ML19093A499
| ML19093A499 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/27/1976 |
| From: | Virginia Electric & Power Co (VEPCO) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML19093A499 (42) | |
Text
EVALUATION.AND THE 1ECH SPECS FOR
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Date Recvd,
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-~ ~-______j RELOAD SAFETY EVALUATION AND CHANGES TO THE TECHNICAL SPECIFICATIONS FOR SURRY POWER STATION
. UNIT 1 -CYCLE 4 -.
SEPTEMBER 27, 1976 VIRGINIA ELECTRIC AND POWER COMPANY 9877
T... l\\BLE OF rnl'J'T'RN'T'!=:
SECTION 1 INTRODUCTION AND
SUMMARY
SECTION 2 REACTOR DESIGN SECTION 3 SAFETY EVALUATION SECTION 4 CONCLUSIONS SECTION 5 REFERENCES SECTION 6 CHANGES TO TECHNICAL SPECIFICATIONS
1.0 INTRODUCTION
AND
SUMMARY
This Reload Safety Evaluation (RSE) report presents a descrip-tion and the associated safety evaluation and changes to the Technical Specifications for the Cycle 4 reload core design of Unit 1 of the Surry Power Station.
Unit 1 is currently in its third cycle of operation and is projected to be refueled during October and November of 1976.
During the refueling, 97 fuel assemblies, including the two Region 4, 17x17 rod array demonstration assemblies, will be replaced with 84 fresh assemblies and 13 previously-burned assemblies.
Of the previously-burned assemblies, four assemblies were initially irradiated in Cycle 2 of Unit 2.
It is planned that Unit 1 will begin Cycle 4 operation by December 1, 1976, and that the cycle will extend until the spring of 1978, for a nominal 18-month cycle, producing approximately 13,200 MWD/MTU (9,299 EFPH) of energy.
All analyses performed for the Cycle 4 reload core design were based on the following assumptio~s:
- 1)
Cycle 3 operation is terminated after 8,800 +/- 1,000 MWD/MTU (6128 +/- 696 E~PH),
- 2)
Adherence to plant operating limitations which are deline-(1) ated in the approved Technical Specifications and the required changes to those Technical Specifications which are described in Section 6.0 of this report, and
- 3)
Cycle 4 operation will not exceed 14,200 MWD/MTU (10,003 EFPH).
All of the postulated accidents, except for the LOCA-ECCS accident, which (2) were analyzed and reported in the FSAR, that could potentially be affected by the Cycle 4 reload core design have been reviewed.
The LOCA-ECCS reanalysis is currently being performed and will be submitted as a supplement to this document.
The conclusions in the FSAR for the Control
e Rod Ejection Accident and the Steamline Break Accident were found to be potentially affected, and these accidents were reanalyzed.
The results of these analyses are included in Section 3.0 of this report.
It has been determined that the proposed Cycle 4 reload core will not adversely affect the safety of the station and, at the same time, will accomplish its energy generation requirement.
The Cycle 4 reload co.re will be operated in compliance with the results of the currently ongoing LOCA-ECCS reanalysis*, as will be documented in the supplement to this RSE report.
2.0 REACTOR DESIGN 2.1 Basic Design Parameters The basic design parameters for Cycle 4 are a core average power of 2441 Mwt, a system pressure of 2250 psia, a coolant flow rate of 2.31 x 10 6 lb/hr-ft 2, a reactor average temperature (TAVG) of 574.4° F, and a core (2) linear power density of 6.2 kw/ft.
2.2 Design Loading Pattern The fuel assembly loading pattern for the Cycle 4 reload core is shown in Figure 1.
The core loading will also contain 720 fresh borosili-cate burnable poison rods and 48 depleted burnable poison rods.
The burn-able poison rod locations and distributions are shown in Figure.2.
Also shown in Figure 2 are the location of two unirradiated secondary sources which will be activated during Cycle 4 operation.
It should be noted that all fuel in Cycle 4 will be in either its first or second cycle of operation.
2.3 Mechanical Desigp,_
The mechanical design of the fresh Region 6 fuel is the same as Region 5, but Regions 6A, 6B, and 6C have different enrichments as noted
e R
p N
6C
'6C 6C 6C*
6B 6C 6C 4C 6C.
4~
5 6C 6C 4C 6C 6B 6C 6C 6C 1-:1 FUEL lIBGION L:.J -DESIGNATION e
.. t d
M.
L 6C 6C 6C 4C.
4C 4C 4C 4C
- GA 4C 4C 4B.*
6A 4C 4C 4C 6A 4C 4C 4C 4C
. 6*c.
6C.
6C
_.,;:**-"'I,.*.,.
- FIGURE 1.
CORE LOADING.PATTERN*
SURRY UNIT 1*- CYCLE 4 IC.*.
J K
- G
. 6C 6C 6C 6C 6C 4B 6C 6B 4C 5
4C 4C 4C 4B 4C
- 6A.
4C 6A
- 4C 2/4A 6A 5
6A
- 6A 4C. 6A 4C
)
5 *.
(>A
- 1.
6A 6A 4C 6A 4e 2/4A 6A 5
6A.
6A 4C 6A 4C 4C 4C 4B 4C 6B 4C 5.
4C 6C 6C 4B 6C 6C 6C 6C F
E
. D
- C 6C 6C 6B 6C 6C 4C 4C 4C 6C 6A 4C 4°C 6C 2/4A 6A 4C 6B
- 6A 4C 4C 4C 5
6A 4B.
