ML14184A792

Forwards Pages Inadvertently Omitted from Bases Changes Submitted by Re Cycle 14 Fuel Reloads

ML14184A792
Person / Time
Site:	Robinson
Issue date:	02/04/1991
From:	Lo R Office of Nuclear Reactor Regulation
To:	Eury L Carolina Power & Light Co
References
NUDOCS 9102110109
Download: ML14184A792 (9)

Text

February 4, 1991 Docket No. 50-261 Mr. Lynn W. Eury Executive Vice President Power Supply Carolina Power & Light Company Post Office Box 1551 Raleigh, North Carolina 27602

SUBJECT:

INFORMATION SUPPORTING CYCLE 14 FUEL RELOADS - H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 Enclosed are copies of the Bases pages changed by letter dated December 13, 1990. The pages were inadvertently omitted from the previous letter.

Sincerely, Original Signed By:

Ronnie H. Lo, Senior Project Manager Project Directorate II-1 Division of Reactor Projects -

I/II Office of Nuclear Reactor Regulation

Enclosure:

As stated cc:

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Mr. L. W. Eury H. B. Robinson Steam Electric Carolina Power & Light Company Plant, Unit No. 2 cc:

Mr. R. E. Jones, General Counsel Mr. Dayne H. Brown, Director Carolina Power & Light Company Department of Environmental, P. 0. Box 1551 Health and Natural Resources Raleigh, North Carolina 27602 Division of Radiation Protection P. 0. Box 27687 Raleigh, North Carolina 27611-7687 Mr. H. A. Cole Special Deputy Attorney General Mr. Robert P. Gruber State of North Carolina Executive Director P. 0. Box 629 Public Staff -

NCUC Raleigh, North Carolina 27602 P. 0. Box 29520 Raleigh, North Carolina 27626-0520 U.S. Nuclear Regulatory Commission Mr. C. R. Dietz Resident Inspector's Office Manager, Robinson Nuclear Project H. B. Robinson Steam Electric Plant Department Route 5, Box 413 H. B. Robinson Steam Electric Plant Hartsville, South Carolina 29550 P. 0. Box 790 Regional Administrator, Region II U.S. Nuclear Regulatory Commission Mr. Heyward G. Shealy, Chief 101 Marietta Street Bureau of Radiological Health Suite 2900 South Carolina Department of Health Atlanta, Georgia 30323 and Environmental Control 2600 Bull Street Mr. R. Morgan Columbia, South Carolina 29201 General Manager H. B. Robinson Steam Electric Plant P. 0. Box 790 Hartsville, South Carolina 29550

Basis 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 maintaining the hot regions of the core 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 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, however, an observable parameter during reactor operation. Therefore, the observable parameters, thermal power, reactor coolant temperature and pressure, have been related to DNB through correlations. DNB correlations have 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, 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 allowable calculated value of the DNB ratio, DNBR, during normal operational transients and anticipated transients is restricted to the safety limit.

A DNB ratio at the safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur ad is chosen as an appropriate margin to DNB for all operating conditions.

The DNB ratio safety limit is a conservative design value which is used as a basis for setting core safety limits.

Based on rod bundle tests, no fuel rod damage is expected at this DNB ratio or greater.

For the standaf mixing vane fuel, the XNB correlation has a DNBR safety limit of 1.17.

For the High Thermal Perfo r~nce fuel, the ANFP correlation has a DNBR safety limit of 1.154.7 The curves of Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent the loci of points of thermal power, reactor vessel inlet temperature, and coolant system pressure for which the DNB ratio is not less than the safety limit.

The area where clad integrity is assured is below these lines.

The temperature limits at low power are considerably more conservative than would be required if they were based upon the minimum allowable DNB ratio but Changed by letter dated 2.1-2 December 13, 1990

are set to preclude bulk boiling at the vessel exit.

An arbitrary upper safety limit of 118% thermal power is shown.

This limit is based on the high flux trip including all uncertainties.

Radial power peaking factors consistent with the limit on FaH given in Specification 3.10.2.1 have been employed in the generation of the curves in Figure 2.1-1.

An additional heat flux factor of 1.03 has been included to account for fuel manufacturing tolerances and in-reactor densification of the fuel.

The safety limit curves in Figure 2.1-1 are based on a minimum RCS flow of 97.3 x 106 Ibm/hr. These curves would not be applicable in the case of a loss of flow transient. The evaluation of such an event would be based upon the analygis presented in Section 15.3 of the FSAR. The minimum RCS flow is 99.8 x 10 lbm/hr, which is the minimum thermal design flow of 97.3 x 10 Ibm/hr with a 2.6% allowance added for instrument uncertainty associated with the precision calorimetric flow measurement.

The Reactor Control and Protection System is designed to prevent any anticipated combination of transient conditions for Reactor Coolant.System temperature, pressure, and therm 4)power level that would result in a DNB ratio less than the safety limit 4based on steady state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less.than or equal to 575.4'F, and a steady state nominal operating pressure of 2235 psig.

Allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power, +40 F in Reactor Coolant System average temperature, and +/-30 psi in pressure.

The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions.

Reference (1) XN-NF-711(P) Rev. 0, "XNB Addendum for 26 Inch Spacer" (2) ANF-1224(P), "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel" (3) WCAP-11889, "RTD Bypass Elimination Licensing Report for H. B. Robinson Unit 2" (4) FSAR Section 15 Changed by letter dated 2.1-3 December 13, 1990

The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified set point assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.(6)

The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed set point above which these trips are unblocked assures their reliability in the power range where needed.

Above 10% power, an automatic reactor trip will occur if two reactor coolant pumps are lost during operation. Above 45% power, an automatic reactor trip will occur if any pump is lost.

