ML14183A137

Safety Evaluation Supporting Amend 121 to License DPR-23

ML14183A137
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Site:	Robinson
Issue date:	01/09/1989
From:	Office of Nuclear Reactor Regulation
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 121 TO FACILITY OPERATING LICENSE NO. DPR-23 RELATING TO TECHNICAL SPECIFICATION FOR RTD BYPASS SYSTEM REMOVAL CAROLINA POWER & LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261

1.0 INTRODUCTION

By letter dated July 26, 1988 (Ref. 1), Carolina Power & Light Company (the licensee) indicated that the reactor coolant temperature measurement system for the hot and cold legs for the H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson), will be modified and requested changes to the plant's Technical Specifications. This modification eliminates the Resistance Temperature Detector (RTD) bypass manifold system and replaces it with dual element RTDs located directly in the reactor coolant system hot and cold leg piping. This will improve availability of the RTD, reduce radiation exposure, and reduce maintenance. The licensee also requested a reduction in the moderator temperature coefficient (MTC) when operating at or above 50% of rated power. In letters dated August 26 and November 1, 1988, the licensee provided additional information related to the proposed changes. The November 1, 1988 submittal also included additional restrictions in the TS to account for instrument response uncertainty and to require certain surveillance. This submittal was made at the request of the NRC staff.

2.0 BACKGROUND

2.1 Current Method The current method of measuring the hot and cold leg reactor coolant temperatures uses the RTD bypass system. This system was designed to address temperature streaming in the hot legs and, by use of shutoff valves, to allow replacement of the direct immersion narrow-range RTDs without draindown of the Reactor Coolant System (RCS).

For increased accuracy in measuring the hot leg temperatures, sampling scoops are located in each hot leg at three locations of a cross section, 120 degrees apart. Each scoop has five orifices which sample the hot leg flow temperature.

The flow from the scoops is piped to a manifold where a direct immersion RTD measures the average hot leg loop flow downstream of the steam generator.

The cold leg temperature is measured in a similar manner with piping to the bypass manifold, except that no scoops are used, as temperature streaming is not a problem due to the mixing action of the RCS pump.

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-2 2.2 Proposed Method The new method proposed for measuring the hot and cold leg temperatures uses narrow-range, dual element, fast response RTDs manufactured by Weed Instruments, Inc. One of each of the RTD dual elements is used while the other is installed as a spare. The RTDs are placed in thermowells to allow replacement without draindown. The thermowells, however, increase the response time.

The three RTDs in the hot legs of loops B and C are to be placed within the existing scoops for the Robinson plant. A hole will be drilled through the end of each scoop so that water will flow through the existing holes in the leading edge of the scoop, past the RTD, and out through the end hole. With the new method, the RTD measures the temperature at one point. This is in contrast to the temperature measurement of the average of the flow from the five sample holes from the hot leg scoops used in the RTD bypdss flow method. However, with the new method, each RTD measurement location is at the same radius as the center hole of the scoop. Therefore, the licensee states, it is the equivalent of the average scoop sample even if a linear temperature gradient exists in the pipe. The three RTDs in each hot leg and the single RTD used in each cold leg are used to calculate the reactor coolant loop differential temperature (delta-T) and average temperature (T ).

avg The RTDs in the hot legs for loop A will not have the three RTDs mounted in the existing scoops as described above due to structural interference.

Instead, these thermowells are to be located approximately 24 inches downstream of the existing scoop in the same circumferential spacing. The thermowells will be in independent bosses with one thermowell at the center and the other two 120 degrees to either side. Although these RTDs are not in scoops, the sensor will be at the same radial location as the other RTDs which are mounted inside the existing scoops. The licensee stated that the response time of the RTD in the hot leg free stream should, theoretically, be slightly faster than those located in the scoops. However, data from another plant indicates that this slight advantage is not reflected in a measured response time with respect to the scoop mounted RTD/thermowells. Since all three RTDs are being moved downstream together, the accuracy of their combined measurement will be the same as if they were all still in the scoops. The flow perturbation caused by the scoop upstream of the new RTD location will have dissipated prior to reaching the RTD. It is expected that this small perturbation will help mix the hot leg fluid and somewhat reduce the magnitude of the temperature streaming.