- 5 6A 4C 4C
'.:4c 2/4A 6A 4C
.6.B 6A 4C -~
~c 4C 4C 4C 6C 6B 6.C 6C 6C 6C
- B 6C 6C 6C
. 4B 6C 6C 6C l.
I
. I I*
i A
I 1
s 6C 7
6C 1
Ge**,
10
- 1).
1~
lS
B.
e
- ~
. J*
ss D
p 8
8 0
0 0
FIGURE 2
. SOURCE AND BURNABLE POISON ROD LOCATIONS AND DlSTRIBUTION SURRY UNIT 1.- CYCLE 4 N
H L
JC. *
- J K
c*
F E.
D 8
8 8
12 s
12 8
12*
8 12 12 8
8 8
12
- 8 12 8*
12 8
12 12 12
- 1.2 l,2 12 8
- 12 8
12 8
12 12*
12 8
8 12 12 8
12 8*
12 8
12 12 12 12 12 12 8
8 12 8
12 8
12
- 8.
8 12 12 8
8 12 s
12 8
12*
\\
8 8
0
~
- NUMBER OF l?R.ESH'?liURNMLK* POiSON RODS r-
- PRIMARY-SECONDARY SOURCE LOCATION c*
8 12 12*
- 12 8
.- DEPLETED ~URNABLE POISON ROD ASSEMBLIES
- FRESH SECONDARY SOURCE LOCATION
- B 8
8 i
r.
I A
ss I
1
- s 7
1.
10 lJ 13 1i lS
in Table 1.
In addition, 24 of the Region 5 assemblies, which were fabricated for use in Cycle 3, were held over to Cycle 4 for initial irradiation and have been denoted as Region 6A.
Clad flattening will not occur during Cycle 4.
Clad flattening time is predicted to be 30,730 EFPH for the limiting region, Region 1, using the current Westinghouse Evaluation Model.(3)
Region 1 had a Cycle 1 fuel residence time of 9,382 EFPH.
Therefore, Region 1 has a Cycle 4 allowed residence time of 21,348 EFPH.
Cycle 4 will nominally operate for 9", 299 EFPH.
Considerable experience with Zircaloy-clad fuel has been obtained, and this experience is extensively described in WCAP-8183. (4)
This report is updated periodically.
2.4 Thermal and Hydraulic Design The DNB evaluations for the Cycle 4 reload core were performed using the same models as were previously used.
The present DNB core limits were found to be adequate and conservative.
The potential effect of rod bow on DNB has been reevaluated based on the data and analyses given in Reference 5 and the interim guidance given by the NRC to Vepco on August 13, 1976.
The effect of rod bow on DNB has been accommodated by a Technical Specifications change (see Section 6.0) which reduces the limiting nuclear enthalpy rise hot channel factor (F~H) as a function of region average burnup.
This reduction in F~H is sufficient to accommodate the rod bow penalty on DNBR associated with the maximum expected end-of-cycle average burnup for each region in the Cycle 4 core.
2.5 Nuclear Design The Cycle 4 loading pattern results in a maximum analytically predicted FQ of 2.2 for routine steady~state and assumed load follow
Table 1 SURRY UNIT 1 -
CYCLE 4 FUEL ASSEMBLY DESIGN PARAMETERS*
Region 1
4B 4C
- 5.
6A 6B 6C S2/4A Enrichment (w/o U 235)*
- 1. 87 2.61 3.33 2.11 2.62 2.60 2.90 2.61 Density(% Theoretical)*
93.6 94.6 94.4 94.6 94.5 95.0 95.0 94.4 Number of Assemblies 1
8 52 8
24 8
52 4
Approximate Burnup at Beginning of Cycle 4 (MWD/MTU) 15,240 8,690 13,560 10,380 0
0 0
9:,890 Conservative Estimate of Burnup at End of Cycle 4 (MWD/MTU) 28,430 23,170 29,930 24,980 16,950 15,220 11,280 26,380
- All regions except Regions 6B and 6C are as-built values; Regions 6B and 6C reflect the nominal values; however, an average density of 94.5% theoretical was used in thermal evaluations.
The conserva-tive estimate of end-of-cycle burnup is based on Cycle 3 end-of-cycle burnup of 9,500 MWD/MTU and on Cycle 4 end-of-cycle burnup of 14,200 MWD/MTU.
- Fuel transferred from Region 4A of Surry Unit No. 2.
operating conditions (Condition I operation).
However, the actual Condition I maximum FQ will be redetermined routinely based on measured data obtained during cycle operation.
This process will be delineated in the Technical Specifications change which will be contained in the LOCA-ECCS supplement to ~Tuis RSE.
A comparison of Cycle 4 core kinetics characteristics with the current limits based on previously submitted accident analyses is pro-vided in Table 2.
It can be seen from Table 2 that all but one of the Cycle 4 parameters (i.e., the delayed neutron fraction) fall within the current limits.
A discussion of the effect of the changes in the value of this parameter is given in Section 3.0.
The beginning-and end-of-cycle control rod worths and shutdown requirements are provided in Table 3.
The minimum shutdown margin require-ment at the end-of-cycle condition is based on the results of a previously submitted design bases Steamline Break Accident analysis.
The avail-able shutdown margin associated with the Cycle 4 core exceeds (at all times during cycle operation) the minimum required shutdown margin.
How-ever, for the Cycle 4 core, there is a decrease in the integral return to power coefficient compared to the integral return to power coefficient limit used __ in _the_ abov_e _applicable Steainl~~e- ~:rea.:k "!-nal_y_sjs. __ Sip.c~~.h_e _effect of this change has not been analyzed and since the minimum required shutdown margin, as opposed to the available shutdown margin, is conserva-tively assumed in the analysis, a reanalysis of the hypothetical Steamline Break Accident is required.
