ML18139C334

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Rev 1 to Reload Safety Evaluation for Surry 2 Cycle 7 (Pattern JB2)
ML18139C334
Person / Time
Site: Surry Dominion icon.png
Issue date: 03/31/1983
From: Cross R, Smith N, Suwal G
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18139C333 List:
References
282, 282-R01, 282-R1, NUDOCS 8305120250
Download: ML18139C334 (20)
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Text

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NFE TECHNICAL REPORT NO. 282-REVISION 1 RELOAD SAFETY EVALUATION FOR SURRY 2 CYCLE 7 (PATTERN JB2) edited by N. A. Smith NUCLEAR FUEL ENGINEERING POWER STATION ENGINEERING DEPARTMENT VIRGINIA ELECTRIC AND POWER COMPANY March, 1983 PAGE 0

Reviewed by: ~~'r;.,Jl,...p, )1. ~L (J

,,--Ef:305120250830509--

l PDR ADOCK 05000281

! p PDR I

I G. M. Suwal Approved by: ~i&} ~

R. W.

C:ross

TAB~E OF CONTENTS Title

1.0 INTRODUCTION

AND

SUMMARY

1.1 Introduction 1.2 General Description 1.3 Conclusions 2.0 REACTOR DESIGN 2.1 Mechanical Design 2.2 Nuclear Design 2.3 Thermal and Hydraulic Design 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION 3.1 Power Capability 3.2 Accident Evaluation 3.2.1 Kinetics Parameters 3.2.2 Control Rod Worths 3.2.3 Core Peaking Factors 3.3 Incidents Reanalyzed 4.0 TECHNICAL SPECIFICATIONS CHANGES

5.0 REFERENCES

e PAGE Page 3

3 3

4 5

5 5

6 "j

7 7

8 9

9 1 1 12 13

Table 1

2 3

4 5

Figu:z::e 1

LIST OF TABLES Title Fuel Assembly Design Pa:z::arnete:z::s Kinetics Cha:z::acteristics Shutdown Requi:z::ernents and Ma:z::gins Reanalysis Assumptions for the Rod Cluster Control Assembly Ejection Accident, BOL Hot Full Power Case Reanalysis Results fo:z:: the Rod Cluster Cont:z::ol Assembly Ejection Accident, BOL Hot Full Powe:z:: Case LIST OF FIGURES Title Core Loading Pattein JB2, -Su:z:::z::y Unit 2 crcle 7 PAGE 2

Page 1.4 15 16 17 18 Page 19

PAGE 3

1.0 INTRODUCTION

AND

SUMMARY

1.1 INTRODUCTION

This report presents an evaluation for Surry Power Station Unit 2, Cycle 7,

which demonstrates that the core reload designed.and designated as Pattern JB2 will not adversely affect the safety of the plant.

described This evaluation was accomplished utilizing the methodology in WCAP-9272, "Westinghouse Reload Safety Evaluation Methodology" (Ref. 1).

Based upon the above referenced methodology, only those.incidents analyzed and reported in the FSAR (Ref. 2) which could potentially be affected by the fuel reload have been reviewed for the C~cle 7 design described herein.

The results for the Hot Full Power Rod ejection event* at Beginning of Cycle were found to be potentially affected.

This event was reanalyzed, and the results of the reanalysis are discussed in Section 3.3.

1.2 GENERAL DESCRIPTION The Surry 2 reactor core is comprised of 157 fuel assemblies arranged 0

in the configuration shown in Figure

1.

During the Cycle 6/7 refueling, 61 fuel assemblies will be replaced with Region 9 fuel. The core loading pattern for Cycle 7 is shown in Figure 1. A summary of the Cycle 7 fuel inventory is given in Table 1.

__J

PAGE 4

Nominal core design parameters utilized for Cycle 7 are as follows:

Core Power CMwt)

System Pressure Cpsia)

Vessel Average Temperature C° F)

Thermal Design Flow Cgpm)

Average Linear Power Density Ckw/ft)

(based on hot, densi£ied core average stack height 0£ 143.6 inches)

1.3 CONCLUSION

S 2441 2250 574.4 265,500 6.2 From the evaluation presented in this report, it is concluded that the Cycle 7

design does not result in the previously acceptable safety limits being exceeded

£or any incident and consequently that no unreviewed safety questions exist as a result 0£ *this reload.