The design for the measurement of the cold leg temperature has also been modified. A single dual-element thermowell RTD will be mounted in each cold leg with one element as a spare. This is in place of the original method in which the measurement was by an external RTD in the cold leg bypass manifold. If the active cold leg RTD fails, then it is disconnected and the installed spare RTD lead is connected in the electronic cabinet in the failed RTD's place.

-3 The hot and cold leg RTD temperature measurements are entered into the calculations for the reactor coolant system delta-T and T temperatures.

Because of the variation in temperature in the cross sect iR of the hot legs due to streaming, the three RTD measurement locations in each hot leg are used to get an average value of the variation. An electronic based system will be used to perform the averaging of the reactor coolant hot leg signals from the three RTDs in each hot leg and transmit the average hot leg temperature signal to protection and control systems. There is a procedure for performing a quality check of the three temperature signals for each hot leg.

Capability exists to add a positive (or zero) bias to the averaging calculation, if needed, in order to compensate for the loss of one of the three hot leg RTD sensor inputs. The bias considers the past history of the previous readings. If two or more of the three hot leg dual elements RTDs or both cold leg RTD elements fail in the same protection channel, then that channel, is considered inoperable and placed in trip.

3.0 ANALYSIS The licensee responded to the NRC staff's questions regarding the response time and uncertainty effects of the new measurement system in a letter dated November 1, 1988 (Ref. 2).

The increased response time has the primary impact on the results of the accident analysis. The uncertainty of the hot leg temperature measurement also affects the accident analysis and is the principal contributor in the analysis for calculating the RCS flow measurement uncertainty.

3.1 RTD Response Time The overall response time of the new thermowell RTD hot leg temperature system as presented in the licensee's July 26, 1988 submittal (Ref. 1) is of the same order as the existing RTD bypass system overall response time of 5.0 seconds. While performing analysis in support of a planned modi fication to eliminate the RTD bypass system the licensee had found an error in the original RTD response time used in the accident analysis (LER 88-002). The correct response time of 5.0 seconds was greater than that originally assumed (Table 15.0.6-1 of the UFSAR shows a value of 2.30 seconds for the delay time of the overtemperature delta-T trip),

necessitating a revision to the accident analysis. The revised analysis provided was-for the new thermowell RTD with an increased overall response time of 4.75 seconds. Because of the increased response time, there are longer delays from the time when fluid conditions in the RCS require an overtemperature delta-T (OTDT) or overpower delta-T (OPDT) reactor trip until a trip is actually generated. The licensee does not take credit for the OPDT trip in the accident analysis. The licensee presented information in Reference 3 concerning the FSAR Chapter 15 non-LOCA accidents that rely on the OTDT reactor trip and which were evaluated for the longer response time.

-4 In the submittal of November 1, 1988 (Ref. 2), the licensee proposed to revise the affected Technical Specifications for the new RTD response time. The licensee stated that the 5.0 second overall response time for the existing design consisted of 0.5 seconds for the RTDs and thermowells and 4.5 seconds for RTD bypass transport, thermal lag in piping, filter and electronic Gelays. The revised safety analysis assumed a 4.75 second overall response time for the new thermowell RTD. This is composed of 4.0 seconds for RTD/thermowell lag time and 0.75 seconds for electronics delay. The licensee stated that this reponse time is achievable with some margin; as the slowest RTD/thermowell response time measured in factory tests ranged from 3.7 to 2.7 seconds.

