ML20064C394
| ML20064C394 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/02/1994 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20064C396 | List: |
| References | |
| NUDOCS 9403100010 | |
| Download: ML20064C394 (20) | |
Text
-
1
$.pa %q(*-
2 E
UNITED STATES i.[i)#(
NUCLEAR REGULATORY COMMISSION
,/
t v
WASHINGTON, D.C. 20555-0001 y
VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 EURRY POWER STATION. UNIT NO. 1 MiEMMENT TO FACILITY OPERATING LICENSE Amendment No.189 License No. DPR-32 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company (the licensee) dated July 2, 1993, as supplemented December 10, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
i 9403100010 940302 DR ADOCK 05000280 p
PDR.
)
i i
1 l
l 1 l
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, l
and paragraph 3.B of Facility Operating License No. DPR-32 is hereby l
amended to read as follows:
(B)
Technical Soecifications The Technical Specifications contained in Appendix A, as revised l
through Amendment No.189, are hereby incorporated in the j
license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall
)
be implemented within 30 days, j
l FOR THE NUCLEAR REGULATORY COMMISSION I
l I
l l
I rbert N. Berkow, Director Project. Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of Issuance: March 2, 1994 l
l l
1
cp* * % 0.5 FO(Z) s (CFQ/0.5) x K(Z) for P s 0.5 where: CFQ = the FO limit at RATED POWER specified in the CORE OPERATING LIMITS REPORT, THERMAL POWER P=
,and RATED POWER K(Z) = the normalized FO limit as a function of core height, Z, as specified in the CORE OPERATING LIMITS REPORT FAH(N) s CFDH x (1 + PFDH x (1 P))
where: CFDH = the FAH(N) limit at RATED POWER specified in the CORE OPERATING LIMITS REPORT, PFDH = the Power Factor Multiplier for FAH(N) specified in the CORE OPERATING LIMITS REPORT, and THERMAL POWER p.
RATED POWER Provision for continued operation does not apply to Control Bank D inserted beyond the insertion limit.
Amendment Nos.189 and 189
i TS 3.12-4 2.
Prior to exceeding 75% of RATED POWER following each core
{
loading and during each effective full power month of operation l
thereafter, power distribution maps using the movable detector j
system shall be made to confirm that the hot channel factor limits of i
this specification are satisfied. For the purpose of this confirmation:
)
The measurement of total peaking factor F$* shall be 1
a.
l increased by eight percent to account for manufacturing l
tolerances, measurement error and the effects of rod bow.
The measurement of enthalpy rise hot channel factor Fh i
shall be compared directly to the limit specified in 1
)
Specification 3.12.B.1. If any measured hot channel factor
]
exceeds its limit specified under Specification 3.12.B.1, the j
reactor power and high neutron flux trip setpoint shall be
]
reduced until the limits under Specification 3.12.B.1 are met.
}
if the hot channel factors cannot be brought to within the l
FO(Z) and Fh limits as specified in the CORE OPERATING
]
LIMITS REPORT within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower AT and
)
Overtemperature AT trip setpoints shall be similarly reduced
]
within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
j b.
The provisions of Specification 4.0.4 are not applicable.
3.
The reference equilibrium indicated axial flux difference (called the target flux difference) at a given power level Po is that indicated
{
axial flux difference with the core in equilibrium xenon conditions l
(small or no oscillation) and the control rod assemblies more than j
190 steps withdrawn. The target flux difference at any other power level P is equal to the target value at Po multiplied by the ratio P/Po. The target flux difference shall be measursd at least once per equivalent full power quarter. The target flux difference must be l
updated during each effective full power month of operation either by actual measurements or by linear interpolation using the most q
recent value and the value predicted for the end of the cycle life.
The provisions of Specification 4.0.4 are not applicable.
i 4.
Except as mcdified by Specifications 3.12.B.4.a b, c, or d below, the indicated axial flux difference shall be maintained within a t 5%
band about the target flux difference (defines the target band on axial flux difference).
