ML20043F057

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Rev 1 to Grand Gulf Unit 1 Cycle 5 Plant Transient Analysis
ML20043F057
Person / Time
Site: Grand Gulf 
Issue date: 05/31/1990
From: Hibbard M, Macduff R
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20043F017 List:
References
ANF-90-021, ANF-90-021-R01, ANF-90-21, ANF-90-21-R1, NUDOCS 9006140127
Download: ML20043F057 (50)
v • d • e


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I ANF-90 021

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REVISION 1 Y

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l1 k1 ADVANCEDNUCLEARFUELS CORPORATION t

1 GRAND GULF UNIT 1 CYCLE 5 PLANT TRANSIENT ANALYSIS T

MAY 1990 l

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ADEMCSDNUCLEARFLELSCORPORATKW ANF-90 021 Revision 1 i

Issue Date: 05/23/90 4

i GRAND GULF UNIT 1 CYCLE 5 PLANT TRANSIENT ANALYSIS r

Prepared by

't. B. Macduff ff i

l BWR Fuel Engineeefng Fuel Engineering and Licensing N, f, a [ [t M. J. Hibbard BWR Fuel Engineering Fuel Engineering and Licensing i

May 1990 4

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Revision 1 Page i I

TABLE OF CONTENTS 1:sno.c Ene

1.0 INTRODUCTION

1-2.0

SUMMARY

4 3.0 THERMAL LIMITS ANALYSIS.......................

16 i

3.1 Introduction..........................

16 3.2 System Transients 16 3.2.1 Design Basis 17 3.2.2 Anticipated Transients 17 3.2.2.1 Loss Of Feedwater Heating la 3.2.2.2 Load Rejection No Bypass............

18 3.2.2.3 Feedwater Controller Failure..........

19 3.2.2.4 Control Rod Withdrawal Error.......... 20 3.3 Flow Excursion Analysis 20 3.4 Safety Limit..........................

21 3.5 Summary of Results.......................

22 3.5.1 Power Dependent Thermal Limits and Values........ 22 3.5.2 Flow Dependent Thermal Limits and Values 23 3.5.3 Exposure Dependent Thermal Limits............ 23 4.0 MAXIMUM OVERPRESSURIZATION 33 4.1 Design Basis.......................... 33

'4.2 Maximum Pressurization Transients 33 4.3 Results 34

5.0 REFERENCES

38 APPENDIX A SINGLE-LOOP OPERATION A-1 j

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ANF-90-021 Revision 1 Page ii LIST OF TABLES IAhlt Elat 2.1 Results of Analyses 6

2.2 Operating Limit Coordinates 8

3.1 Grand Gulf Unit 1 Cycle 5 LFWH Data Summary 24 LIST OF FIGURES Fiaure EARA 1.1 Power / Flow Map Used for Grand Gulf Unit 1 ME00 Analysis 3

4 2.1 Exposure Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 5....

11 l

2.2 Power Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 5 12 2.3 Power Dependent LHGRFAC Value for Grand Gulf Unit 1 Cycle 5 13 2.4 Flow Dependent MCPR Limits for Grand Gulf Unit 1 Cycle 5......

14 2.5 Flow Dependent LHGRFAC Value for Grand Gulf Unit 1 Cycle 5 15 L

3.1 Analysis of LFWH Initial MCPR Versus Final MCPR 25 3.2 Load Rejection Without Bypass Power and Flows) 26 l

3.3 Load Rejection Without Bypass Vessel Pressure) 27 1

3.4 Load Rejection Without Bypass Vessel Level Above Separator Skirt).

28 Feedwater Controller Failure (Power and Flows 29 3.5 Feedwater Controller Failure (Dome Pressure) )............

3.6 30 3.7 Feedwater Controller Failure (Vessel Level Above Separator Skirt) 31 3.8 Grand Gulf Unit 1 Cycle 5 Safety Limit Design Basis Local Power Distribution......................

32 4.1 MSIV Closure Without Direct Scram Power and Flows) 35 4.2 MSIV Closure Without Direct Scram Vessel Pressure) 36 l-4.3 MSIV Closure Without Direct Scram Vessel Level Above Separator Skirt)................

37 A.1. Pump Seizure Event SLO (Power and Flows).............

A-3 A.2 Pump Seizure Event SLO (Vessel Pressure).............

A-4 A.3 Pump Seizure Event SLO (Vessel Level Above Separator Skirt)

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ANF40 021 Revision 1 Page 1

1.0 INTRODUCTION

This report. presents the results of analyses performed by Advanced Nuclear Fuels Corporation (ANF) for reload fuel in Grand Gulf Unit 1 Cycle 5 l

for operation within the Maximum Extended Operating Domain (ME00).

In Cycle 1 (Reference 1) the NSSS vendor performed extensive transient analyses for Grand i

Gulf Unit 1 in conjunction with the extension of the power / flow operating map i

to the ME00.

These analyses established conservative operating limits for ME00 operation.

The initial reload of ANF fuel in Grand Gulf Unit 1 occurred in Cycle 2.

In support of the initial reload of ANF fuel, extensive additional transient analyses were performed by ANF to justify the NSSS vendor operating limits and, where necessary, provide appropriate limits for ANF fuel using ANF methodologies (Reference 2).