The credible !Steamline Break Accident is also reanalyzed to maintain consistency with the above hypothetical Steamline Break reanalysis and to incorporate the impact of an upper reactor vessel
head fluid temperature assumed to be equal to 100% of the temperature in the hot loop.
The results of the reanalysis of both the credible and hypothetical Steamline Break Accidents are discussed in Section 3.0.
The allowable control rod insertion limits, as a function of power, to be used_during_Cyc~~ 4_o_peration ar~ tdentical to those used during Cycle.3 and are __ provided in_!~gure 3. __ The_ ~011:~lusions of applica_!>~_e __,,
safety evaluations are not affected by the Cycle 4 insertion limits, and all
- - - ~- --
power distribution limits are met even when accommodating the NRG-imposed interim thimble ce1-1--rod rise hot channel 3.0 SAFETY EVALUATION 3.1 General bow penalty on N
factor (Ftili)*
DNff tfirougli a reauction in the nuclear enthalpy This section provides, with the exception of the forthcoming LOCA-ECCS reanalysis, an evaluation of the impact of the Cycle 4 reload core on the design basis and postulated incidents previously analyzed in the FSAR or later safety evaluations.
This impact, which is discussed below, does not adversely affect the ability to safely operate the reactor at 100% of rated thermal power during Cycle 4.
A reload core can typically affect accident analyses input parameters in three major areas:
kinetics characteristics, control rod worths, and core peaking factors.
The Cycle 4 reload core parameters in each of these three areas were examined as discussed below to ascertain whether new accident analyses were required.
Kinetics Parameters A comparison of the value of the Cycle 4 kinetics parameters with current limits is given in Table 2. The moderator temperature coefficient will be
z 0 H E-1 H
Cf.l 0
Pol FIGURE 3 SURRY UNIT 1 -
CONTROL BANK INSERTION LIMITS 0
0.1 0.2, 0.3 I
0.4i 0.5 I
' ~
- 0. 6 I
- o. 7 '.;
t"'
111 11 I
[,
B -~
D+++-1-'.,+H+-++++-+-++++++--+++H-H+++-f-++++-+-H-++++-++-t, 1111 lf++++-+-H-++++-++-f+++-+-H-++-+-t-f-++++-+-H-++++-+-'°"-l-++++-+-H+++-f-++++-+-H-1-++-+-+-H+++-H-+++f+f-++-t-H-++++-H+-H-+H+f+++-+-D I
J 0.9
', 1,
' ' :-+/-t=
tel,,, '
1.0 I o.o 0.1 0.2 0.3 O.i 0.5 0.6 0.7 0.8 0.9 1.0.
FRACTION OF RATED POWER
r --
e e
e zero or negative during normal operation, although operation with a slightly positive coefficient is allowed below full power o.peration.
The minimum delayed fraction, S, for the beginning-of-cycle (BOC) condition is slightly lower than the current limit.
The impact of the slightly lower Sat the BOC conditions is addressed in Section 3.2.
Also, the return to power coefficient used in the._hypothetical _Steamliii_e Br_eak analysi~ -has decreased for Cycle 4.
A comparison of the coefficient with the current limit is provided in Table 5.
The impact of the reduced integral return to power coefficient is addressed in Section 3.2.
Control Rod Worth Changes in control rod worths may affect the shutdown ma_rgin, the_
maximum positive reactivity insertion rate, and the magnitude of the rod worths
,-=-==--- --=-=-
-==-----=----=--~ ---. --- -----
assumed in the safety analysis. Table 3 shows that the Cycle 4 reload co-Fe --~_hut-down margin is adequate.
As shown in Table 2, the maximum positiv~ rea~tivity insertion rate associated with the withdrawal of two RCCA.control bank~ movi_E-g together in their highest worth region for Cycle 4 is less than the current
-~-- --- ------
~ ---
\\ ---
limit value of 65 pcm/sec. Finally, the magnitude of all rod worths associated with the Cycle 4 core are conservative relative to those values assl!_med _j_JL previously suI?~itted safety evaluations for those accidents in the FSAR sensitive to rod worths.
Core Peaking Factors Core peaking factors during postulated abnormal conditions influence the maximum fuel rod centerline temperature and the initial stored energy in the fuel.
The maximum linear power density limit of 21.1 kw/ft for Over-power Accidents correspondsto the burnup dependent fuel centerline tempera-ture limit of 47000 F for the limiting regions, Region 6A, 6B, and 6C and
Table 2 SURRY UNIT 1 -
CYCLE 4 KINETICS CHARACTERISTICS Moderator Temperature Coefficient (pcm/°F)**
Least Negative Doppler - Only Power Coefficient, Zero to Full Power (pcm/percent power)
Delayed Neutron Fraction (percent)
Prompt Neutron Lifetime(µ sec)
Maximum Positive React,ivity Insertion Rate from Subcritical (pcm/sec)*
Current Limit( 2,lO)
+3.o* to -35
-11. 5 to -6~ 0
.60 to.50 26
- 65 Cycle 4
-0 to -35
-12.2 to -8.4
.59 to.50 19 63
- The moderator temperature coefficient may be positive up to full power according to the following program:
+3.0 pcm/°F from Oto 50% power and is linearly ramped down to 0.0 pcm/°F from 50 to 100% power (see Reference 1).