Amendment Nos.189 and 189
TS 3.12-16 In addition to the above, the peak linear power density and the nuclear enthalpy rise hot channel factor must not exceed their limiting values which result from the large break loss of coolant accident analysis based on the Emergency Core Cooling System 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 of power distribution, the following hot channel factors are defined:
Fo(Z), Height Deoendent Heat Flux Hot Channel Factor. is defined as the 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 tolerance on fuel pellets and rods.
F$, Engineerina Heat Flux 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 for non statistical applications.
F[g, Nuclear Enthalov 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 for both loss of coolant accident and non-loss of coolant accident considerations.
It should be noted that the enthalpy rise factors are based on integrals and are used as such in the DNB and loss of coolant accident calculations. Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in radial (x-y) power shapes throughout the core. Thus, the radial power shape at the point of maximum heat flux is not necessarily directly related to the enthalpy rise factom. The results of the loss of coolant accident analyses are conservative with respect to the Emergency Core Cooling System acceptance criteria as specified in 10 CFR 50.46 using the upper bound Fo(Z) times the hot channel factor normalized operating envelope given in the CORE OPERATING LIMITS REPORT.
When an Fo measurement is taken, measurement error, manufacturing tolerances, and the effects of rod bow must be allowed for. Five percent is the appropriate allowance for measurement error for a full core map (greater than or equal to 38 thimbles, including a Amendment Nos.189 and 189
TS 3.12-17 minimum of 2 thimbles per core quadrant, monitored) taken with the movable incore detector flux mapping system, three percent is the appropriate allowance for manufacturing tolerances, and five percent is appropriate allowance for rod bow. These uncertainties are statistically combined and result in a net increase of 1.08 that is applied to the measured value of Fo.
In the F$g limit specified in the CORE OPERATING LIMITS REPORT, there is a four l percent error allowance, which means that normal operation of the core is expected to result in dg s CFDH [1 + PFDH (1-P)]/1.04. The 4% allowance is based on the l considerations that (a) normal perturbations in the radial power shape (e.g., rod misalignment) affect F[g, in most cases without necessarily affecting Fo, (b) the operator has a direct influence on Fo through movement of rods and can limit it to the desired value; he has no direct control over F$g, and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests and which may influence Fo, can be compensated for by tighter axial control. An appropriate allowance for the measurement uncertainty for F$g obtained from a full core map (2 38 thimbles, including a minimum of 2 detectors per core quadrant, monitored) taken with the movable incore detector flux mapping system has been incorporated in the statistical DNBR limit.
Measurement of the hot channel factors are required as part of startup physics teste, 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 pattems. The periodic incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would, otherwise, affect these bases.
For normal operation, it has been determined that, provided certain conditions are observed, the enthalpy rise hot channel factor (y limit will be met. These conditions are as follows:
1.
Control rod assemblies in a single bank move together with no individual control rod assembly insertion differing by more than 15 inches from the bank demand position. An indicated misalignment limit of 13 steps precludes a control rod assembly misalignment no greater than 15 inches with consideration of maximum instrumentation error.
2.
Control rod banks are sequenced with overiapping banks as shown in the Control Bank Insertion Limits specified in the CORE OPERATING LIMITS REPORT.
l Amendment Nos.189 and 189
TS 3.12-18 4
3.
The full length Control Bank Insertion Limits specified in the CORE OPERATING LIMITS REPORT are not violated.
4.
Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits are observed. Flux difference refers to the difference between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and the bottom halves of the core.
The permitted relaxation in F$g with decreasing power level allows radial power shape changes with rod insertion to the insertion limits. It has been determined that provided the above conditions 1 through 4 are observed, this hot channel factor limit is met.
A recent evaluation of DNB test data obtained from experiments of fuel rod bowing in thimble cells has identified that the reduction in DNBR due to rod bowing in thimble cells is more than completely accommodated by existing thermal margins in the core design.
Therefore, it is not necessary to continue to apply a rod bow penalty to F$n.
The procedures for axial power distribution control are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers.
Basically, control of flux difference is required to limit the difference between the current j
value of flux difference (AI) and a reference value which corresponds to the full power equilibrium value of axial offset (axial offset - Al/ fractional power). The reference value of flux difference varies with power level and burnup, but expressed as axial offset it varies only with burnup.