As noted in NUREG-0809 (Reference 5), extensive RTD testing has revealed RTD time response degradation with aging. In view of this, surveillance tests are needed. The licensee provided a revision to Technical Speci fication Table 4.1-1 which includes the surveillance test schedule for RTD response time. The required RTD response time is given in Technical Specification Section 2.3.3. The NRC-approved in-situ method for measuring RTD response time is the Loop Current Step Response (LCSR) method, which is the method to be used at Robinson.

3.2 RTD Uncertainty The new method of measuring each hot leg temperature with three thermowell RTDs manufactured by Weed Instruments, Inc., in place of the RTD bypass system with three scoops, has been analyzed by the licensee. The new method measures at one point for each scoop center hole location, compared to the former method in which there werefive sample points in a 5-inch span of the scoop measurements. This may result in a small streaming error relative to the former scoop flow measurement because of a temperature gradient over the 5-inch scoop span. However, this gradient has been calculated to have only a small effect. Also, since possible temperature uncertainties from imbalanced scoop flows are eliminated, the overall result is more accurate and is within the bias shift allowance in the analysis. In addition, since the new methoo uses three RTDs for each hot leg temperature measurement, it is statistically a more accurate temperature measurement than the former method which used only one RTD for each hot leg temperature measurement. Additional uncertainties are introduced when the signals are processed for averaging before being sent to the processing system.

The licensee stated that sufficient allowance has been made in the reactor protection system setpoints. Therefore, the current values of nominal setpoints in the Technical Specifications are still valid.

In the proposed new configuration T is first calculated by averaging three (3) independent RTD measuremenW!. Since the contribution of each RTD is divided by three in this averaging process, a single T RTD will have to shift three times the original four degrees (12 degreP2tF) to trigger an alarm. This T signal is then used to calculate T in the same manner as before. EM though a single RTD must shift furf r to trigger an alarm, the controlling function (T ) has still only shifted 2 degrees F.

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The same logic is applied to the delta-T deviation alarm. Again the influence of a single T RTD is only one third of the total. Therefore, a single T RTD must 9H ft 6 degrees before the 2 degree deviation alarm is trigger H. A channel check is performed on a shift basis. The temperatures are recorded on a monthly basis and the records are available to supply a bias to compensate for the loss of one of the hot leg RTD sensor inputs as discussed in Section 2.0. Upon loss of one hot leg RTD, the preferred method for correction is to substitute the spare sensor of each dual element RTD by switching leads in the electronic cabinet.

The licensee states that for re-calibrating the RTDs at each refueling prior to startup, they use the Westinghouse recommended RTD cross calibration method. Also, there is a small allowance in the OTDT reactor trip analysis to account for RTD delta-T drift. The platinum resistance temperature sensors are believed to be very stable and to have relatively small calibration drifts. However, according to several sources (Refs.

5, 6, 7, 8), RTDs have been known to shift in calibration with time and corrections were necessary. The licensee addressed the ability to check the accuracy of the new thermowell RTD measuring method, with the RTD bypass system removed, as compared to the previous method.

The licensee stated that they will perform a comparison of the delta-T temperature indications after the modification with measurements prior to the modification. The NRC will be notified of the results of this comparison.

3.3 Flow Measurement Uncertainty The thermal design flow (TDF) for the Robinson plant is 97.29x10 6 lb/hr or 258,900 gpm based on an inlet temperature of 550.20 F. From the results of the lateat precision calorimetric RCS measurement, the indicated flow is 103.8x10 lb/hr or 273,400 gpm. The current flow measurement uncer tainty (FMU) is 1.87%. Therefore, the cogesponding minimum flow, including flow measurement uncertainty, is 101.9x10 lb/hr or 270,250 gpm. These values are well above the TDF. The licensee provided a flow measurement uncertainty analysis in References 2 and 3 that accounts for changes due to the RTD bypass system removal. The licensee stated that the methodology used was the same as that used for the Shearon Harris Unit 1 plant (Ref.