Cycle 5 for Grand Gulf Unit I will include a first reload of ANF 9x9 5 fuel.

The nominal cycle energy remains 1698 GWd and the cycle length remains 18 months.

The reload fuel for Cycle 5 is ANF 9x9 5 (Reference 15).

New methods are employed for the analysis and include the use of the CASMO 3G/MICROBURN-B codes (Reference 7), COTRANSA2 system analysis methods

- (Reference 5), revised safety limit methodology (Reference 9), and the use of ANFB Critical Power Correlation (Reference 14) in XCOBRA and XCOBRA T.

XCOBRA and XC9 BRA-T are changed only by the inclusion of ANFB, The Cycle 5 transient analysis consists of recalculation of the limiting transients at state points having the least margin to operating limits to confirm that the i

effects of the Cycle 5 changes on transient results are small relative to L

available margin and/or establish appropriate limits.

Reanalysis of the limiting transients for Cycle 5 assures that the less limiting transients l_

which were previously addressed will continue to be protected by the established operating -limits for Cycle S.

The power / flow conditions analyzed in Cycle 5 are presented in Figure 1.1.

Analyses were performed at EOC-2000 mwd /MTV, at EOC, and at E0C+30 EFPD (Effective Full Powsr Days),

. providing limits for Cycle 5 that are exposure dependent.

These analyses establish the Grand Gulf Unit 1 Cycle 5 Technical Specification MCPR at rated conditions, establish MAPLHGR limits for Cycle 5 T

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ANF 90 021 Revision 1 Page 2 operation, and establish revised thermal limits for off rated conditions for the transition toward an all 9x9 5 core.

The analyses also demonstrate that vessel integrity is protected during the most limiting Cycle 5 pressurization

event, i

i The MCPRp and MCPRr limits have been revised to reflect ANF calculated limits using new ANF methodology.

Consistent with the new methodology, MAPFACf limits have been replaced with LHGRFAct limits.

Similarly, MAPFACp limits have been replaced with LHGRFAC limits.

LHGR protection has been p

established for both 8x8 and 9x9 5 fuel in Cycle 5.

The Grand Gulf Unit 1 power and flow dependent MCPR analyses for Cycle 5 were performed at limiting power / flow conditions.

Appropriate analyses were performed to substantiate the LHGRFACp values.

Flow dependent LHGRFAC values were confirmed with analyses performed on the 100% rod line with the initial core flow increasing from 40% to 107% of rated flow.

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1 ANF 90 021 Revision 1 Page 4 2.0 SWMARY The results of the Grand Gulf Unit 1 Cycle 5 transient analyses support appropriate thermal limits for the Grand Gulf core including the ANF 1.4 9x9 5

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reload.

ANF thermal limits have been provided for MCPR between 40% and 70%

p power that are based on generic ANF Control Rod Withdrawal Error (CRWE) analyses (Reference 4).

Above 70% power, the thermal limits have been determined from transient analysis with an operating limit MCPR of 1.19.

MCPR limits have been verified or established at powers less than 40% based

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p on the LRNB transient.

Additionally, MCPRf limits and LHGRFACr values (Reference 12) have been verified or established for both " loop manual" and "non loop manual" operation.

j With the change to LHGRFAC from MAPFAC, separate MAPLHGR limits for each fuel type are no longer required since LHGR limits will be monitored directly.

Consequently, the 8x8 MAPLHGR (Reference 16) has replaced all previous MAPLHGR l

limits for 8x8 fuel in the Technical Specification and a new MAPLHGR limit for 9x9 5 fuel has been introduced.

MAPLHGR limits satisfy the requirements l

specified by 10CFR50.46 of the U.S. Code of Federal Regulations.

The 8x8 and 9x9 5 LHGR limits will be protected at off-rated conditions by applying LHGRFACf and LHGRFACp multipliers on the Technical Specification LHGR limits, l

Table 2.1 summarizes the transient analyses results applicable to Grand-Gulf Unit 1 Cycle 5.

These results, together with the Grand Gulf Unit 1 Cycle 5 calculated safety limit MCPR of 1.08, support use of a 1.19 MCPR operating limit (at rated conditions) for Cycle 5 operation between BOC and E00-2000 mwd /MTU.

The operating limit (at rated conditions) from E00-2000 mwd /MTU to EOC is supported at 1.29.

For extended operation from EOC to EOC+30 EFPD the operating limit (at rated conditions) is 1.30.

Figure 2.1 presents the exposure dependent MCPR, limit as a function of core average exposure.

The calculated safety limit of 1.08 includes the assessment of the channel bow impact using appropriate ANF methods (Reference 9).

The plant transient and safety limit analyses results reported herein establish power dependent Minimum Critical Power Ratio (MCPR ) limit.

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ANF-90 021 Revision 1 Page 5 power dependent Linear Heat Generation Factor (LHGRFAC ) is presented for p

Cycle 5 operation for all fuel types.

The revised MCPR limits, the LHGRFAC p

p values, and the corresponding results of ANF's analyses are presented in Figures 2.2 and 2.3.