This latter trip will prevent the minimum value of the DNB ratio, DNBR, from going below the safety limit during normal operational transients and anticipated transients when only two loops are in operation and the overtemperature AT trip setpoint is adjusted to the value specified for three loop operation.

The turbine and steam-feedwater flow mismatch trips do not appear in the specification, as these settings are not used in the transient and accident analysis.

(FSAR Section 15)

The RCS temperature measurement response time parameters define the time delay between when the OTAT reactor trip conditions are reached and when the control rods are released and free to fall and is based on a sensor lag of 4.0 seconds for the narrow range temperature measurement with a 0.75 second electromechanical delay.(7)(8)(9)

References (1) FSAR Section 15.4 (2) FSAR Section 15.0 (3) FSAR Section 15.6 (4) Deleted (5) FSAR Section 15.3 (6) FSAR Section 15.2.

(7) FSAR Section 7.2.2.2.2 Changed by letter dated 2.3-6 December 13, 1990

Specification 3.1.1.1.b requires that all three reactor coolant pumps be operating during power operation to provide core cooling in the event that a loss of flow occurs.

The flow provided will keep DNB well above the safety limit.

Therefore, cladding damage and release of fission products to the reactor coolant will not occur.

Specification 3.1.1.1.c is designed to allow for adequate mixing of the reactor coolant to maintain a uniform boron concentration during dilution, and to provide a means of boron injection.

Should no residual heat removal pump or reactor coolant pump be available, boration via natural circulation shall be initiated.

A boron concentration corresponding to 1% 6k/k at 200*F (which assumes most reactive rod stuck out) would prevent a return to criticality during the cooldown phase of the postulated steam line break event.

The pressurizer is of no concern because of the low pressurizer volume and because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant.

The purpose of Specification 3.1.1.1.d is to limit pressure surges exhibited in the RCS during a RCP startup.,These pressdre surges can be controlled in one of two ways.

One method would be to require a steam bubble in the pressurizer and thus control pressure using pressurizer controls.

The other method would be to limit the temperature difference (< 50F) between the RCS average temperature and the idle pump's cold leg water temperature.

3.1.1.2 Steam Generator At least two steam generators shall be operable whenever the average primary coolant temperature is above 350*F.

Basis One steam generator capable of performing its heat transfer function will provide sufficient heat removal capability to remove core decay heat after a normal reactor shutdown. The reactor cannot be made critical without water in all three steam generators, since the low-low steam generator water level trip prevents this mode of operation._ Two operable steam generators are therefore adequate.

Changed by letter dated 3.1-3 December 13, 1990

equilibrium conditions in terms of fueL loading patterns and anticipated control bank worths. These measurements will augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well as analytical basis.

Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.

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.

Second, the minimum DNBR in the core must not be less than the safety limit in normal operation or in short term transients.

In addition to the above, the initial steady state conditions for the peak linear power for a Loss-of-Coolant Accident must not exceed the values assumed in the accident evaluation. This limit is required in order for the maximum clad temperature to remain below that established by the ECCS Acceptance Criteria. To aid in specifying the limits on power distribution the following hot channel factors are defined.

a. FQ, Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

b.

F., Nuclear Heat Flux Hot Channel Factor, is defined as the maximum local fuel rod linear power density divided by the avcrage fuel rod linear power density, assuming nominal fuel pellet and rod dimensions.

c.

F, Engineering 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 Changed by letter dated 3.10-12 December 13, 1990

An upper bound envelope of peaking factors has been determined from extensive analysis considering all operating maneuvers consistent with the technical specifications on power distribution control as given in Section 3.10.2.

The specifications on power distribution control ensure that xenon distributions are not developed, which at a later time, could cause greater local power peaking even though the flux difference is then within limits. The results of a Loss-of-Coolant Accident analysis based on this upper bound envelope indicate that a peak clad temperature would not exceed the 22000F limit.

The nuclear analyses of credible power shapes consistent with the power distribution control procedures have shown that the F limit is not exceeded.

Q For transient events, the core is protected from exceeding 21.1 kw/ft locally, and from going below the DNBR safety limit by automatic protection on power, flux difference, pressure and temperature.

Measurements of the hot channel factors are required as part of startup physics tests and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors.

In the specified limit of F there is a 5 percent allowance for (5)

Q uncertainties(5 ) which means that normal operation of the core within the defined conditions and procedures is expected to result in a measured FN 5 Q

percent less than the limit, for example, at rated power even on a worst case basis. When a measurement is taken, experimental error must be allowed for, and 5 percent is the appropriate allowance for a full core representative map taken with the movable incore detector flux mapping system.

N In the specified limit of F.., there is an 8 percent allowance for design prediction uncertainties, which means that normal operation of the core is N

expected to result in FAH at least 8 percent less than the limit at rated power. The uncertainty to be associated with a measurement of F N by the movable incore system, on the other hand, is 4 percent, which means that the normal operation of the core shall result in a measured F N at least 4 percent AH less than the value at rated power. The logic behind the larger design uncertainty in the case is that (a) abnormal perturbation in the radial power N.

shape (e.g., rod misalignment) affects FAH in most cases without necessarily Changed by letter dated 3.10-16 December 13, 1990

DISTRIBUTION Docket-F-ie NRC PDR Local PDR PDII-1 Reading S. Varga (14E4)

G. Lainas E. Adensam P. Anderson R. Lo OGC D. Hagan (MNBB 3302)

E. Jordan (MNBB 3302)

G. Hill (4) (P1-137)

Wanda Jones (P-130A)

L. Kopp J. Calvo (1103)

ACRS (10)

GPA/PA OC/LFMB cc:

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