13) and is consistent with NUREG/CR-3659 (Ref. 14). There is a statistical advantage to using three RTDs for the hot leg temperature measurement in the new method rather than the one RTD in the former method. This analysis used the plant-specific instrumentation for-the Robinson plant. The results of the analysis indicated that the flow-measurement uncertainty value is 2.6% (including the cold leg elbow taps and 0.1% for feedwater venturi fouling).

With the modifications due to the RTD bypass system removal, the new flow measurement uncertainty value of 2.6% will be used in place of the former value of 1.87%. This FMU is used for calculating the RCS flow from the

-6 precision calorimetric measurement for comparison against the TDF. The minimum allowable flow (MAF) is obtained from the following equation:

MAF = TDF * (1+FMU)

For the Robinson plant, MAF= 99.8 x10 6 lb/hr.

The licensee has committed to have a procedure to require cleaning the feedwater venturi meters at each refueling before the precision calori metric is made. We have found the flow measurement uncertainty analysis to be acceptable.

3.4 Positive Moderator Temperature Coefficient The licensee requested a change in the moderator temperature coefficient (MTC) as reflected in the accident analysis. Originally, Technical Specification 3.1.3, Mininmum Conditions for Criticality, stated that "the reactor shall not be made critical at any temperature above which the MTC is greater than + 5.0 pcm/oF at 50% of rated power and linearly decreasing to 0 pcm/oF at rated power."

This has been changed to state that "The reactor shall not be made critical at any temperature above which the MTC is greater than 0 pcm/oF at 50% of rated power and above." This is acceptable as it is a change in the conservative direction.

3.5 Non-LOCA Accidents Reanalyzed The licensee stated that the primary impact of the RTD bypass system elimination is the different response time characteristics of fast response thermowell RTDs. Thus, only those events which rely on the OTDT and OPDT reactor trips are impacted. The Robinson plant does not take credit for the OPDT trip in the accident analysis.

The Chapter 15 accidents in the FSAR were examined by the licensee and the following non-LOCA accidents affected by the longer response time were reanalyzed: (1) Loss of External Electrical Load/Turbine Trip; (2) the Uncontrolled Rod Cluster Control Assembly (RCCA) Withdrawal at Power; and (3) Control Rod Misoperation. The licensee stated that the approved plant transient computer code PTSPWR2 (Ref. 9) was used for the analysis of these events. Its output is used as input to the approved XCOBRA-IIIC methodology (Ref. 10) to predict the minimum departure from nucleate boiling (MDNBR) for the event. The MDNBRs were calculated with the approved XNB critical heat flux correlation (Ref. 11) for which the critical heat flux limit (CHFL) is 1.17. The analyses were structured to support a Technical Specification F delta-H limit of 1.65.

The initial operating conditions included a power level of 102% of full power or 2R46 MWt, an RCS flow rate set at the thermal design flow rate of 97.29 x 10 lbm/hr, pressurizer pressure of 2220 psia and core inlet temperature of 550.2 0 F. Basic assumptions used in the analyses included:

(a) precluding the withdrawal function of automatic rod control, (b) non-positive MTC above 50% power for the rod drop transient, and (c) a k1 value in the OTDT trip function of 1.24 including uncertainties (this is for the RTD overall response time value of 4.75 seconds).

The discussion of the three reanalyzed accidents follows.

-7 3.5.1 Loss of External Load This accident is described in Section 15.2.2 of the FSAR and the reference analysis is presented in Reference 12. The licensee's analysis addresses the DNBR part of the Loss of External Load event. The other part relates to a challenge of the vessel pressurization criteria and was not considered.

This is because the mitigating features of the pressurizer spray and pressurizer relief valves were assumed to function which results in a conservative evaluation of the MDNBR for this event. The analysis included the effect of the RTD response time on the OTDT trip function. The event initiates with closure of the turbine control valves and after several pressurizer PORV and pressurizer and steamline valve openings, the reactor scram occurs on OTDT, with rod insertion at 15.5 seconds. The DNBR challenge results from core power and primary coolant temperature increase.