- pcm= 10-5 llK/K
Table 3 SURRY UNIT 1 -
CYCLE 4 SHUTDOWN REQUIREMENTS AND.MARGINS Control Rod Worth(% Ap)
All Rods Inserted Less Most Reactive Stuck Rod (1) Less 10% Uncertainty Control Rod Requirements (% ~p)
Reactivity Defects (i.e., Doppler, Moderator, Void, and Redistribution)(
Rod Insertion Allowance (2) Total Requirements Shutdown Margin * { (1)- (2)}. (% lip)
Required Shutdown Margin (% lip)
Cycle 4
- BOC EOG 7.51 6.76 1.83 0.90 2.73 4.03 1.00 7.99 7.19 2.83 0.90 3.73 3.46
- 1. 77
e is in excess of the maximum local linear power density calculated for these conditions during Cycle 4 operation.
The time dependent fuel densi-fication model was used.for this evaluation. (6)
The Cycle 4 peaking factor associated with the Rod Ejection Accident are less than those used in the previously submitted rod ejection safety evaluation.
Finally, the LOCA-ECCS limiting peaking factors are currently being reevaluated for compliance with the criteria delineated in 10 CFR 50.46.
This reevalua-tion is being performed with methods, which are in compliance with Appendix K, 10. CFR 50, and will incorporate the guidance.detailed in the 0 d (7}
NRC ) r er for Modification of License dated August 27, 1976.
If this analysis indicates that the core peaking factor produced during normal operation could exceed the above LOCA-ECCS limiting peaking factor values, appropriate restrictions on the power distribution limits and/or surveil-lance requi~ements on the power distribution would be required and will be submitted in the supplement to this RSE report.
3.2 Incidents Reanalyzed For the beginning-of-cycle Rod Ejection Accident cases, the minimum delayed neutron fraction, S, is below the current limit value.
However, the results.9_f __ the ___ ro_d _ej ec_t:j..on_ analysis ar_~dep~~~e_JJ.~_ on tli~
ejected rod reactivity expressed in units of dollars, which is defined as the (8) ratio of ejected rod worth to delayed neutron fraction.
For Cycle 4, ejected rod worths *.are lower than the current limits, and this more.than compensates for the slightly reduced delayed neutron fraction. Table 4 summarizes the results of this evaluation and shows that the ejected rod reactivity for Cycle 4 is less than the current limit.
Therefore, it is concluded that all safety criteria are met and that there is no requirement to:reanalyze the Rod Ejection Accident.
The hypothetical Steamline Break accident transients were reanalyzed due to a decrease in the integral return to power coefficient.
Table 4 ROD EJECTION PARAMETERS
- 1.
HZP -
BOC Max. Ejected Rod Worth (6K/K)
Delayed Neutron Fraction (Seff)
Ejected Rod Reactivity($)
- 2.
HFP -
BOC Max. Ejected Rod Worth (6K/K)
Delayed Neut~on F~action (Seff)
Ejected Rod Reactivity ($)
Current Limit
.0092
.0060 1.56
.00336
.0060
.56 Cycle 4
.0064
.0059
- 1. 28
.00326
.0059
.553
Four hypothetical _Steamline Break possible conditions are postulated.
The Case 1 - Complete severance of a pipe outside the containment, downstream of the steam flow measuring n0zzle, with the plant initially at EOL no load condition, full reactor coolant flow with offsite power available.
Case 2 -
Complete sevel7ance of a pipe.inside the containment at the outlet of the steam generator with the other conditions being the same as Case 1.
Case 3 -
Same conditions as Case 1 with the loss of 0ffsite Case 4 -
Same conditions as Case 2 with the loss of off site power.
power.
same design methods, as used in the previously submitted analysis~9) were used.
The reanalysis also included.the effects 0f a fluid tempera-ture in the upper reactor vessel head assumed to be equal to.100% ?f. tbe temperature in the hot loop.
The results are given in Table 5 for the limiting cases and show that for the hypothetical Steamline Break Accident the minimum l
DNBR is still greater than 1. 30.
For the credible Steamline Break Accident~ which
-* all safety -criteria are met and the conclusions of the* FSAR remain--val-+/-d-.-, ----.
4.0 CONCLUSION
S The Cycle 4 reload will not adversely affect the safety of the reactor while accomplishing the requirement for energy generation for up to 14,200 MWD/MTU.
This conclusion is based on the successful develop-ment of a core loading pattern which provides the necessary amount of reactivity while maintaining most reactor and safety design parameters within their current limits.
The impact on the safety of the reactor, except for the LOCA-ECCS analysis which is currently being reanalyzed,
e Table 5 STEAMLINE BREAK REANALYSIS PARAMETERS AND RESULTS A.
PARAMETERS:
- 1.
Integral.Return to Power Coefficient (pcm)
Relative Power (%)
Current Limit Cycle 0
0 0
10
-1380.
759.
20
-2150.
-1152.
30
-2460.
-1422.
B.
RESULTS:
Current Limit
- 1.
HyEothetical Break***
Peak Core Average Power, %
Reactor Inlet Temp., Failed Loop, Reactor Inlet Temp., Intact Loops, Reactor Coolant Pressure, psia Reactor *Coolant Flow, % of Nominal Min. DNBR
- 2.
Credible Break Reactivity at Closest Approach to Criticality, %LlK
- Inside Break {Case 2) with Power.
- Inside Break (Case 4) without Power.
15.8*
s.5**
OF 370 332 OF 497 519 794 875 100
- 24.
>1.3
>1.3
-.31 4
Cycle 4
- 28. 6*
9.2**
373 347 502 52i 1167 1215 100
- 23.
