 86-05 Morst Case Fire Analysis for CR and CSR 8.

NRR Question No.

9 - Emergency Lighting 9.

NRR Question No.

10 - Number of ADS Values Needed for S.S.

10.

NRR Question No.

12 - Use of Both Divisions for S.S.

ll.

NRR Question No.

13 - Equipment Needed for S.S.

12.

NRR Question No.

16 - Testing of S.S.

Components 13.

NRR Question No.

17 - Use of Installed Transfer Switches 14.

NRR Question No.

21 - SM-8 Cabinet Transfer Switches 15.

NRR Question No.

18 - Operability of Six SRVs 16.

NRR Question No.

19 - Process Instrumentation 17.

NRR Question No. 20 - Required Repairs for S.

S.

18.

NRR Question No.

23 - RHR V-8 and V-9 LOCA Prevention 19.

NRR Question NO. 26 - Minimum Staffing Needed for S.S.

20.

T.I. 2515 Emergency Lighting 21.

T.I. 2515 Appendix R, S.S.

JE

B.'lectrical En ineer s

1.

LER No. 84-03 Lo-L6 - Appendix R Deficiencies 2.

Information Notice 86-106 - Interaction Between Fire Protection and Security Systems 3.

Information Notice 85-09, Isolation Transfer Switches 4.

Unresolved Item 86-05 Common Enclosure Analysis 5.

Unresolved Item 86-25 Associated Circuits Analysis 6.

Unresolved Item 86-25 Hi-Low Pressure Interface Analysis 7.

Unresolved.Item 87-19 Loss of All Sources of AC Power 8.

Unresolved Item 86-05 Worst Case Fire Analysis for CR and CSR 9.

NRR Question No. 8 - Conformance with Regulatory Guide 1.75 10.

NRR Question No.

11 - Conformance with Appendix R and BTP 9.5-1 11.

NRR Question No.

12 - Protection of Both Divisions for S.S.

12..

NRR Question No.

16 - Testing of S.S.

Components 13.

NRR Question No.

17 - Testing of Installed Transfer Switches 14.

NRR Question No. 21 - SN-8 Cabinet Transfer Switches 15.

NRR Question No. 22 - Coordinated Circuit Protection 16.

NRR Question No. 23 - Power to RHR Valves Y-8 and V-9 17.

NRR Question No.

24 - Divisional Power to RHR Valves Y-8 and V-9 18.

NRR Question No'.

25 - 3 Phase Faults at Hi-Lo Pressure Interfaces 19.

T.I. 2515 Appendix R., S.S.

i

Fire Protection En ineer s

2.

3.

5.

6.

7.

9.

10.

11.

12.

13.

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

LER 84-031 Lo-L6 - Appendix R Deficiencies LER 87-29 Lo - Fire Rated Floor Penetrations LER 87-30 Lo - Unqualified'ire Wall Part 21 Report 86-14 P - Automatic Sprinkler Model C Yalves Part 21 Report 86-14 P1 - Automatic Sprinkler Model C Valves Information Notice 61 Interaction of Fire Protection and Security Systems.

Unresolved Item 87-02 Unqualified Barriers for Protection of S,S.

Components Unresolved Item 86-05 Morst Case Fire Analysis for CR'nd CSR Unresolved Item 86-25 Fire Main Beneath Safety-Related Structures Unresolved Item 86-25 Deficiencies Not Documented in LER 84-031 Unresolved Item 86-25 CSR Design Unresolved Item 86-25 Fire Detection

System Design

Unresolved

!tern 86-25 Emergency Lighting Unresolved Item 87-19 Loss of All AC Power Due to Fire Unresolved Item 87-19 Fire Threat to Shutdown Division and Oil Transfer Pump Rooms NRR Question No.

1 - Combustible Inventory Fire Area TG-1 NRR Question No.

2 - Bus Duct Penetration Fire Barriers NRR Question No.

3 - Closure of Fire Dampers Under Air Flow NRR Question No. 4 - Protection of Instrument Sensing Lines NRR Question No.

5 - Separation of Redundant Trains in Seismic Gap NRR Question No.

6 - Inspection of Mater Control Yalves NRR Question No.

7 - Holan Cylinder Storage Procedures NRR Question No.

9 - Emergency Lighting NRR Question No.

11 - Appendix R and BTP 9.5-1 Conformance NRR Question No.

12 - Protection of Both Divisions for S.S.

NRR Question No.

15 - Protection for Division II SRVs NRR Question No.

27 - Regulatory Guide 1.39 Procedure Conformance NRR Question No. 28 - NFPA Administrative Procedure Conformance

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Module 64704 - Routine Program Implemention 30.

T.I. 3515 Emergency Lighting 31.

T.I. 2515 Appendix R Safe Shutdown

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Review Mater ia"Is Re uired

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Ins ection Re orts 397/86-05 397/86-25 U'97/87-02M 397/87-19 v'.

Licensee Event Re orts LERs 84-31 Lo-L6 87-29 Lo 87-30 Lo 3.

Information Notices 61-06 o

85-09 t.

~Pt Pit t

86-14 P

86<<14 P1, 5.

Safet Evaluation Re orts SERs NUREG 0892 SER attached to letter dated November 11, 1987 ~

6.

Licensee's Documents Associated Circuits Analysis Safe Shutdown Equipment List January 11, 1988 Response to NRR Open Items Responses to Inspection Report Open Items