A plot of DNBR versus time was provided. The minimum DNBR occurred at 16.3 seconds and was computed with a correlated CHFL of 1.19. Since this is above the limit of 1.17, we find that the acceptance criterion has been met.

3.5.2 Uncontrolled Control Rod Assembly at Power This accident is described in Section 15.4.2 of the FSAR. The licensee's analysis addresses the limiting uncontrolled rod withdrawal transient resulting in reactor trip on the OTDT turbine trip. The reference analysis is presented in Reference 12. The reactivity insertion rate was that resulting in minimum DNBR and in simultaneous OTDT and power range high flux trips. The limiting event was a reactivity insertion ramp of 2 pcm/sec from beginning of cycle (BOC) full power initial conditions with positive reactivity feedback. The pressure increase due to coolant expansion and insurge flow to the pressurizer was limited to a maximum of 2274 psia by the PORVs.

Increasing core power and temperature resulted in a reactor trip on the OTDT reactor trip at 27.2 seconds. The minimum DNBR of 1.19 occurred shortly after at 27.4 seconds. Since the minimum DNBR was greater than the correlated CHFL of 1.17, we find that the acceptance criterion has been met.

3.5.3 Control Rod Misoperation (System Malfunction or Operator Error)

This accident is described in Section 15.4.3 of the FSAR. The licensee's analysis addresses the limiting rod drop transient resulting in a reactor trip on the OTDT reactor trip. The event is initiated by a dropped rod cluster control-assembly. This promptly inserts negative reactivity which reduces reactor power and turbine runback begins. The analysis indicated that turbine load reaches its programmed value at 9 seconds. The average coolant temperature first decreased in response to the power reduction and then increased due to the reduced secondary load demand. The temperature increase caused an insurge to the pressurizer, resulting in an opening of the PORVs at 16.1 seconds. In addition, there was a reactor trip due to OTDT at 61.1 seconds. The minimum DNBR occurred at 61.2 seconds with a value of 1.23. Since the minimum DNBR was greater than the correlated CHFL of 1.17, we find that the acceptance criterion has been met.

-8 3.6 LOCA Evaluation The elimination of the RTD bypass system impacts the uncertainties associated with RCS temperature and flow measurement. The licensee stated in Reference 2 that the magnitude of the uncertainties are such that RCS inlet T in and outlet T out temperatures, thermal design flow rate and the steam generator performance data used in the LOCA analyses will not be affected. Past sensitivity studies concluded that the inlet temperature effect on peak clad temperature is dependent on break size. As a result of these studies, the LOCA analyses are performed at a nominal value of T. without consideration of small uncertainties.

The RCS flow rate and slam generator secondary side temperature and pressure are also determined using the loop average temperature (T ) output. These nominal values used as inputs to the analyses are notIffected due to the RTD bypass elimination. It is concluded that the elimination of the RTD bypass piping will not affect the LOCA analyses input and hence, the results of the analyses remain unaffected. Therefore, the plant design changes due to the RTD bypass elimination are acceptable from a LOCA analysis standpoint without requiring reanalysis.

4.0 EVALUATION OF TECHNICAL SPECIFICATIONS Based on the evaluation stated above, the following addresses each of the specific proposed Technical Specification (TS) changes.

Technical Specification 2.1 A new statement was put in the TS relating to the minimum allowable flow and the flgw measurement uncertainty. This states - "The minimum RCS flow is 699.8x10 lbm/hr, which is the minimum thermal design flow of 97.3x10 lbm/hr with a 2.6% allowance added for instrument uncertainty associated with the precision calorimetric flow measurement."

This is acceptable as discussed in Section 3.3 of this Safety Evaluation (SE)

The 2.6% flow measurement uncertainty includes 0.1% for feedwater venturi meter fouling. A reference to the uncertainty calculation has been added to this TS section.