The flow dependent Minimum Critical Power Ratio (MCPR ) limit and the f

results of ANF's analysis; are presented in Figure 2.4.

The flow dependent Linear Heat Generation Rate Factor (LHGRFAC ) is presented in Figure 2.5.

f These flow dependent LHGRFACf values and MCPRf limits have been verified or established for Cycle 5 to support both the " loop manual" and the "non-loop manual" mode of operation.

These curves are based on conservative maximum core flow rates.

Table 2.2 shows the coordinates used to construct Figures 2.1 through 2.5.

The implementation of the MCPR operating limit requires that the most l

restrictive operating limit be chosen from among the three MCPR curves based l

on exposure, flow, and power.

Thus, the greater value of MCPR as given by MCPR,, MCPR, or MCPR is selected as the operating limit in accordance with f

p the state point of operation (Figures 2.1, 2.2, and 2.4).

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The results of the maximum system pressurization transient analysis are presented in Table 2.1.

The results show that the Grand Gulf Unit I safety valves have sufficient capacity and performance to protect the vessel pressure safety limit of 1375 psig during Cycle 5.

The fuel related Technical Specification limits for Cycle 5 operation are included in the reload analysis report (Reference 3).

4

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i ANF-90 021 Revision 1 Page 6 Table 2.1 Results of Analyses 1

THERMAL LIMITS Transient Delta CPR Loss of Feedwater Heating (all conditions) 0.11 Control Rod Withdrawal Error (100% power, Ref. 4) 0.10 Delta-CPR EOC-2000 EOC 111. 2 d AKA_ 211_

Feedwater Controller Failure

  • 0.07 0.08 0.13 0.13 (Without Bypass) (104.2/108)

MAXIMUM SYSTEM PRESSURIZATION l.

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% Power /% Core Flow Vessel Lower Plenum Steam Dome MSIV Closure 104.2/108 1291 psig 1268 psig 104.2/75 1286 psig 1269 psig l

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Table 2.1 Results of Analyses (Continued) i Load Reiection Without Bvoass Delta CPR

% Power /% Core Flow E00-2000 EOC EOC+30EFPD Ax8 2x2_

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104.2/108*

0.06 0.05 0.20 0.20 0.21 0.21 104.2/75 0.02 0.02 0.04 0.05 J

70/40 0.05 0.05 0.06 0.07 40/108 0.10 0.11 0.16 0.19 40/108**

0.29 0.34 0.28 0'32 J

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25/73 0.82 1.10 0.79 1.05 0.77 1.03 4

0.61 0.84 0.60' O.82 25/30,,

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i ANF 90 021 Revision 1 i

Page 8 Table 2.2 Operating Limit Coordinates l

GRAND GULF UNIT 1 CYCLE 5 MCPRfe) Limits j

(Figure 2.1)

Core Average Exposure MCPR(e)

GWd/MTU i

12.872 (BOC) 1.19 22.948 (EOC 2000) 1.19 22.948 1.29 24.948 (EOC) 1.29 24.948 1.30 25.766 (E0C+30 EFPD) 1.30 MCPRio) Limits (Figure 2.2)

Percent of Rated MCPR(p)

Core Power 100 1.19 70 1.24 70-1.40 40 1,48*

40 1.85 40 2.10**

25 2.05**

25 2.20*

  • Core Flow 1 50%.

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ANF-90 021 Revision 1 Page 9 Table 2.2 Operating Limit Coordinates (Continued)

LHGRFACfo) Limits (Figure 2.3) j Percent of Rated Core Power LBGE&fdg1 100 1.00 40 0.60**

40 0.69*

24.4 0.57**

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    • Core Flow > 50%.

MCPRff) Limits (Figure 2.4)

Percent of Rated Core Loop Non Loop F10w BADMAl Manual 20 1.35 1,63 30 1.35 1.63 i

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-Table 2.2 Operating Limit Coordinates (Continued)

LHGRFAC(f) LIMIl$.

(Figure 2.5)

Percent of Loo Non Loop _

Rated Core Manu!1 Manual Flow 110.0 1.00 1.00

&l.0 1.00 1.00 90.0 1.00-0.992.

84.3 1.00 80.0 0.977 0.904 70.0 0.928 0.827 60.0 0.880 0.757 50.0 0.8?7 0.695 40.0 0.794 0.638 30.0 0.752 0.586 20.0 0.752-0.586 4

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ANF-90 021 i

Revision-1 Page 16-3.0 THERMAL LIMITS ANALYSIS 3.1-Introduction The scope of the thermal limits analysis includes system transients, localized core events, and safety limit analysis.

Results of these analyses.

are used to confirm or' establish power, flow, and exposure - dependent MCPR limits and-LHGRFAC values as appropriate.

COTRANSA2 (Reference 5), XCOBRA-T -(Reference 6), XCOBRA (Reference 19),

and MICROBURN B (Reference 7) are the major codes used in the thermal limits analyses as described in ANF's THERMEX Methodology Report (Reference 8). and.

Neutronics Methodology Report (Reference 7).

COTRANSA2 is a system transient simulation code which includes an axial one-dimensional neutronics model.

