ML20058E889
| ML20058E889 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/27/1982 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20058E885 | List: |
| References | |
| NUDOCS 8207300247 | |
| Download: ML20058E889 (31) | |
Text
_ - - - _ _ - _ _ _ _ _ _
c ATTACHMENT A Proposed Technical Specification Changes 8207300247 820727 PDR ADOCK 05000029 P
e Summary of Technical Specification Changes.
Delete the Following Pages Insert the Following Pages 2-2 2-2 2-3 2-3 B2-1 B2-1 B2-2 E2-2 3/4 1-1 3/4 1-1 B 3/4 1-1 B 3/4 1-1 3/4 1-7 3/4 1-7 B 3/4 1-2 B 3/4 1-2 3/4 1-23 3/4 1-23 3/4 1-29 3/4 1-25 P 3/4 1-5 B 3/4 1-5 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7
,3/4 2-7 B 3/4 2-2 E 3/4 2-2 B 3/4 2-1 B 3/4 2-1 5-1 5-1
e l
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660 I
640 w
.c 620
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y,AIS COOLAN' C
SYSTEM PPISSURE E.
600 Ey Cycle Dependent To be generated for each 3
reload cycle 3
580 2600 psia O
E 2400 psia S
E 560 2200 psia A
ov 540 2000 psia c7 z
1800 psia 520 I1600 psia 500 70 80 90 100 110 120 130 Indicated Reactor Power, Percent REACTOR CORE SAFETY LIMIT - ALL LOOPS IN OPERATION FIGURE 2.1-1 YANKIE-ROVE 2-2
660 640 o'
620 0
COOL W U
600 SYSTF.
c.
E Cycle Dependent p7yggggg e
To be generated for each g
reload cycle a
3 2600 psia O
E 400 psia 560 m
c 3
2200 psia ou 540 2
- E 2000 psia 520 1800 psia 500 (600 psia 50 60 70 80 90 100 110 Indicated Reactor Power, Percent REACTOR COPI SAFETY LIMIT - 3 LOOPS IN OPERATION TICURE 2.1-2 YANKEE-ROk'E 2-3
o 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the main coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation, and therefore, THERMAL POWER and main 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 uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30.
This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.
The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Main Coolant System pressure and cold leg temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. Because of flow instability, DNB may occur prematurely should the core exit quality become too great. The limiting core exit quality for preventing flow instability is taken conservatively as 0.08.
The values for these limits are calculated for each fuel cycle to assure a consistent margin to limits by accounting for variations in core power distributions from cycle to cycle. The cycle-specific limits are provided to the plant prior to startup. The basis for the calculation of the limits will be documented in the cycle-specific Core Performance Analysis Report.
The limiting hot channel factors used in determining the thermal limit curves are higher than those calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion.
l B 2-1
d SAFETY LIMITS BASES The cugves are based on the following nuclear hot channel factors:
l F of 2.76; FAH of 1.80; and a reference cosine with a peak of 1.44 for axial p6wer shape.
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion.
2.1.2 MAIN COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Main Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the main coolant from reaching the containment atmosphere.
The reactor pressure vessel, pressurizer and pumps are designed to Section VIII of the ASME Boiler and Pressure Vessel Code for Nuclear Power Plants, including all addenda through 1956, which permits a maximum transient pressure of 110 percent, 2735 psig, of design pressure. Pressure relief devices must be provided that will prevent pressure from exceeding 110 percent of the design pressure. The Main Coolant System piping and valves are designed to ANSI (formerly ASA) Standards, Power Piping Code, Section B31.1, 1955 Edition, and B16.5,1957 Edition, respectively, which allows the design to be based on normal operating pressure and temperature and also allows exceeding the design conditions for periods of time. The stress level can be increased 15 percent above the Code allowable design value for not more than 10 percent of the design life and up to 20 percent above the allowable for up to 1 percent of the design life. Since normal plant operating pressure is 2000 psig, there is no conflict with either design condition. The setting of the Main Coolant System safety valves could allow pressure to increase to 2560 psig during a transient. The amount of time this condition is expected to exist is well within the allowances of B31.1.
The Safety Limit of 2735 psig is, therefore, consistent with the design criteria and associated code requirements.
The entire Main Coolant System was hydrotested at 3435 psig,138 percent of design pressure, to demonstrate integrity prior to initial operation.
YANKEE ROWE B 2-2
e 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be > A A K/K, for Main Coolant Core Average Temperatures _> 5150F.
The SHUTDOWN MARGIN shall be j; B A K/K, for Main Coolant Core Average Temperatures < 4850F.
0 The SHUTDOWN MARGIN requirement is a linear function between 485 F and 5150F.
APPLICABILITY: MODES 1, 2* and 3.
ACTION:
With the SHUTDOWN MARGIN less than required, immediately initiate and continue boration at J: 26 gpm of 2200 ppm boron concentration or equivalent until the required SHUTDOWN MARGIN is restored.
SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determincd to be > that required:
Within one hour af ter detection of an inoperable control rod (s) and at a.
least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable. If the inoperable control rod (s) is immovable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).
b.
When in MODES 1 or 2#, at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.5.
c.
When in MODE 2##, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticality, by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.5.
d.
Prior to initial operation above 5 percent RATED THERMAL POWER af ter each fuel loading, by consideration of the f actors of e below, with the control banks at the maximum insertion limit of Specification 1
3 1.3.5.
- See Special Test Exception 3.10.1
- With Ke f f _,1.0
- With K gg 1.0 e
YANKEE R0WE 3/4 1-1
e 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, Main Coolant System boron concentration, and Main Coolant The most restrictive condition occurs at EOL, with T at System Tavg.
avg no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled Main Coolant System cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of BA k/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with accident analysis assumptions. The value of AA k/k is incorporated to provide added SHUTDOWN MARGIN and reflects the actual excess of shutdown margin available at the plant. With Tavg < 3300F, the reactivity transients resulting from a postulated steam line break cooldown are minimal.
5% A k/k SHUTDOWN MARGIN (with all rods inserted) provides adequate protection to preclude criticality for all postulated accidents with the reactor vessel head in place. The amount of SHUTDOWN MARGIN varies from fuel cycle to fuel cycle because of the changes in reactivity of the core from cycle to cycle. Thus, the values for A and B are fuel cycle dependent and are determined for each fuel cycle. The cycle-specific limits are provided to the plant prior to startup, and the basis for the calculation of the limits is documented in the cycle-specific Core Performance Analysis Report.
To eliminate possible errors in the calculations of the initial reactivity of the core and the reactivity depletion rate, the predicted relation between fuel burnup and the boron concentration, necessary to maintain adequate control characteristics, must be adjusted (normalized) to accurately reflect actual core conditions. Normally, when full power is reached af ter each refueling, and with the control rod groups in the desired positions, the boron concentration is measured and the predicted steady-state 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 af ter about 10% of the total core burnup. Thereaf ter, actual boron concentration can be compared with prediction and the reactivity status of the core can be continuously evaluated, and any deviation would be thoroughly investigated and evaluated.
B 3/4 1-1
e REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING C0tIDIT'_ON FOR OPERATION 3.1.1.4 The moderator temperature coefficient (NIC) shall be:
a.
Negative at hot zero power; b.
More negative than CA k/k/0F at RATED THERMAL POWER; and c.
Less negative than DA k/k/ F.
APPLICABILITY: MODES 1 and 2*#.
ACTION:
With the moderator temperature coefficient outside any one of the above limits, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.1.4.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.
4.1.1.4.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:
a.
Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
b.
When restarting from the first shutdown longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after
>60% of core life.
- With K gf _,1.0 e
- See Special Test Exception 3.10.4 YANKEE R0WE 3/4 1-7
a 3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 950 gpm provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Main Coolant System. A flow rate of at least 950 gpm will circulate an equivalent Main Coolant System volume of 2940 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron reductions will therefore be within the capability for operator recognition and control. Restriction of boron dilution with Main Coolant System temperature < 2500F prevents inadvertent criticality due to excess dilution below the temperature limit for criticality.
3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in Main Coolant System boron concentration associated with fuel burnup. The confirmation that the measured and appropriately compensated MTC value is within the allowable tolerance of the predicted value provides additional assurances that the coefficient will be maintained within its limits during intervals between measurement.
The values of the MTC (values C and D) are cycle dependent and are determined for each cycle. The cycle-specific limits are provided to the plant prior to startup, and the basis for the calculation of the limits is documented in the cycle-specific Core Performance Analysis Report.
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YANKEE ROWE B 3/4 1-2 l
a REACTIVITY CONTROL SYSTEMS i
3/4.1.3 MOVABLE CONTROL RODS CONTROL ROD OPERABILITY LIMITING CONDITION FOR OPERATION 3.1.3.1 All control rods which are inserted in the core shall be OPERABLE and positioned within + 8 inches (indicated position) of every other rod in their group.
APPLICABILITY: MODES 1* and 2*.
ACTION:
a.
With one or more control rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3 1 1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With more than one control rod inoperable or misaligned from any other rod in its group by more than + 8 inches (indicated position), be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c.
With one control rod inoperable or misaligned from any other rod in its group by more than + 8 inches (indicated position), POWER OPERATION may continue provided that within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:
1.
The rod is restored to OPERABLE status within the above alignment requirements, or l
2.
The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.
POWER OPERATION may then l
continue provided that:
l a) An analysis of the potential ejected rod worth is performed within 3 days and the rod worth is determined to be < E Ap at zero power and < hap at RATED THERMAL POWER for the remainder of the fuel cycle, and l
t
- See Special Test Exceptions 3.10.2 and 3.10.4 I
1 l
. YANKEE R0WE 3/4 1-23 l
100 80
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Cycle Dependent y
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'80 70 90 CONTROL R0D GROUP C POSITION (INCHES WITHDRAWN)
- Allowable THERMAL Power based on the main c'oolant pump combination in operation.
FIGURE 3.1-1 YANKEE-ROWE 3/4 1-29
___.__-.-y
-_,.__m
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL RODS (Continued) assurance of fuel rod integrity during continued operation. The reactivity worth of a misaligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumptions used in the accident analysis.
The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with Tavg 1.
515 F and with all main coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
The values of the ejected rod worth (E and H) and the control rod insertion curve (Figure 3.1-1) are cycle dependent and are determined for each cycle. The cycle-specific limits are provided to the plant prior to startup, and the basis for the calculation of the limits is documented in the cycle specific Core Performance Analysis Report.
YANKEE R0WE B 3/4 1-5 l
l t
POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) f.
Power level uncertainty, 1.03.
Heatfluxengineeringfactor,F(,1.04.
g.
h.
Core average linear heat generation rate at full power, G kw/f t.
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, YANKEE R0WE 3/4 2-3 i-
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1a 90-e CYCLE RVERAGE BURNUP (1000 MWD /NTU)
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GROUP C POSITION (INCHES WITHDRAWN)
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F 0 Measurement FIGURE 3.2-2 Factor F as a Function of Rod Insertion YANKEE R0WE 3/4 2-5
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CYCLE EXPOSURL ICWD/MfUI Figure 3.2-4 Multiplier for Reduced Power as a Function of Exposure a
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:
(a) maintaining the minimum DNBR in the core t 1.30 during normal operation and in short-term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.
3/4.2.1 PEAK LINEAR HEAT GENERATION RATE Limiting the peak linear heat generation rate (LHGR) during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 22000 exceeded.
F is not When operating at constant power, all rods out, with equilibrium xenon, power peaking in the Yankee Rowe core decreases monotonically as a function of cycle burnup.
This has been verified by both calculation and measurement on Yankee cores and is in accord with the expected behavior in a core that does not contain burnable poison. The all-rods out power peaking measured prior to exceeding 75% of RATED THERMAL POWER af ter each fuel loading thus provides an upper bound on all-rods out power peaking for the remainder of that cycle.
Thereaf ter, the measured power peaking shall be checked every 1000 equivalent full power hours and the latest measured value shall be used in the computation. The only effects which can increase peaking beyond this value would be control rod insertion and xenon transients and these are accounted for in calculating peak LHGR.
The core is stable with respect to xenon, and any xenon transients which may be excited are rapidly damped.
The xenon multiplier in Figure 3.2-3 was selected to conservatively account for transients which can result from control rod motion at full power.
i l
The limits on power level and control rod position following control rod insertion were selected to prevent exceeding the maximum allowable linear l
heat generation rate limits in Figure 3.2-1 within the first few hours following return to power after the insertion. With Yankee's highly damped core, the 24-hour hold allows sufficient time for the initial xenon maldistribution to accommodate itself to the new power distribution.
The restriction on control rod location during these 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the return to allowable fraction of full power will not cause additional redistribution due to rod motion.
(
YANKEE ROWE B 3/4 2-1 i
4 3/4.2 POWER DISTRIBUTION LIMITS BASES (Continued)
The value G and Figures 3.2-1, 2, 3, 4 are cycle dependent and must be determined for each cycle. The cycle-specific limits are provided to the plant prior to startup, and the basis for the calculation of the limits is documented in the cycle-specific Core Performance Analysis Report.
Af ter 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at zero power, the average xenon concentration has decayed to about 20% of the full power concentration. Since the xenon concentrations are so low, an increase in power directly to maximum allowable power creates transient peaking well below the value imposed by the xenon redistribution multiplier. Thus, any increase in power peaking due to this operation is below the value accounted for in the calculation of the LHCR.
These conclusions are based on plant tests and on calculations performed with the SIMULATE three-dimensional nodal code used in the analysis of Core XI (reference cycle) described in Proposed Change No.115, dated March 29, 1974.
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux and enthalpy hot channel factors ensure that
- 1) the design limits on peak local power density and minimum DNBR are not exceeded, and 2) in the event of a LOCA, the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit.
Each of these hot channel factors are measurable, but will normally only be determined periodically as specified in Specification 4.2.2.1 and 4.2.3.1.
This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:
Control rods in a single group move together with no individual rod a.
insertion differing by more than + 8 inches from any other rod in the group.
b.
Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.
f c.
The control rod insertion limits of Specification 3.1.3.5 is I
maintained.
N as a function of THERMAL POWER allows ghanges in The relaxation in FAH will be the radial power shape for all permissible rod insertion limits. FAH l
maintained with its limits provided conditions a through c above are l
maintained.
When an F measurement is taken, experimental error, engineering i
q tolerance and fuel densification must be allowed for.
5% is the appropriate allowance for a full core map taken with the incore detector flux mapping l
system, 4% is the appropriate allowance for engineering tolerance, and 3% is l
the appropriate allowance for fuel densification.
l YANKEE ROWE B 3/4 2-2 1
5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.
LOW POPULATION ZONE i
5.1.2 The low population zone shall be as shown in Figure 5.1-2.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel spherical shell having the following design features:
a.
Nominal inside diameter = 125 feet.
b.
Minimum thickness of steel shell = 7/8 inches.
c.
Net free volume = 860,000 cubic feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 3.4.5 psig and a temperature of 2490F.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 76 fuel assemblies with each fuel assembly containing nominally 230 or 231 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 91 inches. Each fuel assembly shall contain a nominal total weight of 234 kilograms uranium.
Reload fuel shall have an enrichment no greater than 4.0 weight percent U-235.
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YANKEE R0WE 5-1
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ATTACEME!;T B Cycle Dependent - Data 4
D 1his attaclment provides aQ example of the type of submittal' planned for future cycles under these proposed tect.nical specifications. The cycle specific parameters for the current operating cycle are provided for ease of
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i SAMPLE LETTER United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention:
Office of Nuclear Reactor Pegulation
References:
(a) License No. DPR-3 (Docket No. 50-29)
Subject:
50.59 Core Reload - Transmittal of Data
Dear Sir:
This letter transmits data associated with the next reload core (e.g.,
Cycle 15*) at the Yankee Nuclear Power Station (YNPS). The data is being provided for your information.
Pursuant to Section 50.59 of the Commission's Codes and Regulations, the Yankee Atomic Electric Company has performed an analysis of plant operation for the next reload core.
As a result of this analysis we have evaluated the adequacy of our existing Technical Specifications.
In addition, we have generated new cycle dependent parameters as identified in our Technical Specifications - items A-C and Figures 2.1-1, 2.1-2, 3.1-1, and 3.2-1 through 3.2-4.
Appropriate values for Core 15* are provided in the Attachment. A comparison to previous cycle values is also included. The cycle dependent parameters were generated with analysis methods used in previous core reload submittals and accepted by the NRC. A report has been published providing justification for operation of the next fuel cycle including each cycle dependent parameter. This report will be available at the plant site for NRC purposes.
We trust you will find this sutmittal satisfactory; however, should you desire additional information, please contact us.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY
- Core 15 is the present operating cycle and is being used here as an example.
CYCLE 15 PARAMETERS Value for Core XV Previous Cycle Reference to Core XV Core 1
Performance Analyses Report A
5.5%A K/K 4.72% A K/K
- p. 87 B
4.72% A K/K (See Note 1)
- p. 87 C
-0.2 x 10-4A K/K/0F
-0.2 x 10-4A K/K/0F
- p. 72-76, p. 87 D
-3.82 x 10-4 A K/K/0F
-3.82 x 10-4A K/K/0F
- p. 91 E
0.93% A p 0.93% Ap
- p. 95 H
0.50% Ap 0.50% Ap
- p. 95 G
4.40 KW/FT 4.40 KW/FT
- p. 51 NOTE 1:
The addition of shutdown margin vs. temperature requirements was first used in Core 15.
Previously, Modes 1, 2, and 3 were considered to be temperature independent, with only a 4.72% A K/K requirement.
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