>1.3
>1.3
-.34
- While four cases are evaluated for the hypothetical Steamline Break Accident, as detailed in Section 3.2, only the two most limiting cases are reported.
,*!**,t' e
was assessed by reanalyzing those accidents for which the parameters for the Cycle 4 reload core design had exceeded their current limits. It was determined that the consequences of those accidents were well within appropriate limiting criteria.
As a result of the reload core design and safety analyses evaluation, administrative changes to the currently approved Technical Specifications have been proposed.
. 5. 0 REFERENCES
- 1.
Technical Specifications, Surry Power Station, Units 1 and 2.
- 2.
Final Safety Analysis Report, Surry Power Station, Units 1 and 2.
- 3.
R. A. George, et al, "Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-proprietary), July 1974.
- 4.
Placido, V. J., and Schreiber, R. E., "Operational Experience with Westinghouse Cores," WCAP-8183, Revision 3, May 1975.
- 5.
Letter from C. M. Stallings (Vepco) to B. C. Rusche (NRC), Serial Number 194, August 18, 1976.
- 6.
Hellman, J, M. (Ed.), "Fuel Densification Experimental Results and Model for Reactor Operation," WCAP-8218-P-A, March 1975 (Proprietary) and WCAP-,8219-A, March 1975 (Non-proprietary).
- 7.
Letter from Robert W. Reid (NRC) to W. L. Proffitt (Vepco), August 27, 1976 with Order for Modification of License dated Aguust 27, 1976.
- 8.
"An Evaluation of the Rod Ejection Accident in Westinghouse PWR's Using Spatial Kinetics Methods," WCAP-7588, Revision 1-:-A, January 1975.
- 9.
Letter from C. M. Stallings (Vepco) to K. R. Goller (NRC), Serial Number 458, March 12, 1975.
- 10.
"Fuel Densification Surry Units 1 and 2 Low Pressure Analysis,"
WCAP-8116 (Proprietary) and WCAP-8117 (Non-proprietary), April 1973.
e e
6.0 CHANGES TO TECHNICAL SPECIFICATIONS Based on the information and supporting analyses provided in the preceding sections of this document, changes to Section 2.1 (Safety Limit, Reactor Core), Section 3.12 (Control Rod Assemblies and Power Distribution Limits), Section 4.10 (Reactivity Anomalies!, and Section 5,3 (Reactor) of the Technical Specifications are provided in the following pages, The specific changes are concerned with (1) allowable.fuel residence time, (2) control rod insertion limits for banks C and D, (3) maximum linear power density limits, and (4) the enthalpy rise hot channel factor limit, F~H' The energy-weighted D bank insertion limit which was applicable for Unit 1, Cycle 3 operation is no longer necessary.
The specific changes to the Technical Specifications are identi-fied by a vertical line in the right margin.
The Technical Specifications Change Number appears to the right of the vertical line.for those items that are being changed as a result of this submittal.
An asterisk appears to the right of the vertical line for those items that could potentially change based on the results of the reanalysis of the LOCA-ECCS, which, as required by the **O:rd*er fo:!=' *Modd.if.ication of License issued by the NRC on August 27, 1976, is currently in progress.
The results of this reanalysis will be submitted to the NRC at a later date as a supplement to this docu-ment.
Until this is done, the heat flux hot channel factor limits, Fq, which have been changed from a value of 2.1 for both units are being limited to values of 1.80 for Unit 1 and 1,82 for Unit 2 as required by the Order.
e PROPOSED TS CHANGE NO. 47 SURRY UNITS 1 & 2 SEPTEMBER 24, 1976
e TS 2.1-2
- 4.
The reactor thermal power level shall not exceed 118% of rated power.
B.
The safety limit is expanded if the combination of Reactor Coolant System average temperature and thermal power level is at any time above the appropriate pressure line in TS Figures 2.1-1, 2.1-2 or 2.1-3; or the core thermal power exceed 118% of rated power.
C.
The fuel residence time. shall be limited to 21,348 effective full power hours(EFPH) for Cycle 4 of Unit 1 and to 6699 EFPH for Cycle 3 of Unit 2.
Basis 47 To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions.
This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the reactor coolant saturation temperature.
The upper boundary of the nucleate boiling regime is termed Departure From Nucleate Boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure.
DNB is not, how -
ever, an observable parameter during reactor operation.
Therefore, the observ-able parameters; thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 correlation.
The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially
e e
TS 2.1-4 than the loci of points of thermal power, coolant system average temperature, and coolant system pressure for which either the DNB ratio is equal to 1.30 or the average enthalpy at the exit of the core is equal to the saturation value.
At low pressures or high temperatures the average enthalpy at the exit:of the core reaches saturation before the DNB ratio reaches 1.30 and, thus, this arbitrary limit is conservative with respect to maintaining clad integrity.
The plant conditions required to violate these limits are precluded by the protection system and the self-actuated safety valves on the steam generator.
Upper limits of 70% power for loop stop valves open and 75% with loop stop valves closed are shown to completely bound the area where clad integrity is assured.
These latter limits are arbitrary but cannot be reached due to the Permissive 8 protection system setpoint which will trip the reactor on high nuclear flux when only two reactor coolant pumps are in service.
Operation with natural circulation or with only one loop in service is not allowed since the plant is not designed for continuous operation with less than two loops in service.
N TS Figures 2.1-1 through 2.1-3 are based on a F~H of 1.55, a 1.55 cosine axial flux shape and a DNB analysis as.described in Section 4. 3 of the report Fuel Densification Surry Power Station, Unit 1 dated December 6, 1972 (including the effects of fuel densification).