The technical specifications on power distribution control given in Specification 3.12.B.4 together with the surveillance requirements given in Specification 3.12.B.2 assure that the Limiting Condition for Operation for the heat flux hot channel factor is met.
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 bumup proceeds). This value, divided by the fraction of full power at which the core Amendment Nos.189 and 189
TS 5.3-3 b.
The moderator temperature coefficient in the power operating range is less than or equal to the limits specified in the CORE OPERATING LIMITS REPORT.
The maximum upper limit for the moderator temperature coefficient shall be:
1)
+ 6 pcm/ F at less than 50% of RATED POWER, or 2)
+ 6 pcm/ F at 50% of RATED POWER and linearly decreasing to O pcm/ F at RATED POWER.
Capable of being made subcritical in accordance with Specification c.
3.12. A.3.C.
B.
The design of the Reactor Coolant System complies with the code requirements specified in Section 4 of the UFSAR.
l 2.
All piping, components, and supporting structures of the Reactor Coolant System are designed to Class 1 seismic requirements, and have been designed to withstand:
Primary operating stresses combined with the Operational seismic a.
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 simultaneous vertical ground Amendment Nos.189 and 189
i l
TS 6.2-1 l
6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS Specification A.
The following actions shall be taken for Reportable Events:
t 1.
A report shall be submitted pursuant to the requirements of Section 50.73 to 10 CFR, and 2.
Each Reportable Event shall be reviewed by the SNSOC The Vice President - Nuclear Operations and the MSRC shall be notified of the results of this review.
I B.
Immediate notifications shall be made in accordance with Section 50.72 to 10 CFR.
C.
CORE OPERATING LIMITS REPORT Core operating limits shall be established and documented in the CORE i
OPERATING LIMITS REPORT before each reload cycle or any remaining I
part of a reload cycle. Parameter limits for the following Technical Specifications are defined in the CORE OPERATING LIMITS REPORT:
1.
TS 3.1.E and TS 5.3.A.6.b - Moderator Temperature Coefficient 2.
TS 3.12.A.2 and TS 3.12.A.3 - Control Bank Insettlon Limits 3.
TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Umits Amendment Nos.189 and 189
TS 6.2-2 l l
l The analytical methods used to determine the core operating limits identified above shall be those previously reviewed and approved by the NRC, and identified below. The core operating limits shall be determined so that applicable limits (e.g., fuel thermal mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown l
l margin, and transient and accident analysis limits) of the safety analysis i
are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided for information for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
REFERENCES 1.
VEP-FRD-42, Rev.
1-A, "Reload Nuclear Design Methodology,"
September 1986 (Methodology for TS 3.1.E and TS 5.3.A.6.b - Moderator Temperature Coefficient; TS 3.12.A.2 and 3.12.A.3 - Control Bank Insertion Limit; TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor and Nuclear l
Enthalpy Rise Hot Channel Factor) 2a.
WCAP-9220 P A, Rev.1, " Westinghouse ECCS Evaluation Model-1981 Version," February 1982 (W Proprietary)
(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2b.
WCAP 9561-P-A, ADD. 3, Rev.1, 'BART A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients Special Report: Thimble Modeling in W ECCS Evaluation Model,' July 1986 (W Proprietary)
(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2c.
WCAP-10266-P-A, Rev. 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987 (W Proprietary) 1 (Methodology for TS 3.12.8.1 and TS 3.12.B.2 - Heat Flux Hot Chann i Factor) i i
Amendment Nos.189 and 189
TS 6.2 3 2d.
WCAP 10054 P-A, ' Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (W Proprietary)
(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2e.
WCAP 10079-P-A, 'NOTRUMP, A Nodal Transient Small Break and General Network Code," August 1985 (W Proprietary) 1 (Methodology for TS 3.12.8.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 3a.
VEP NE-2-A, " Statistical DNBR Evaluation Methodology,' June 1987 (Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor) l 3b.
VEP NE-3 A, " Qualification of the WRB-1 CHF Correlation in the Virginia I
Power COBRA Code," July 1990 (Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor) 4 l
Amendment Nos.189 and 189
.