XCOBRA-T is a transient thermal-hydraulic code used in the analysis of thermal margins-of 'the limiting fuel assembly.

MICR0 BURN-B is a three-dimensional steady state core simulation code which is used for Control Rod Withdrawal l

Error - (CRWE), Loss of Feedwater Heating (LFWH), and flow excursion events (LHGRFAC ).

XCOBRA is-a steady state Thermal Hydraulic - code used in the f

l analysis of. slow flow excursion events (MCPR ).

The ANFB Critical Power f

Correlation (Reference 14)- evaluates the thermal margins of the fuel L

assemblies.

This correlation has been generically approved by the NRC L

(Reference'17).

3.2 System Transients Thermal limits have been appropriately revised based upon ANF methods used in the Cycle 5 analysis.

Figure' 1.1 shows the eight power / flow conditions.that were analyzed in support of the Cycle 5 reload.

System response for pressurization transients from these state points were analyzed for Cycle 5 using COTRANSA2.

The Load Reject No Bypass (LRNB) pressurization transient analysis was performed at each of the eight state points including a single-loop operation state point.

The Feedwater Controller Failure (FWCF) p analysis was performed at 104.2%/108%.

ASME pressurization analyses were performed at state points 104.2%/108% and 104.2%/75%.

LFWH analyses were performed with MICR0 BURN-B at eight exposure points in the cycle.

Analyses

in

'ANF 90 021 Revision 1 Page 17 have been performed considering the ANF 9x9-5 fuel to assure that the power i

dependent limits supported by analyses for control rod withdrawal error remain applicable to Grand Gulf Unit 1 Cycle 5.

These analyses show less restrictive results ~or iittk change from the Cycle 4 analyses due to Cycle 5 changes, thus justifying that the-less limiting transients not analyzed for Cycle 5 will continue to be protected. The pump seizure event and load. reject without~

bypass were analyzed for. single-loop operation for Cycle 5.

Load reject results for the SLO point are shown in Table 2.1 while the accident results of T

the Pump Seizure event are presented in Appendix A.

3.2.1 Desian Basis The LRNB and FWCF transients have been determined to be most-limiting at i

and of full power capability when control rods are fully withdrawn-from the core. Between BOC and EOC-2000 mwd /MTV, the LFWH transient is most limiting.

From nominal E00-2000 mwd /MTU to EOC+30 EFPD, the LRNB and FWCF transients-1 remain limiting.'

The delta-CPR calculated for EOC-2000 mwd /MTV, E00, and

- E0C+30 EFPD is conservative for cases where control rods are partially inserted.

The analysis for Grand Gulf Unit I with ME00 was performed using conservative analytical limits for trips and setpoints.

Events initiated at core powers below 40% rated were analyzed with the direct scram due to turbine control and stop valve fast closure disallowed, -and with the recirculation i

pump high to low speed transfer selectively enabled for high dome pressure.

3.2.2 Anticinated Transients ANF's transient methodology report for jet pump BWRs. (Reference. 5) considered eight categories of anticipated transients.

The most limiting transients were evaluated at various power / flow points within ME00 to verify the. power dependent thermal margin for Grand Gulf Unit 1 Cycle 5.

The i

limiting transients analyzed for Grand Gulf Unit 1 Cycle 5 were:

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Loss of Feedwater Heating j

Load Rejection No Bypass Feedwater Controller Failure No Bypass

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ANF-90-021 Revision 1 Page 18 Other transients are inherently ~non-limiting or bounded by one of the above as shown in the NSSS vendor. ME00 analyses for Cycle 1 and the ANF Grand-Gulf Unit-1 Cycle 2 analyses.

Control Rod Withdrawal Error is an exception in that it has been analyzed. generically and reassessed using the ANFB Critical-Power Correlation and considering the ANF 9x9-5 fuel design.

3.2.2.l' Loss Of Feedwater Heatina Analysis of the loss of feedwater heating event was performed in Cycle 4 y

to reflect reactor operation over the ME00 operating power versus flow map and l

conditions anticipated during actual Grand Gulf reactor operation.

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Calculations performed for Cycle 5 assumed a conservative reduction of l -

100*F in the feedwater temperature. Table 3.1 provides the conditions of each 3

case analyzed.in terms of cycle exposure, core power, and core flow.

The initial and final NCPR values are presented for each case.

Analysis of the data showed that-the correlation between the initial and

- final MCPR-developed for Cycle 4 (Reference 18)- remained bounding

..This bounding relationship (Equation 3.2) is. presented in Figure 3.1.

_This relationship is repeated below as:

o MCPR(initial) = -0.0514 + 1.1130

(3.1)

L The operating limit MCPR was then shown to be defined by:

p OLMCPR(LFWH) = -0.0112 + 1.1130

Substituting the SLMCPR of 1.08, the L

MCPR operating limit for the LFWH cvent for all operating conditions analyzed is 1.19.

L 3.2.2.2 Load Re.iection No Bvoass The Load Rejection No Bypass (LRNB) event is the most limiting of the class of transients characterized by rapid vessel pressurization for Grand l

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ANF-90-021-L Revision 1 Page 19-Gulf Unit'1.