The conclusions are based on the following:

1. c~~le 6 burnups ranging from 14000 -

16800 MWD/MTU

2. Cycle 7 burnup will not exceed 15500 MWD/MTU (nominal end of reactivi~y plus ~pproximately 1500 MWD/MTU of coastdown)
  • 3.. There is adherance to plant operating limititions as given in the Technical Specifications

PAGE 5

2.0 REACTOR DESIGN 2.1 MECHANICAL DESIGN The mec~anical design of the Region 9 fuel assemblies is the same as the Region 8 assemblies.

Table 1 compares pertinent design parameters of the various fuel regions.

The Region 9 fuel has been designed according to the fuel performance model in Reference 3.

The fuel is designed* and operated* so that clad fla~tening will not occur, as by the Westinghouse model (Ref. 4).

For all fuel regions, predicted the fuel rod internal pressure design basis, which is acceptable as shown in Reference 5, is satisfied.

Westinghouse has had considerable experience with Zircalby clad fuel.

This experience is extensively described in WCAP-8183, "Operational Expe~ience with Westinghouse Cores" (Ref. 6). This report is updated annually.

2.2 NUCLEAR DESIGN The Cycle 7

core loading has a LOCA FQ limit of 2.18 under normal operating conditions.

The maximum analytically predicted FQ for Cycle 7

is 2.15; F2 is less than the limit at all core elevations for this cycle.* Therefore, frequent axial power distribution monitoring is not required.

PAGE 6

Table 2

provides a

summary of changes in the Cycle 7 kinetics characteristics compared with the current limits based on previously submitted accident analyses.

As shown in the table, only one of the Cycle 7 parameters, the most negative Doppler Temperature Coefficient, falls outside its current limit.

Section 3.0.

This parameter is evaluated in Table 3

provides the control rod worths and requirements at the most limiting conditions during the cycle.

The required shutdown margin is based on previously submitted accident analyses (Ref.

2).

The available shutdown margin exceeds the minimum required.

The control rod insertion limits applicable for Cycle 7 are those which are shown in Reference 7.

The loading contains a total of 900 fresh burnable poison reds located in 12 Region 8 and 57 Region 9 fuel assemblies.

There are no depleted burnable poison rods.

Three secondary sources will be used as shown in Figure 1.

2.3 THERMAL AND HYDRAULIC DESIGN No s~gnificant variations in thermal margins will result from the Cycle 7

reload.

The present DNB core limits (Re£. 7) have been found to be conservative for Cycle 7.

PAGE 7

3.0 POWER CAPABIL1TY AND ACCIDENT EVALUATION 3.1 POWER CAPABILITY The plant power capability is evaluated considering the consequences of those incidents examined in the FSAR (Ref. 2) using the previously accepted design basis.

It is concluded that the core reload will not a~versely affect the ability to safely operate at 100 percent of rated power during Cycle 7~

For the eualuation performed to address overpower concerns, the fuel centerline temperature limit of 4700°F can he accommodated with margin in the Cycle 7 core using the methodology discussed in Reference 1.

The time dependent densification model (Ref.

8) was used for these fuel temperature evaluations.

The LOCA limit at rated power can he met by maintaining F2 at, or below, 2.18.

3.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the FSAR (Ref. 2) were examined.

In most cases, it was found that the effects were accommodated within the conservatism of the assumptions used in the previously applicable safety analyses. For the incident which was reanalyzed, it was determined that the applicable design basis limits are not exceeded, and therefore the conclusions

_presented in*the FSAR are still valid.

A c6re reload can typically affect atcident analysis input parameters in the following areas=

core kinetic characteristics, control rod

e PAGE 8

worths, and core peaking £actors.

Cycle 7 parameters in each 0£ these three areas were examined as discussed below to ascertain whether new accident analyses were required.

3.2.1 KINETICS PARAMETERS A

summary. of the evaluation of Cycle 7 core physics parameters with current limits is given in Tabl~ 2.

The delayed neutron £ractions, moderator temperature coefficients, and prompt neutron lifetime are within the bounds of the current limits.

The moderator temperature coefficient will be zero or negative during normal operation, although operation with a

slightly positive coefficient is allowed below full power operation.

The most negative Doppler temperature coefficient is

-2.0 pcm/°F compared to the limit of -1.6 pcm/°F.. This coefficient is used in conjunction with *the Doppler power coefficient

£or fuel temperature changes in transients where the core water temperature drops.

For the most severe reactivity addition accident (startup of an inactive loop),

this amounts to less than a 37. increase in total positive reactivity insertion.

This would yield a negligible increase in peak power which can be accommodated in all of the FSAR c~oldown events.

--~

I=-=-'~-~----.===~=-=-'~---I-e PAGE 9

3.2.2 CONTROL ROD WORTHS Changes in control rod worth may affect differential rod worths, shutdown

margin, ejected rod
worths, and trip reactivity.

Table 2 shows that *the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 7 meets the current limit.

Table 3

shows that the Cycle 7

shutdown margin requirements are satisfied.

Ejected rod worths for Cycle 7 are within the bounds of the current limits.

As a

condition for using the rod swap technique for measuring rod

worths, the NRC has required that a

comparison be made between Westinghouse and Vepco shutdown margin calculations (Reference 9). This 1*

comparison will be perfoimed and reviewed by the Station N~clear Safety and Operating Committee and by the Safety Evaluation and Control Staff prior to Unit 2 startup.

3.2.3 CORE PEAKING FACTORS The peaking factors for the steamline break have been evaluated and are within the bounds of the previous safety analysis limits.

The peaking factors following control rod ejection are also within the limits of previous analysis

values, except at the hot full power condition at Beginning Of Life CBOL). The impact of this increase is discussed in Section 3.3.

As stated in Section 2.2, the maximum analytically predicted local peaking factor for Cycle 7

is less than the F2 limit.

Therefore,

e PAGE 10 frequent axial power distribution monitoring will not be required during Cycle 7.

e PAGE 11 3.3 INCIDENTS REANALYZED The Control Rod Ejection accident analysis is affected by an increase in power peaking factor for the BOL HFP case. This case was reanalyzed.

The hot*spot fuel rod and system parameters do not exceed the limiting criteria of the FSAR and Reference 10 for this case. Therefore, the conclusions of the FSAR remain valid. Reanalysis assumptions for the Control Rod Ejection accident are given in Table 4, and the results are presented in Table 5.

i*

-~-*-

e PAGE 12 4.0 TECHNICAL SPECIFICATIONS CHANGES No changes to the Unit 2 Technical Specifications are required as a result of the Cycle 7 reload.

e PAGE 13 5.*o REFERENCES

1. F. M. Bordelon, et al.," Westinghouse Reload Safety Evaluation Methodology," WCAP-9272, March 1978.
2. Surry P~wer Station Unit 1 and 2, Final Safety Analysis Report, December 1, 1969.
3. Miller, J. V., C Ed), "Improved Analytical Models used in Westinghouse Fuel Rod Design Computations", WCAP-87B5, October 1976.
4. George, R. A., et al.,"Revised Clad Flattening Model", WCAP-8377, July 1974.
5. Risher, D. H., et al.,"Safety Analysis ~or the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964, June 1977.
6. Jones, R. G., Iorii, J. A., "Operational Experience with Westinghouse Cores", WCAP-8183, Revision 11, May 1982.

1.

7. Surry Power Station Units 1 and 2, Technical Specifications, Docket Kos. 50-280 and 50-281, as amended.
8. Hellman, J. M., (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.
9. Letter from R. L. Tedesco CNRC) to W. N. Thomas CVepco), dated November 7, 1980.
10. Risher, D. H., "An Evaluation of ~he Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods", WCAP-7588, Rev. 1-A, January 1975.

\\

Table 1 Fuel Assembly Design Parameters Surry Unit 2 Cycle 7 Batch*

6B 7A 7B 8

9 Enrichment 3.20

3. 1 3 3.41 3.60 3.60 (LJ/0 U235)

Density 94.5 94.8 94.7 94.7 94.5 1 C~ Theoretical)

Humber o:f 8

8 12 68 6 1 Assemblies Burnup at 24400 23600 24400 15300 0

Beginning o:f Cycle 7 CMWD/MTU) 2 Burnup at 39700 38900 33900 30000 17600 End o:f Cycle 7 (MWD/MTU) 3 MTU per Region" 3.66 3.67 5.49

31. 21 28.00
1. Densities are as-built except :for Region 9, which is value..
2. Assumes end-of-Cycle 6 burnup o:f 14000 MWD/MTU.

3. Assumes end-of-Cycle 7 burnup o:f 15500 MWD/MTU; Batch batch average burnups do not exceed 37000 MWD/MTU.

4. Initial MTU a nominal 6 and 7
  • Table 2 Kinetics Characteristics Surry 2 Cycle 7 Parameter Moderator Temperature Coefficient Cpcm/°F)1 Most Negative Doppler-only Temperature Coefficient Cpcm/°F)

Least Negative Doppler~

only Power Coefficient, HZP to HFP (pcm/Yo power)

Minimum Delayed Neutron Fraction, BOL to EOL CY.)

Maximum Prompt Neutron Lifetime (micro sec)

Maximum Differential Rod Worth of Two Banks Moving Together (pcm/sec)

1. 1* pcm= 10-s d}Vk Z. S~e Section 3.2.1.

Current Limit 3.0 to -35.0

-1. 6

-11.4 to -6.00 0.55 to 0.44 26 75 e

PAGE 15 Cycle 7 Values within current limits

-2.0 2 within current limits within current limits 1*

<26

<75

e Table 3 Shutdown Requirements and Margins 1 Surry 2 Cycle 7 Control Rod Worths (7. dk/k)

All Rods Inserted less Worst Stuck RodC1)

C 1) less 10,. (2)

Control Rod Requirements (7. dk/k)

Reactivity Defects (combined Doppler, moderator, void, and £lux redist-ribution e££ects)

Rod Insertion Allowance fotal Requirements (3)

Shutdown Margin ((2)-(3))

Required Shutdown Margin

1. All BOC calculations based on EOC6=14000 MWD/MTU All EOC calculations based on EOC6=16800 MWD/MTU BOC 6.60 5.94
1. 85
1. 30
3. 15 2.79
1. 77 PAGE 16 EOC 7.44 6.70 2.99

.so 3.49 3.20 L77

===~---~

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PAGE

  • Table 4 REANALYSIS ASSUMPTIONS FOR THE ROD CLUSTER CONTOL ASSEMBLY EJECTION ACCIDENT BOL, HOT FULL POWER CASE 17 PARAMETER CURRENT ANALYSIS PREVIOUS ANALYSIS Power Level Ejected rod worth, 7. dk/k Delayed neutron fraction Feedback reactivity weighting Trip reactivity, 7. dk/k FQ before rod ejection FQ after rod ejection Moderator coefficient, pcrn/°F 102 0. 2 1

.0055

1. 30 4.00 2.55 6.50

+3~0 102 0.30

.0055

1. 30 4.00 2.55 5.46 1*

+3.0

C ---

  • Table 5 REAHALYSIS RESULTS FOR e

THE ROD CLUSTER CONTROL ASSEMBLY EJECTIOH ACCIDEHT BOL, HOT FULL POWER CASE

. PAGE PARAMETER CURREHT AHALYSIS Design Limit**

Maximum fue*1 avg temp,°F 4198 Maximum fuel centerline temp, or 5007 Maximum clad avg temp,°F 2543 2700 Maximum fuel stored energy, BTU/LB 332.7 360 Fuel Melting, 7.

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