Technical Specification 2.3.1.2 The licensee performed an anlaysis and demonstrated that in the formula to determine the maximum value of OTDT (item d of TS 2.3.1.2), a calculated value for the constant k1 of 1.24 is adequate to conservatively prevent DNB. To allow for the estimated uncertainty, the licensee proposed the k1 value of 1.1365 to be used in the TS. The staff finds this TS k1 value to be acceptable since it incorporates the appropriate uncertainty into the calculated value.

Technical Specification 2.3.3 This is a new TS item that states that "The RCS narrow range temperature sensors response time shall be less than or equal to a 4.0 second lag time constant." This is acceptable as discussed in Section 3.1 of this SE.

-9 Technical Specification Bases 2.3 The reference statement on the RTD bypass piping as it affects transport time has been removed and replaced by a statement on the transport and response time of the RTDs. This is an editorial change to reflect the new design and is acceptable.

In addition, reference to FSAR Section 15.4.2 on the RTD bypass system piping is deleted. Since this piping will be removed, it is no longer pertinent. Also, a new paragraph is inserted which states - "The RCS temperature measurement response time parameters define the time delay between when the OTOT 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 (Refs. 7, 8, 9)."

New References 7, 8, and 9 were also added. These changes reflect the modification and are acceptable.

Technical Specification 3.1.3.1 This change reduces MTC by a specification that states that the MTC shall be 0 pcm/oF at 50% of rated power or above. Formerly, the specification stated that the MTC shall be +5.0 pcm/oF at 50% of rated power and linearly decreased to 0 pcm/oF at rated power. This change is acceptable as discussed in Section 3.4 of this SE.

Technical Specification Bases 4.1 This is a new addition to the Technical Specifications which states - "For RCS narrow range temperature sensors, verification of response time will be a part of calibration. Cross calibration of RCS narrow range temperature sensors will be performed on a refueling interval."

This is acceptable as discussed in Section 3.1 of this SE.

Technical Specification Table 4.1-1

-Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels Item 4 of this table, Reactor Coolant Temperature, has several additions.

The narrow range RTD response time will be tested and cross calibrated at least once every 18 months. These changes are'acceptable as discussed in Sections 3.1 and 3.2 of this SE.

5.0

SUMMARY

The impact of the RTD bypass elimination for H. B. Robinson, Unit No. 2 on the FSAR Chapter 15 non-LOCA accident analyses has been evaluated and found to be acceptable. For the events impacted by the increase in the channel response time, it has been demonstrated that the conclusions presented in the FSAR remain valid. For the remaining Chapter 15 non-LOCA

U0

-10 events, the effect of the increased initial RCS average temperature error allowance has been ascertained by separate evaluations. In all instances, the conclusions presented in the FSAR remain valid under this error allowance assumption and the DNBR limit value is met. The reduction of the MTC when operating at or above 50% rated power was found to be acceptable as it is in the conservative direction.

The licensee's analysis to support an RCS flow measurement uncertainty value, which includes the new hot leg RTD temperature accuracy, were found to be acceptable and pertinent Technical Specification changes were proposed.

The Technical Specification changes were found to be acceptable. The licensee has committed to obtain data for comparison of the temperature indications dfter modification with measurements prior to modification.

NRC will be notified of the results of the comparison. The licensee has also committed to implement new procedures to clean the feedwater venturi meters at refueling before the precision calorimetric is made at each refueling.

6.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on December 19, 1988 (53 FR 51021). Accordingly, based on the environmental assessment, the Commission has determined that the issuance of these amendments will have no significant effect on the quality of the human environment.

7.0 CONCLUSION

The Commission has issued a Notice Of Consideration of Issuance of Amendment of Facility Operating License and Opportunity for Hearing which was published in the Federal Register (53 FR 30879) on August 16, 1988. No petition to intervene or request for hearing has beer filed on this action.

The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by opertation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defence and security or to the health and safety of the public.