The load' rejection causes a fast closure of the turbine control u

L valves. - The resulting compression wave travels through the steam lines into the vessel and creates the rapid pressurization condition. - A reactor scram and a recirculation pump transfer are initiated by fast closure of the control valves.

Condenser bypass flow, which can' mitigate the pressurization j

j effect, is not allowed.

The excursion of the core power due to void collapse p

is primarily terminated by reactor scram and void-growth due' to the

- i recirculation pump high'to low speed transfer.

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Figures 3.2, 3.3, and 3.4 present the response of various reactor and plant parameters to the LRNB event initiated at the Reload Licensing Analysis condition (104.2% power /108% core flow).

Prior to EOC-2000 mwd /MTV, the LRNB p

is not the limiting transient. The MCPR operating limit of 1.29 is determined for EOC conditions. _

At E0C+30 EFPD, the MCPR operating limit at rated conditions is 1.30.

Table 2.1 lists the delta-CPRs-for' this transient at the1 power / flow conditions and exposure conditions considered.

1 3.2.2.3 Feedwater Controller Failure -

The failure of the feedwater controller'to maximum demand (FWCF) is the most-limiting of the vessel inventory. increase transients.

Failure of the

' feedwater control system to maximum demand would result in an increase in the coolant level in the reactor vessel.

Increased feedwater flow results in lower temperatures at the core inlet, which in turn cause an increase in-core

. power level.

If the feedwater flow stabilizes at the increased value, the core power will stabilize at a new, higher value.

If the flow increase continues, the water level in the downcomer will eventually reach the high t

level setpoint, at which time the turbine stop valve is closed to avoid damage to the turbine from excessive liquid inventory in the steamline.

The high m

- water level trip-also initiates reactor scram, and subsequent turbine ' trip-leads to recirculation pump high to low speed transfer.

The core power excursion is terminated by the same mechanisms that end the LRN8 transient.

Figures 3.5, 3.6, and 3.7 present the response of various reactor and plant parameters to the FWCF without bypass event initiated at' the Reload

-n

ANF-90 021 Revision 1--

Page 20 Licensed Analysis condition (104.2% power /108% core flow).

The delta-CPR for this event was calculated to be 0.13 at EOC. This delta-CPR is bounded by the LRNB delta CPR.

At E00-2000 mwd /Mid the delta-CPR for this event remains bounded - by the LFWH event delta-CPR of 0.11.

For the case of FWCF without bypass and with feedwater heaters out of service (-100 *F), the delta-CPRs-remain bounded by the FWCF without bypass at EOC.

3.2.2.4 Control Rod Withdrawal Error Reference 4 documents ANF's generic CRWE analysis for Grand Gulf Unit 1 operation within the ME00.

CRWE analyses were performed with MICR0 BURN-B-

.using 'ANFB critical power correlation.

Based on Reference 4 operating conditions and analytical procedures, one and two foot CRWE events were simulated.

The results of these analyses were statistically corchined to produce a 95/95 upper. limit for various power levels.

These analyses results are also demonstrated to bound Cycle 5 specific analyses.

As illustrated by Figure 2.2, the MCPR operating limit between 40% and 70% bounds the CRWE p

analyses results by a significant margin and therefore remains applicable for Cycle 5.

~3.3 Flow Excursion Analysis The flow excursion transient is analyzed to determine the flow dependent thermal limits and values (MCPRf and LHGRFAC ).

This transient is analyzed by f

assuming a failure of the recirculation flow control system such that the

. recirculation flow increases slowly to the physical maximum attainable by the-equipment. Two modes of operation are analyzed for Grand Gulf Unit 1 Cycle 5,

" loop manual" and "non-loop manual."

These two modes of operation correspond to -a single recirculation loop flow excursion event and a dual recirculation loop flow excursion event, respectively.

The results of the flow excursion transient analyses were used to establish new flow dependent thermal limits of MCPR.

Thus, the existing 7

limit (Reference 12) is being replaced.

For these analyses the change in critical power along the flow ascension path was calculated with XCOBRA e

ANF-90 021 Revision 1 Page 21 (Reference 8).

Peaking factors were selected such that' the bundle' with the least margin' would. reach the safety limit MCPR of 1.08 at the maximum flow.

Figure ' 2.4 presents the MCPR limits for maximum achievable core flows for f

both events,. conservatively assuming t%t u, res;rculation system equipment is capable of 110% of rated flow on the limiting rod line.

For flow rates less than 30% rated flow, the recirculatia system operates at low speed restricting the maximum possible flow.

Because of-this restriction, the MCPRf curve remains fixed between 20% flow and 30% flow.

The Cycle 5 LHGRFACf analysis was performed consistent with base line analyses performed in Cycle 2 and confirmed with CASM0/MICROBURN and the revised MCPR correlation ANFB.

The results of this analysis demonstrate the.

baseline analysis remains bounding and that the flow dependent multiplier (Figure 2.5) remains applicable.

Figure 2.5 shows the results of this.

analysis.

Because of restrictions in flow rates attainable for operation with core flows at less than 30% of rated, the LHGRFACf remains constant for core flow rates between 20%.and 30%.

3.4 Safety limit The safety limit MCPR is defined as the minimum value of the critical power ratio at which the fuel could be operated, with the expected number. of rods in boiling transition not exceeding 0.1% of the fuel rods in the' core.

The safety limit is the minimum critical power ratio which would be permitted to occur during the limiting. anticipated operational occurrence.

The safety limit MCPR for all fuel types in Grand Gulf Unit 1 Cycle 5 operation was

. calculated to be 1.08 using the methodology presented in-References 9 and-11.

The determination of the safety limit explicitly includes the effects of channel bow and relies on the following assumptions:

+

1.

Cycle 5 will not use channels for more than one fuel bundle lifetime.

2.

The channel exposure at discharge will not exceed 40,000 mwd /MTV based on the fuel bundle average exposure.

3.

The Cycle 5 core will contain GE and Cartech supplied channels.

e

L ANF-90 021 Revision 1 Page.22 4.

The-limiting module contains a

conservative exposure configuration (two twice-burned assemblies adjacent to a fresh assembly),

j The input parameter values for uncertainties used in the safety limit MCPR' analysis are unchanged from the Cycle 2 analysis presented in-Reference 2 except for the uncertainties associated with the new ANFB

- correlation, its implementation in the safety limit evaluation, channel bow, j

and the uncertainties appropriate for CASM0/MICR0 BURN analysis.

The. limiting local power distribution used to determine the safety limit MCPR is shown in Figure 3.8.

The effects of channel bow were modeled in the safety limit evaluation.

3.5 ' Su-arv of Results,

The results of the Grand Gulf Unit 1 Cycle 5 thermal limits analysis show a. Cycle 5 safety limit MCPR of 1.08 and a MCPR operating limit of 1.19 at rated conditions for E0C-2000 mwd /MTU.

A MCPR operating limit of 1.29 at rated conditions is shown at EOC.

For EOC+30 EFPD a MCPR operating limit 1.30 is supported.

These exposure dependent limits are shown in Figure 2.1.

The q

MCPR operating limit considers the effects of exposure (MCPR,), flow (MCPR ),

f and power (MCPR ). - The operating-limit of interest is the larger of the three p

values for a given reactor operating condition.

3.5.1 Power Denendent Thermal Limits and Values

'The power; dependent MCPR limit (MCPR ) protects against exceeding: the p

safety limit MCPR during anticipated operational occurrences from off-rated conditions.

The MCPR limit bounds the sum of the delta-CPR for the limiting p

event and the calculated safety limit MCPR.

-The power dependent LHGRFAC (LHGRFAC ) is used to protect against both p

fuel melting and 1% clad strain during anticipated system transients from off-rated conditions.

The conservative LHGR values for protection against fuel failure during anticipated operational occurrences are given in References 10 i

w

.ANF-90-021 Revision 1.

Page 23 and 13.

The results are presented in a fractional form for application to the LHGR operating limit.

The MCPR limits and LHGRFACp valuas for Cycle 5 are shown in Figures 2.2 p

and 2.3, respectively.

Between 40% and 70% power the MCPR limit is based on p

f the ANF.CRWE limit.

Below 40% power, the MCPR limit is either confirmed or

[

p revised based on Cycle 5 transient. analyses.

At rated power the MCPR limit p

is-also based on Cycle 5 transient analyses.

The Cycle 5 LHGRFACp values

- reflect the change to the off-rated LHGR multipliers methodology for Cycle 5.

h L

3.5.2 Flow Denendent Thermal Limits and Values i

The flow dependent MCPR limit (MCPR ) protects against exceeding the f

safety limit MCPR for slow flow excursion events.

The results of the' MCPRf analysis for Grand Gulf Unit 1 Cycle 5 are presented in Figure 2.4.

The flow dependent LHGRFAC (LHGRFAC ) protects against both fuel melting and 1% clad f

strain.

The LHGRFACf values to' be used in Cycle 5 are presented in Figure 2.5.

L i

j.

3.5.3 Exoosure Denendent Thermal Limits Th=. exposure dependent MCPR limit (MCPR,) protects..against exceeding the

. safety limii MCPR during the operation of the core.

The results of the 4

exposure dependent analysis for Grand Gulf Unit 1 Cycle 5 are ' presented in Figure 2.1.

l l

l l

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Q ANF-90 021 Revision 1 Page-24 Table 3.1 Grand Gulf Unit 1 Cycle 5 LFWH Data-Summary -

Initial State Final State Cycle Total' Core Total Core Core Total Cort ' Total Core Core Exposure Power Flow Minimum Power-Flow Minimum (GWd/MT)

(MWt)

(M1b/hr)

CPR (MWt)

(M1b/hr)

CPR 0.00 3833 97.87 1.37 4327 97.87 1.29 1.00 3833 96.75 1.34 4328 96.75 1.26 2.00 3833 96.75 1.34 4335 96.75 1.26

' 3.00 -

3833 95.62 1.33 4332 95.62 1.25 4.00 3833 96.75 1.31 4328 96.75 1.23 5.00 5833 96.75 1.30 4319 96.75 1.22 6.00 3833 95.06 1.29 4309 95.06 1.22 7.00 3833 95.62 1.32 4305 95.62 1.24 8.00 3833 96.19 1.31 4302 96.19 1.23 9.00 3833 97.87 1.33 4290 97.87 1.25 10.00 3833 99.00 1.36 4278 99.00 1.28 11.00 3833 108.00 1.39 4273 108.00 1.32 12.08 3833 118.12 1,46 4276 118.12 1,38 n

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-Figure 3.8 Grand-Gulf Unit 1 Cycle 5 Safety Limit Design Basis Local Power Distribution e

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4 ANF-90 021

-i Revision 1 Page 33 q

4.0 MAXIMUM OVERPRESSURIZATION Maximum system pressure has been calculated for the containment isolation event J (rapid closure of' all main. steam isolation valves) with an adverse i

scenario as specified in the ASME Pressure Vessel Code.

This analysis showed that the Grand Gulf Unit I safety valves have sufficient capacity and performance to prevent pressure from. reaching the established transient pressure safety limit 'of 110% of design pressure (1.1 x '1250 - 1375 psig).

The maximum. vessel-pressures at the most limiting power / flow point (104.2% power /108% flow) are shown in Table 2.1.

t 4.l' Qgsion Basis During the transient, the most critical active component (direct scram on L

MSIV closure) was assumed to fail.

The event was terminated by the high flux l-scram.

Credit was taken for actuation of only 13 of the-20 safety / relief valves:

6 in the relief mode' and 7 in the safety mode.

The calculation was performed with ANF's plant simulation code, COTRANSA2, which includes an axial one-dimensional neutronics model.

The safety valve analysis setpoints for this calculation included a conservative 6% tolerance.

Relief valve sctpoints for this analysis remain unchanged from Cycle 4.

4.2 Maximum Pressurization Transients 1-Scoping analyses described in Reference 5 found the closure of all main steam isolation valves (MSIVs) without direct scram to be limiting.

The MSIV closure was found to be limiting when all transients are evaluated on the same

' basis (without direct scram) because of the smaller steam line volume p

. associated with MSIV closure.

Though the closure rate - of the MSIVs is substantially slower than turbine stop or control valves, the compressibility

.of the additional fluid in the steam lines associated with a turbine isolation causes these faster closures to be less severe.

Once the containment is isolated, the subsequent core power production must.be absorbed in a smaller volume compared to that of a turbine isolation resulting in higher vessel pressures.

o 1

,j ANF-90 021 Revision l=

Page 34 i

4 '. 3 Results The results of the maximum system pressurization tr.alysis are presented j

in Table 2.1.

Figures 4.1, 4.2, and 4.3 present the response of various.

reactor and plant ~ parameters during the MSIV closure event from 104.2%

power /108% flow.

These results show that the Grand Gulf Unit I safety valvet l

have. sufficient capacity and performance to protect the previously established j

maximum vessel pressure safety limit of 1375 psig for Cycle 5.

Two state points were analyzed in order to cover the ME00 range for full power operation.

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s ANF 90 021 Revision 1 Page 38

5.0 REFERENCES

1.

Lester L. Kintner, USNRC, letter to 0. D. Kingsley, Jr., MP&L, " Technical Specification Changes-to Allow Operation with One Recirculation loop and Extended Operating Domain," August 15, 1986.

2.

" Grand _ Gulf Unit 1 Cycle 2 Plant Transient Analysis," XN NF-86-36',

Revision 3, Exxon Nuclear Company, Inc., Richland, WA, August 1986.

3.

" Grand Gulf Unit 1 Cycle 5 Reload Analysis,"- ANF-90-022, Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA, May 1990.

-4.

"BWR/6 Generic Rod Withdrawal Error Analysis; MCPRo for Plant Operations Within the Extended Operation Domain," XN-NF-825(P)(A), Supplement 2, Exxon Nuclear Company, Inc., Richland, WA, October 1986.

5.

"COTRANSA2:- A Computer ' Program for Boiling Water Reactor Transient Analysis," ANF-913, Volume 1, Supplements 1, 2, and 3.

6.

"XCOBRA-T:

A Computer Code for BWR Transient Thermal Hydraulic Core Analysis," XN-NF-84-105(P)( A), Volume 1, Exxon Nuclear Company, Inc.,

Richland, WA, February 1987.

7.

" Exxon Nuclear Methodology for Boiling Water Reactors:

Neutronics Methods for Design and Analysis,"

XN-NF-80-19( A),

Volume 1

and Supplement-3, Exxon Nuclear Company, Inc., Richland, WA, March 1983.

8.

" Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: -Thermal Limits Methodology. Summary. Description," XN-NF-80-19fP)(A), Volume : 3, Revision 2, Exxon Nuclear Company, Inc., Richland, WA, January 1987.

9.

" Advanced Nuclear Fuels Critical Power Methodology for Boiling Water

-Reactor," XN-NF-524(P), Revision 2, and Supplements, Advanced - Nuclear Fuels Corporation, Richland, WA, April 1989.

10. " Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"

XN-NF-85-67(P)(A), Revision 1, Exxon Nuclear Company, Inc., Richland, WA, September 1986.

11.

" Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, Inc., Richland, WA, June 1986, 12.

" Grand Gulf Nuclear Station Unit 1 Revised Flow Dependent Thermal Limits," NESDD-88-003, MSU System Services Inc., November 1988.

13.

" Generic Mechanical Design for Advanced Nuclear Fuels 9x9-5 BWR Reload,"

ANF-88-152, Amendment 1, September 1989.

14.

"ANFB Critical Power Correlation," ANF-1125(P), Supplement 1, April 1989.

j i

i l

ANF 90-021 Revision 1 Page 39

]

15.

"Granc Gulf 1 ANF-1.4 Design Report, Mechanical. Thermal Hydraulic, and Neutronic Design for Advanced Nuclear Fuels 9x9-5 Fuel Assemblies,"

ANF-89171(P), Volumes 1 and 2 January 1990, 16.

" Grand Gulf Unit 1 LOCA Analysis," XN NF-86-38, June 198o.

17.

" Acceptance for Referencing of Topical Report ANFll25(P) and Supplement 1,

'ANFB Critical Pow?

Correlation',"

Lr.tter from A. C. Thadani (NRC) to R. A. Copeland (AN.') March 8,1990.

18.

" Grand Gulf Unit 1 Cycle 4 Plant Transient Analysis," ANF 88 150, November 1988.

19.

"XCOBRA Code Users Nanual," XN-NF-CC-43, Revision 1. January 1980.

L v

9 4

1 L

ANF 90 021 Revision 1 Page A 1

)

i APPENDIX A SINGLE. LOOP OPERATION Analyses have been provided that demonstrate the safety of plant operation with a single recirculation loop out of service for an extended

)

period of time. These analyses confirm that during single-loop operation, the plant cannot reach the normal bundle power levels and nodal power levels that are possible when both recirculation systems are in operation.

The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power.

Because the ANF 9x9 5 fuel was 1

designed to be compatible with the co resident 8x8 fuel in thermal hydraulic,

?

nuclear, and mechanical design performance, and because the ANF methodology has given results which are consistent with those of the previous analyses for two loop operation, the analyses performed by the NSSS supplier for single-loop operation are also applicable to single loop operation with fuel and analyses provided by ANF.

o A.1 PUMP SEIZURE ACCIDENT The pump seizure i s a

postulated accident where the operating recirculation pump suddenly stops rotating.

This causes a rapid decrease in core flow, a decrease in the rate at which heat can be transferred from the fuel rods and a decrease in the critical power ratio.

COTRANSA2 and XCOBRA-T l

are used to calculate the MCPR for ANF fuel during a pump seizure from single-loop operation.

COTRANSA2 was used to simulate system response to a pump seizure in i

single-loop operation at the power flow point of 70.6% rated power and 54.1%

rated flow.

The operating recirculation pump rotor was stopped quickly j

causing a sudden decrease in the active jet pump drive flow.

During the

[

event, the inactive jet pump diffuser flow went from negative flow to positive flow.

Figures A.1, A.2, and A.3 show the graphical representation of important system parameters during the accident.

,--.-.n,-

nn.-. -- -,

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

,--,------n-

l ANF 90 021 Revision 1 Page A 2 4

Thermal hydraulic analysis using ANF safety limit methodology has shown that less than 10% of the rods in the core would experience boiling transition l

during this event.

Therefore.the two loop MCPR limit provides the required j

p proter.. tion below 70% of rated core power such that any postulated fuel failures would not result in exceeding a small fraction (<10%) of the 10CFR100 requirements.

i A.2 MCPR SAFETY LIMIT For single-loop operation ANF has determined that the two loop safety limit of 1.08 provides sufficient protection to account for increased tip uncertainties and increased flow measurerent uncertainties associated with single-loop operation.

ANF has evaluated the effects of these uncertainties using ANF safety lir.iit methodology and determined that the two loop safety L

limit MCPR is also applicable to ANF fuel during single loop operation for Cycle 5.

A.3 FLOW DEPDOENT THERMAL LIMITS It is conservative to use the reduced flow two loop operating MCPR limit for single-loop operations. The reduced flow MCPR limit is to protect against boiling transition during flow excursions to maximum flow.

The loop manual limit assures that there is even more thermal margin under si..gle-loop conditions than under two-loop condi'. ions.

A.4 MAPLHGR LIMITS ANF has established that the two-loop MAPLHGR limits for ANF 8x8 and 9x9 5 fuels multiplied by a reduction factor of 0.8 may be conservatively applied for single-loop operation.

Application of this reduction factor ensures that the peak clad temperature from a single-loop operation LOCA is bounded by the two loop LOCA analysis.

The application of these limits is valid for average planar burnups of 50000 mwd /MTV and 55000 mwd /MTU for ANF 8x8 and 9x9 5 fuels, respectively.

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ANF 90 021 Revision 1 Issue Date: 05/23/90 GRAND GULF UNIT 1 CYCLE 5 PLANT TRANSIENT ANALYSIS Distribution R. A. Copeland W. S. Dunnivant L. J. Federico N. L. Garner N. E. Garrett D. E. Hershberger N. J. Hibbard T. L. Krysinski R. B. Nacduff R. S. Reynolds S. E. State R. B. Stout C. J. Volmer G. N. Ward H. E. Williamson SERI/N. L. Garner (40)

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