They are also valid for the following limit of the enthalphy rise hot channel factor:
F:H = 1.55 (1 + 0.2 (1-P)) x T(BU) 47 where Pis fraction of rated power and T(BU) is the interim thimble cell rod bow penalty on F~H given in TS Figure 3.12-9.
These hot channel factors are higher than those calculated at full power over the range between that of all control rod assemblies fully withdrawn to
e TS 2.1-6 to this limiting criterion. Additional peaking factors to account for local peaking due to fuel rod axial gaps and reduction in fuel pellet stack length have been included in the calculation of this limit.
The fuel residence time is limited to 21,348 EFPH for Cycle 4 of Unit 1 and to I 47 6699 EFPH for Cycle 3 of Unit 2 to assure no fuel clad flattening will occur in the cores without prior review by the Regulatory Staff.
References
- 1)
FSAR Section 3.4
- 2)
FSAR Section 3.3
- 3)
FSAR Section 14.2
e TS 3.12-3
- 7.
Delete B.
Power Distribution Limits
- 1.
At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:
For Unit 1 FQ(Z). < (1.80/P) x K(Z) for P > 0.5 FQ(Z)
< (3.60) x K(Z) for'P < 0.5 For Unit 2 F (Z)
< (1.82/P) x K(Z) for P>> 0.5 Q
FQ(Z)
< (3.64) x K(Z) for PK 0.5 For Unit 1 and 2
< 1.55 (1 + 0.2. (1-P)) x T (BU) where Pis the fraction of rated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8, Z is the core height location of FQ' and T(BU) is the interim thimble cell rod bow penalty on F~H given in TS Figure 3.12-9.
47 47
TS 3.12-4
- 2.
Prior to exceeding 75% power following each core loading, and during each effective full power month of operation thereafter, power distribution maps using the movable detector system, shall be made to confirm that the hot channel factor limits of this specification are satisfied.
For the purpose of this confirma-tion:
- a.
The measurement of total peaking factor, F~eas, shall be increased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.
- b.
The measurement of enthalpy rise hot :channel factor, F~H.*
shall be increased by four percent to account for measure-ment error.
If either measured hot channel factor exceeds its limit specified under 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under 3.12.B.1 are met.
If the hot channel factors cannot be brought to within the limits FQ ~ 1.80 x K(Z) for Unit 1 and Fq ~ 1.82 x K(Z) for Unit 2, and F~H ~ 1.55 x T(BU) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the overpower ~T and overtemperature
~T trip setpoints shall be similarly reduced.
147
TS 3.12-7 Alarms shall normally be used to indicate the deviations from the axial flux difference requirements in 3.12.B.4.a and the flux Difference time limits in 3.12.B.4.b. If the alarms are out of service temporarily, the axial flux difference shall be logged, and conformance to the limits assessed, every hour 'for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.
- 5.
The allowable quadrant to average power tilt is T = 2.0 + 50 (1.435 /F
- 1) 2,. 10%
xy where F is 1.435, or the value of the unrodded horizontal plane xy peaking factor appropriate to F as determined by a movable incore Q
detector map taken on at least a monthly basis; and Tis the per-centage operating quadrant tilt limit, having a value of 2% if F
is 1.435 or a value up to 10% if the option to measure F xy xy is in effect.
- 6.
If the quadrant to average power tilt exceeds a value T% as selected in 3.12.B.5, except for physics and rod exercise testing, then:
- a.
The hot channel factors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the power level adjusted to meet the specification of 3.12.B.1, or
- b.
If the hot channel factors are not determined within two hours, the power and high neutron flux trip setpoint shall be reduced from rated power, 2% for each percent of quadrant tilt.
- c.
If the quadrant to average power tilt exceeds+ 10% except for physics tests, the power level and high neutron flux trip setpoint will be reduced from rated power, 2% for each percent of quadrant tilt.
TS 3.12-13 malpositioned control rod assemblies are observable from nuclear and process information displayed in the Main Control Room and by core thermocouples and in-core movable detectors.
Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (part-length of full length control rod assembly 12 feet out of alignment with its bank) operation at":50% steady state power does not result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been performed.
An inoperable control rod assembly imposes additional demands on the operators.
The permissible number of inoperable control rod assmeblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
Two criteria have been chosen as a design basis for fuel performance related to fission gas release, pellet temperature and cladding mechanical properties.
First, the peak value of linear power density must not exceed 21.1 kw/ft for Unit 1 and 20.4 kw/ft for Unit 2.
Second, the minimum DNBR in the core must not be less than 1.30 in normal operation or in short term transients.
In addition to the above, the peak linear power density must not exceed the limiting kw/ft values which result from the large break loss of coolant accident analysis based on the ECCS acceptance criteria limit of 2200°F on peak clad temperature.
This is required to meet the initial conditions assumed for the loss of coolant accident. To aid in specifying the limits on power distribution the following hot channel factors are defined.
TS 3.12-14 F (Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the Q
maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods~
F~, Engineering Heat Fuel Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances.
The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad.
Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.
N F~H' Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
It should be noted that F:His based on an integral and is used as such in the DNB calculations.
Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core.
Thus the horizontal power shape at the point of maximum heat flux is not necessarily directly N
related to F6H.
An upper bound peaking factor envelope of 1.80 for Unit 1 and 1.82 for Unit 2 times the normalized\\,;
axial dependent of TS Figure 3.12-8 has been determined from extensive analyses considering all operating maneuvers consistent with the technical specifications on power distribution control given in Section 3.12.B.4.
The results of the loss of coolant accident analyses are conservative with
~espect to the ECCS acceptance criteria as specified in 10CFRS0.46.
TS 3.12-15 When an F measurement is taken, both experiemental error and manufacturing Q
tolerance must be allowed for.
Five percent is the appropriate allowance for a full core map (~ 40 thimbles monitored) taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerances.
N In the specified limit of Fb.H there is an eight percent allowance for uncertain-ties which means that normal operation of the core is expected to result in F~H
..::_ 1.55(1 + 0.2(1-P)) x T(BU)/1.08 where T(BU) is the interim thimble cell rod bow N
penalty on Fl>.H given in TS Figure 3.12-9.
The logic behind the larger uncertainty 47 in this case is that (a) normal perturbations in the radial power shape (e.g.
N rod misalignment) affect Fl>.H, in most cases without necessarily affecting FQ, (b) the operator has a direct influence on F through movement of rods, Q
and can limit it to the desired value, he has no direct control over FN
, and (c) an error in the predictions for radial power shape, which b.H may be detected during startup physics tests can be compensated for the N
FQ by tighter axial control, but compensation for Fb.H is taken, experi~
mental error must be allowed for and four percent is the appropriate allowance for a full core map (~ 40 thimbles monitored) taken with the movable incore qetector flux mapping system.
Measurement of the hot channel factors are required as part of startup physics tests, during each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors.
The incore map taken following core loading provides :confirmation of the basic nuclear design bases including proper fuel loading patterns.
The periodic incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would, otherwise, affect these bases.
TS 3.12-17 these hot channel factor limits are met.
In Specification 3.12.B.1, FQ is arbitrarily limited for P <
.5 (except for physics test~).
Delete
.47 Movable incore instrumentation thimbles for surveillance are selected so that the measurements are representative of the peak core power density.
By limiting the core average axial power distribution, the total power peaking factor FQ(Z) can be limited since all other components remain relatively fixed.
The remaining part of the total power peaking factor can be derived based on incore measurements, i.e., an effective radial peaking factor, R, can be determined as the ratio of the total peaking
TS 3.12-20 The technical specifications on power distribution control given in 3.12.B.4 assure that the F upper bound envelope of 1.80 for Unit 1 and 1.82 for Unit 2 Q
times Figure 3.12-8 is not exceeded and xenon distributions are not developed which at a later time, would cause greater local power peaking even though the flux difference is then within the limits specified by the procedure.
The target (or reference) value of flux difference is determined as follows.
At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full length rod control bank more than 190 steps withdrawn (i.e. normal full power operating position appropriate for the time in life, usually withdrawn farther as burnup proceeds).
This value, divided by the fraction of full power at which the core was operating is the full power va1ue of the target flux difference.
Values for all other core power levels.are obtained by multiplying the full power value by the fractional power.
Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of +6 to -9% ~I are permitted from the indicated reference value.
During periods where extensive load following is.re-quired, it may be impractical to establish the required core conditions for measuring the target flux difference every month.
For this reason, the specification provides two methods for updating the target flux difference.
Strict control of the flux difference (and rod position) is not as necessary during part power operation.
This is because xenon distribution control as part power is not as significant as the control at full 47 l
- e TS 3.12-22 as possible.
This is accomplished, by using the boron system to position the full length control rods to produce the required indicated flux dif-ference.
At the option of the operator, credit may be taken for measured decreases in the unrodded horizontal plane peaking factor, F xy This credit may take the form of an expansion of permissible quadrant tilt limits over tilt limits over the 2% value, up to a value of 10%, at which point specified power reductions are prudent.
Monthly surveillance of F
bounds the quantity because it decreases with burnup (WCAP-7912 L).
xy A 2% quadrant tilt allows that a 5% tilt might actually be present in the core because of intensitivity of the excore detectors for disturbances near the core center such as misaligned inner control rods and an error allowance.
No increase in F occurs with tilts up to 5% because misaligned Q
control rods producing such tilts do not extend to the unrodded plane, where the maximum F occurs.
Q
z O'.
H E-1 H.
(I) 0
-~
o.o 0.2 0.4 0.6
,~--
P=l 0.8 1.0 o.o TS FIGURE 3.12-lA
- r
- cf 0.2 0.4 0.6 0.8 FRACTION OF RATED POWER FIGURE 3.12-lA CONTROL BANK INSERTION-LIMITS FOR 3-LOOP NORMAL OPERATION-UNIT 1 1.0 47
e I
0.99 0.98
- o. 97 j *.
.0.94 0.93:
o o
+
5
,I-"
+
10 15 20 25 REGION AVERAGE BURNUP (1000 MWD/MTU)
N FIGURE 3.12-9 INTERIM THIMBLE CELL ROD BOW PENALTY ON F l1H
. SURRY UNITS Nd~ 1 AND 2 TS FIGURE.3.12-9 47 H
30,
35
e TS 4.10-1 4.10 REACTIVITY ANOMALIES Applicability Applies to potential reactivity anomalies.
Objective To require evaluation of applicable reactivity anomalies within the reactor.
Specification A.
Following a normalization of the computed boron concentration as a function of burnup, the actual boron concentration of the coolant shall be compared monthly with the predicted value. If the difference between the observed and predicted steady-state concentrations reaches the equivalent of one percent in reactivity, an evaluation as to the cause of the discrepancy shall be made and reported to the Nuclear Regulatory Commission per Section 6.6 of these Specifications.
B.
During period of power operation at greater than 10% of power, the hot N
channel factors, FQ and F~H' shall be determined during each effective full power month of operation using data from limited core maps.
If these factors exceed values of For Unit 1 FQ(Z) < (1.80/P) x K(Z) for P > 0.5 FQ(Z) < (3.60) x K(Z) for P < 0.5
- For Unit 2 FQ(Z) 2- (1.82/P) x K(Z) for P > 0.5
~(-Z~)-2.-E-3.--64-}-x-K:-E-Z--}-for--~-0-;-5 For Unit 1 and 2 F~H 2-1.55(1+0.2(1-P)) x T(BU) 47
e Basis TS 4.10-2 (where Pis the fraction of rated power at which ti.he core is operating, K(Z) is the function given in TS Figure 3.12-8, Z is the core he+/-ght location of N
FQ, and T(BU) is the interim thimble cell rod bow penalty on Ft.H given in TS Figure 3.12-9), an evaluation as to the cause of the anomaly shall be made.
BORON CONCENTRATION To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel 47 burnup and the boron concentration necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions.
When full power is reach initially, and with the control rod assembly groups in the desired positions, the boron concentration is measured and the predicted curve is adjusted to this point.
As power operation proceeds, the measured boron concentration is compared with the predicted concentration, and the slope of the curve*.relating burnup and reactivity is~compared with that predicted.
This process of normalization should be completed after about 10% of the total core burnup.
Thereafter, actual boron concentration can be compared with prediction, and the reactivity status of the core can be continuously evaluated.
Any reactivity anomaly greater than 1% would be un-expected, and its occurrence would be tho~oughly investigated and evaluated.
The value of 1% is considered a safe limit since a shutdown margin of at least 1% with:the most reactive control rod assembly in the fully withdrawn position is always maintained.
e e
TS 5.3;1 5.3 REACTOR Applicability Applies to the reactor core, Reactor Coolant System, and Safety Injection System.
Objective To define those design features which are essential in providing for safe system operations.
Specifications A.
Reactor Core
- 1.
The reactor core contains approximately 176,200 lbs of uranium dioxide in the form of slightly enriched uranium dioxide pellets.
The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods.
All fuel rods are pressurized with helium during fabrication.
The reactor core is made up of 157 fuel assemblies.
Each fuel assembly contains 204 fuel rods except for two demonstration fuel assemblies in Unit 2 which are part of Region 4 fuel.
The demonstration 147 assemblies each contain 264 fuel rods.
- 2.
The average enrichment of the initial core is 2.51 weight per
~C-ent-0-f---U-2--3-5-.~'I'-h-r--ee-f-u-el-enrichments-are---us-eu-in-ttn~-nut---ral core.
The highest enrichment is 3.12 weight per cent of U-235.
TS 5. 3-2
- 3.
Reload fuel will be similar in design to the initial core.
The enrichment of reload fuel will not exceed 3.60 weight percent of U-235.
- 4.
Burnable poison rods are incorporated in the inital core.
There are 816 poison rods in the form of 12 rod clusters, which are located in vacant control rod assembly guide thimbles.
The burnable poison rods consist of pyrex clad with stainless steel.
- 5.
There are 48 full-length control rod assemblies and 5 part-length control rod assemblies in the reactor core.
The full-length control rod assemblies contain a 144-inch length of silver-indium-cadmium alloy clad with stain-less steel.
The part-length control rod assemblies contain a 36-inch length of silver-indium-cadmium alloy with the remainder of the stainless steel sheath filled with Al2o3.
- 6.
Surry Unit 1, Cycle 4, Surry Unit 2, Cycle 3, and subsequent cores will meet the following criteria at all times during the operating lifetime.
- a.
Hot channel factors:
For Unit 1 FQ(Z) < (1,80/P) x K(Z) for P > 0.5 FQ(Z) < (3.60) x K(Z) for P ~ 0.5 For Unit 2 FQ(Z) < (1.82/P)x K(Z) for P > 0.5 FQ(Z) < (3.64) x K(Z) for P ~ 0.5 For Unit 1 and 2 F~H ~ i.55(1+0.2(1-P~ x T(BU) where P is the fraction of r.ated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8, Z is the core height of Fq, and T(BU) is the interim thimble cell rod bow penalty on F!H given in TS Figure 3.12-9.
47 47
B.
e TS 5.3-3
- b.
The moderator temperature coefficient in the power operating range is less than or equal to:
- 1) +3.0 pcm/°F at less than 50% of rated power, or
- 2) +3.0 pcm/°F at 50% of rated power and linearly decreasing to O pcm/°F at rated power.
- c.
Capable of being made subcritical in accordance with Specification 3.12 A.3.C
- 7.
Up to 10 grams of enriched fissionable material may be used either in the core of available on the plant site, in the form of fabricated neutron flux detectors for the purposes of monitoring core neutron flux.
- 1.
- 2.
The design of the Reactor Coolant System complies with the code requirements specified in Section 4 of the FSAR.
All piping, components, and supporting structures of the Reactor Coolant System are designed to Class 1 seismic requirements, and have been designed to withstand:
- a.
Primary operating stresses combined with the Operational seismic stresses resulting from a horizontal ground acceleration of 0.07g and a simultaneous vertical ground acceleration of 2/3 the horizontal, with the stresses maintained within code allowable working stresses.
- b.
Primary operating stresses when combined with~the Design Basis Earthquake seismic stresses resulting from a horizontal ground acceleration of 0.15g and a simulataneous vertical ground 47