Prinicpal Contributor:

Harry Balukjian Dated:

January 9, 1-989

11

8.0 REFERENCES

1. Letter from L. W. Eury, Carolina Power and Light Company, to USNRC, dated July 26, 1988.

2. Letter from L. W. Eury, Carolina Power & Light Company, to USNRC, dated November 1, 1988.

3. WCAP-11889, RTD Bypass Elinination Report for H. B. Robinson, Unit 1, June 1988 (Proprietary), WCAP-11890 (Non-Proprietary with addendum 1) dated October 1988.

4. ANF-88-094, H. B. Robinson, Unit 2 Chapter 15 Overtemperature Delta-T Trip Event Analysis for Elimination of RTD Bypass Piping, July 1988.

November 1, 1988.

5. NUREG-0809, Safety Evaluation Report, Review of Resistance Temperature Detector Time Response Characteristics, August 1981.

6. NUREG/CR-4928, Degradation of Nuclear Plant Temperature Sensors, June 1987.

7. K. R. Carr, An Evaluation of Industrial Platinum Resistance Thermometer Temperature -

Its Measurement and Control in Science and Industry, ISA publication, Vol. 4, Part 2, 1972, pages 971-982.

8. B. W. Mangum, The Stability of Small Industrial Platinum Resistance Thermometers, Journal of Research of the NBS, Vol. 89, No. 4, July-August 1984, pages 305-350.

9. XN-NF74-5(A) and Sups. 1-6, Rev. 2, "Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTSPWR)," Exxon Nuclear Company, Richland, WA, October 1986.

10.

XN-NF-82-21(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, Richland, WA, September 1983.

11.

XN-NF-621(A), Rev. 1, "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, Richland, WA, September 1983.

12. XN-NF-84-74, Rev. 1, "Plant Transient Analysis for H. B. Robinson Unit 2 at 2300 MWt With Increased FNH" Exxon Nuclear Company, Richland, WA, April 1986.

13.

WCAP-11168 Rev. 1, "RCS Flow Uncertainty for Shearon Harris Unit 1,"

October 1986.

14. NUREG/CR-3659, "A Mathematical Model for Assessing the Uncertainties of Instrumentation Measurements for Power and Flow of PWR Reactors,"

February 1985.

7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-261 NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No.121 to Facility Operating License No. DPR-23 to the Carolina Power & Light Company (the licensee), which revised the Technical Specifications for operations of the H. B. Robinson Steam Electric Plant, Unit No. 2, located in Darlington County, South Carolina. The amendment is effective as of the date of its issuance.

The amendment changes certain reator parameters to reflect the correct reacto coolant loop resistance temperature detector (RTD) system response time and to support the elimination of the RTD bypass system. The amendment also reduces the range of reactor operation over which the allowable Moderator Temperature Coefficient could be positive.

The application for amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments.

Notice of Consideration of Issuance of Amendment and Opportunity for Hearing in connection with this action was published in the FEDERAL REGISTER on August 16, 1988 (53 FR 30879).

-2 Also in connection with this action, the Commission prepared an Environmental Assessment and Finding of No Significant Impact, which was published in the FEDERAL REGISTER on December 19, 1988 (53 FR 51021).

For further details with respect to the action, see (1) the application for amendment dated July 26, 1988, as supplemented August 26, and November 1, 1988, (2) Amendment No. 121 to Facility Operating License No. DPR-23, and (3) the Commission's related Safety Evaluation. All of these items are available for public inspection at the Commission's Public Document Room, 2120 L Street, N.W., Washington, D.C., 20555 and at the Hartsville Memorial Library, Nuclear Information Depcsitory, 220 N. Fifth Street, Hartsville, South Carolina 29550.

Dated at Rockville, Maryland this 9th day of January 1989.

FOR THE NUCLEAR REGULATORY COMMISISON Edward A. Reeves, Jr., Acting Project Director Project Directorate TI-1 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation