ML17325B248
| ML17325B248 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/09/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17325B247 | List: |
| References | |
| NUDOCS 8906220287 | |
| Download: ML17325B248 (26) | |
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1 UNITED STATES i NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.126 TO FACILITY OPERATING LICENSE NO.
DPR"58 INDIANA MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 DOCKET NO. 50"315
- 1. 0 INTRODUCTION By letter dated October 14, 1988, as supplemented December 30, 1988, and June 5, 1989,the Indiana Michigan Power Company (the licensee) requested an amendment to the Technical Specifications (TSs) appended to Facility Operating License No.
DPR-58 for the Donald C.
Cook Nuclear Plant, Unit No. 1.
The proposed amendment would permit the operation of future reload cycles of Unit 1 at reduced primary system temperature and pressure conditions.
The reduced temperature and pressure (RTP) conditions will decrease the steam generator U-tube stress corrosion cracking of the type observed at the D.
C.
Cook Nuclear Plant, Unit 2.
The licensee's contractor (Westinghouse) has determined that this RTP program should more than double the time to reach a given level of steam generator U-tube corrosion in comparison to the original temperatures and pressure.
D.
C.
Cook, Unit 1 is presently -licensed to operate at 3250 MWt, which is rated thermal power defined by Definition 1.3 of the Technical Specifications.
Some'ransient and accident analyses are performed at a higher power level to position Unit 1 for a potential power uprating.
However, not all of the analyses have been performed at this higher power level.
The small break loss-of-coolant accident (LOCA) analysis was, for example performed at a power of level of 3250 MWt with the high head safety injection cross-tie valve shut
~ and at 3588 MWt for all other analyzed plant conditions.
The staff's review of the RTP program for Unit 1 did not consider any issues related to a future power uprating.
The licensee performed analyses and evaluations to support the RTP program for D.
C.
Cook, Unit 1.
The licensee's efforts addressed full rated thermal power operation (3250 MWt) with a range of vessel average temperature between 547 F and 576.3 F.
Two discrete values of the pressure, 2100 psia and 2250
- psia, were used in the analyses and evaluations.
The analyses and evaluations support a maximum average tube plugging level of lOX, with a peak steam generator tube plugging level of 15X.
The licensee will select the desired operating temperature and the pressure on a cycle-by-cycle basis.
The licensee performed the safety analyses and evaluations at conservatively high power levels and high primary system temperatures in order to position both of the D.
C.
Cook units for future power uprating and in order to support potential future operation of Unit 2 at reduced temperatures and pressure.
The potential uprated power for Unit 1 that is partially supported by this analysis and evaluation is 3425 MWt, which corresponds to a reactor power level of 3413 MWt.
The design power capability parameters are given in Table 2.1-1 of Reference 2.
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- 2. 0 EVALUATION 2.1 NUCLEAR STEAM SUPPLY SYSTEM (NSSS 2.1.1 Lar e and Small Break LOCA Anal ses The licensee performed a large break LOCA analysis using the 1981 version of the Westinghouse ECCS Evaluation Model, which uses the BASH computer code.
The analysis assumptions include a total peaking factor, F
, of 2.15, a hot channel enthalpy rise factor,'-delta H, of 1.55, lOX safely injection flow degradation, a reactor power level of 3413 MWt, and 15K uniform steam generator tube plugging level.
A range of hot-leg temperatures of 580.7'F to 611.2'F and a range of cold-]eg temperatures of 513.3'F to 546.2'F, consistent with the temperature range of the RTP program, were considered in the analysis.
In the
- analysis, the reactor coolant system pressure was varied to justify plant operation at either 2100 psia or.2250 psia.
. A large-break LOCA analysis was also performed with the RHR cross-tie valve closed.
For this case, a reduced core power of 3250 MWt was used to compensate for the reduction in safety injection flow caused by the closed RHR cross-tie valve.
For those limiting pressure and temperature conditions which produced the largest peak clad temperature, a full break spectrum of discharge coefficients was performed.
The limiting break size was determined to be a cold-leg guillotine break with a discharge coefficient, C
, of 0.6, a hot-leg temperature of 611.2'F and a primary system pressure oil 2250 psia, assuming maximum safety injection flow.
The peak clad temperature was calculated to be 2180.5 F.
Based on these results, the requirements of 10 CFR 50.46 have been met for the Unit 1 large-break LOCA anaTysis.
The licensee performed a small-break LOCA analysis using the Westinghouse small-break ECCS Evaluat'ion Model, which uses the NOTRUMP code.
The analysis assumptions included a total peaking factor of 2.32, a hot channel enthalpy rise factor of 1.55, safety injection flow rates based on pump performance curves degraded 10K below design head and including the effect of closure of the high head safety injection cross-tie valve, and a uniform 15K steam generator tube plugging -level.
The analysis was performed at a core power level of 3250 MWt, a range of operating core average temperatures of 547~F to 581.3'F, and reactor pressure of either 2100 psia or 2250 psia.
All other plant conditions were analyzed't a power of 3588 MWt.
The licensee analyzed a spectrum of cold-leg breaks at the limiting reactor coolant system temperature and pressure conditions.
The limiting break size from this analysis was then analyzed at other temperature and pressure points of the operating range.
The limiting case was determined to be a three-inch diameter cold-leg break at a pressure of 2100 psia and at a core average temperature of 547'F.
This limiting break resulted in a peak clad temperature of 2122'F.
Based on these results, the requirements of 10 CFR -50.46 have been met for the Unit 1 small-break LOCA analysis.
The licensee reviewed the effect of the RTP program on the post-LOCA hot-leg recirculation time to prevent boron precipitation.
This time is affected by powe) level and various systems'ater volumes and boron concentrations.
Because these systems'ater volumes and boron concentrations are not affected by the RTP program, there is no effect on the post-LOCA hot-leg switchover time.
The licensee reviewed the effect of the RTP program on the post-LOCA hydrogen generation rates.
The assumption of 120 F maximum normal operations containment temperature bounds, for the analysis of record, the effect of the primary system temperature changes of the RTP program on the post-LOCA hydrogen generation rates.
- 2. 1.2 Non-LOCA Transients and Accidents The licensee has evaluated the impact of the RTP program on the non-LOCA events presented in Chapter 14 of the D.
C.
Cook, Unit j. FSAR.
The approved reload core design methodology and design codes were used.
'The evaluations were performed to support the operation of Unit 1 at a core power of 3250 MMt over a vessel average temperature range between 547~F and 576.3~F at a primary system pressure of either 2100 psia or 2250 psia.
The evaluation assumes a steam generator tube plugging level of lOX, with a peak steam generator tube plugging level of 15K.
The non-LOCA safety evaluation supports the parameters of the RTP program with the exceptions of the steamline break mass and energy releases outside containment, which were evaluated at a full power vessel average temperature no greater than the current D.
C.
Cook Unit 1 full power average temperature, T
, of 567.8'F....
avg'he evaluation performed by the licensee also considered the parameters for a potential uprating of Unit 1 to reactor core power level of 3413 MWt, with a vessel average temperature range between 547'F and 578.7 F at a primary system pressure of either 2100 psia or 2250 psia.
The steam generator tube plugging level is assumed to be the same as for the RTP program.
Even though the non-LOCA evaluation may have been performed for the uprated core power and its associated parameters, the staff's review of this license amendment does not address a D.
C.
Cook Unit 1 power uprating.
The licensee revised certain reactor trip and engineered safeguards features (ESF) setpoints to provide adequate operating margins for the RTP operating conditions.
Revised reactor trip setpoints were incorporated in the over'temperature-delta T (OTDT) and overpower-delta T (OPDT) trip functions.
The revised ESF setpoints affects the low steamline pressure value of the high-high steamline flow coincident with a low steamline pressure actuation logic.
The new OPDT and OTDT reactor trip setpoints were developed by the licensee for a new set of core thermal safety limits for the RTP program at a reactor core power level of 3413 HMt.
The approved setpoint methodology of Reference 3 was used.
For those events analyzed with the approved Improved Thermal Design Procedure (ITDP), Reference 4,
a safety-lim'it value of 1.45 was used for the Departure from Nucleate Boiling Ratio (DNBR).
This is conservative compared to the design DNBR value of 1.32 for a thimble cell and 1.33 for a typical cell required to meet the DNB design basis.
In, the -safety analysis for D.
C.
Cook, Unit 1, the licensee assumed the high pressurizer water level trip setpoint of 100X (nominal reactor setpoint).
Furthermore, the reference average temperature used in the OPDT and OTDT trip setpoint equations are rescaled to the full power average temperature each time the cycle average temperature is changed.
Similarly, the appropriate value of primary system pressure of either 2100 or 2250 psia was used in the two tr ip setpoint equations.
For the revised ESF setpoint of the high-high steamline flow'oincident with low steamline pressure, the low steamline pressure setpoint was lowered from 600 psig,to 500 psig to accommodate the range of conditions of the RTP program and a potential power uprating.
- 2. 1.3 Steamline Break Mass/Ener Releases The current mass and energy releases for the inside containment analysis is based on analyses performed for Cook Unit 2, which are also applicable to Cook Unit 1.
Data are represented in Chapter 14 of the FSAR for Unit 2 at power levels of 0, 30, 70, and 100X power.'or the "at power" analyses, the initial primary system temperature and secondary steam pressures of the RTP program are lower than those in the Unit 2 FSAR analyses.
The mass blowdown rate is dependent on steam pressure and since the steam pressure will be less than the current
- analyses, the initial mass blowdown rate will be lower.
The lower steamline pressure setpoint (500 psig) of the ESF actuation signal does not significantly impact the analysis because the lead-lag compensation results in a steamline pressure signal which anticipates the rapid decrease in pressure caused by a steamline break.
Based on these considerations,,the licensee concludes that the RTP program will result in a lower integrated energy release into containment and that the data used in the Unit 2 FSAR remains bounding.
A study was performed for Unit 1 of the. mass,and energy release. outside containment to address equipment qualification issues (Ref. 5).
Cases at 70'nd lOOX power were analyzed.
The analysis presented in Reference 5 assumed the full power vessel average temperature to be 567.8 F.
Any reduction in full power T from the analyzed T
and the associated reduction in initial steam pressure Sill result in,less 1Ni)ting releases.
The low steamline pressure value assumed in the analysis supports the reduced value of the setpoint to.500 psig.
The increased level of steam generator tube plugging is acceptable because the analysis assumed better heat transfer characteristics.
The licensee concludes that the current mass and energy release analysis is acceptable for the RTP program as long as the full power T is equa'1 to or less than 567.8'F.
avg
- 2. 1. 4 Startu of an Inactive Loo The licensee evaluated the startup of an inactive loop event.
This event cannot occur above the P-7 permissive setpoint of 10K power as restricted by the Technical Specifications.
The parameters assumed in the FSAR analysis for
.three-pump operation at 10K power remain bounding for the parameters for 10K power condition.
The licensee concludes, therefore, that the conclusions presented in the FSAR remain valid.
2.1.5 Uncontrolled Rod Bank Withdrawal from a Subcritical Condition The uncontrolled rod bank withdrawal from a subcritical condition transient causes a power excursion.
This power excursion is terminated, after a fast power rise, by the negative Doppler reactivity coefficient of the fuel, and a
reactor trip on source, intermediate, or power range flux instrumentation.
The power excursion results in a heatup of the moderator/coolant and the fuel.
The analysis used a reactigity insertion rate of 75 pcm (note that one pcm is equal to a reactivity of 10 delta K/K).
This reactivity insertion rate is greater than for the simultaneous withdrawal of the two sequential control banks having the greatest combined worth at the maximum speed of 45 inches/minute.
The neutron flux overshoots the nominal full power value; however, the peak heat flux is much less than the full power nominal value because of the inherent thermal lag of the fuel.
The analysis, with the reduced system pressure of 2100 psia, yields the minimum value of DNBR.
The analysis is performed using the Standard Thermal Design Procedure (STDP).
The W-3 DNB correlation was issued to evaluate DNBR in the span between the lower non-mixing vane grid and
-'5" the first mixing vane grid.
The MRB-1 DNB correlation is applied to the remainder of the fuel assembly.
From the analysis performed, the licensee concludes that the DNB design bases are met for all regions of the core, and therefore, the conclusions in the FSAR remain applicable for a reduction in nominal system pressure to 2100 psia.
2.1.6 Uncontrolled Control Rod Assembl Bank Withdrawal at Power The uncontrolled rod bank withdrawal from a power condition transient leads to a power increase.
The transient results in an increase in the core heat flux and an increase in the reactor moderator/coolant temperature.
The reduction in pressure for the RTP program is non-conservative with respect to DNB.
In
- addition, a revised Overtemperature Delta-T setpoint equation is being assumed in the Cook Unit 1 analyses.
The Power Range High Neutron Flux and Overtempera-ture Delta-T reactor trips provide the primary protection against DNB.
Both minimum and maximum reactivity cases were analyzed over a range of reactivity insertion rates.
The licensee provided quantitative results for the maximum reactivity feedback case for power leve)s of 10X, GOX, and lOOX power for a range of reactivity insertion rates.
The results indicate that the DNBR limit is met for all the cases.
The licensee examined a number of cases associated with the pressurizer water volume transient caused by an uncontrolled control rod assembly bank withdrawal-at-power event.
It was determined that credit for high pressurizer water level reactor trip was required to prevent the pressurizer from filling.
The licensee assumed a value of 100X narrow range span (NRS) for the high pressurizer water level reactor trip setpoint.
A time delay of 2 seconds was assumed for trip actuation until rod motion becomes adequate to terminate the transient.
Thus the high neutron flux and o>ertemperature-delta T reactor trips provide adequate protection over the range of possible reactivity.insertion rates in that the minimum value of DNBR remains above the safety-limit DNBR value.
In addition, the high pressurizer water level reactor trip prevents the pressurizer from filling.
2.1.7 Rod Cluster Assembl Misali nment The rod cluster control assembly misalignment events consist of three separate events:
(1) a dropped control rod, (2) a dropped control bank, and (3) a statically misaligned control rod.
These events were reanalyzed because the reduction in pressure for the RTP program is nonconservative with respect to the DNB transient.
A dropped control rod or control bank may be detected in the following manner:
(1) by a sudden drop in the core power as seen by the nuclear instrumentation system; (2) by an asymmetric power distribution as seen by the excore neutron detectors or the core exit thermocouples; (3) by rod bottom signal; (4) by the rod position deviation monitor; and (5) by rod position indicators.
A misaligned control rod may be detected in the following manner; (1) by an asymmetric power distt ibution as seen by the excore neutron detectors or the core exit thermocouples; (2) by the rod position deviation monitor; and (3) by rod position indicators.
The resolution of the rod position indicator channel is t5 percent or 212 steps
(+7.5 inches).
Deviation-of any control rod from its group by twice this distance (124 steps or 415 inches) will not cause power distribution worse than the design limits.
The rod position deviation monitor provides an alarm before a rod deviation can exceed 0 24 steps or t 15 inches.
The dropped rod event was analyzed using an approved methodology (Ref. 6).
A dropped rod or rods from the same group will result in a negative reactivity insertion which may be detected by the negative neutron flux rate trip circuitry.
If detected, a reactor trip occurs in about 2.5 seconds.
For those dropped rod events for which a reactor trip occurs, the core is not adversely impacted because the rapid decrease in reactor power will reach an equilibrium value dependent on the reactivity feedback or control bank withdrawal (if in automatic control).
The limiting case for this class of events is 'the case with the reactor in automatic control.
For this case a power overshoot occurs before an equilibrium power condition is reached.
The licensee states that, using the methodology of Reference 6, all analyzed cases'result in DNBR values which are within the safety-limit DNBR value.
The licensee states that a dropped rod bank results in a reactivity insertion of at least 500 pcm.
This will be detected by the negative neutron flux rate trip circuitry and cause a reactor trip within about 2.5 seconds of the initial motion of the rod bank.
Power decreases rapidly and there is, therefore, no adverse impact on the reactor core.
The most severe misalignment cases, with respect to DNBR, are those in which one control rod is fully inserted or where control bank "D" is fully inserted but with one control rod fully withdrawn.
Multiple alarms alert the operator before adverse conditions are reached.
The control bank can be inserted to its insertion limit with any control rod fully withdrawn without DNBR falling below the safety-limit DNBR value, as shown by analysis.
An evaluation performed by the licensee indicates that control rod banks other than the control bank would give less severe results.
For the case with one rod fully inserted, DNBR remains above the safety-limit DNBR value.
For all cases foll'owing identification of a control rod misalignment, the operator is required to perform actions in accordance with plant Technical "Specifications and procedures.
- 2. 1.8 Chemical and Volume Control S stem Malfunction The boron dilution event was analyzed by the licensee for startup and power operation.
The analysis is performed to show that sufficient time is available to the operator to determine the cause of the dilution event and take corrective action before the shutdown margin is lost.
The licensee reports that 45 minutes is available for Mode 1 (power operation) and 68 minutes for Modes 2 or 3 (startup or hot standby conditions)
(Ref. 7).
- 2. 1.9 Loss of Reactor Coolant Flow The loss-of-flow transient causes the reactor power to increase until the reactor trips on either a low-flow trip signal or reactor coolant pump power supply undervoltage signal.
The reactor power increase causes a reactor moderator/coolant temperature increase.
This initial coolant temperature increase causes a positive reactivity insertion because of the positive moderator temperature coefficient.
The licensee analyzed both a partial loss-of-flow (loss of one pump with four coolant loops in operation) transient and a complete loss-of-flow transient (loss of four pumps with four coolant loops in operation).
For the partial loss-of-flow transient, the reactor is assumed to be tripped on a low-flow signal.
For a complete loss-of-flow transient, the reactor is assumed to be tripped on a pump undervoltage signal.
For either event, the average and hot channel heat fluxes do not increase significantly above their initial values and the DNBR remains above the safety-limit DNBR value.
2.1.10 Locked Rotor Accident The locked rotor accident causes a rapid reduction in the fluid flow through the affected loop.
The reactor trips on a low-flow signal which rapidly reduces the neutron flux upon control rod.insertion.
Control rod motion starts 1 second after the flow in the affected loop reaches 87K of its nominal value.
The licensee evaluated this accident assuming that offsite power is available.
No credit is taken for the pressure-reducing effect of the pressurizer relief valves, pressurizer
- spray, steam dump, or controlled feedwater flow after reactor trip.
The licensee performed an analysis to determine the DNB transient and to demonstrate that the peak system pressure and the peak clad temperature remain below limit values.
The peak reactor coolant system pressure of 2588 psia reached during the transient is less than that which would cause stresses to exceed the faulted conditions stress limits.
The peak clad temperature reached is 1959~F.
Less than 4.5X'of the fuel rods in the most limiting fuel assembly reach values of DNBR less than the safety-limit DNBR value.
These results indicate that the RTP program assumptions give acceptable consequences for the locked rotor accident.
- 2. 1.ll, Loss of External Electrical Load The loss-of-external-electrical-load event was analyzed by the licensee to show the adequacy of pressure-relieving devices and to demonstrate core protection.
This reanalysis was necessary because of changes in reactor pressure and temperature conditions for the RTP program and because of changes.to the Overtemperature-Delta T reactor trip setpoint equation.
Maximum and minimum reactivity feedback cases were examined, with the case analyzed with and without credit for pressurizer sprays and power-operated relief valves.
For the minimum reactivity feedback case with pressurizer pressure control, the reactor trips on a high pressurizer pressure signal.
For the maximum reactivity feedback case with pressurizer pressure control, the reactor trips on a low-low steam generator water level signal.
For the minimum reactivity feedback case without pressurizer pressure control, the reactor trips on a high pressurizer pressure signals For all four cases, the minimum value of DNBR remains well above the safety-limit.DNBR value and the Overtemperature-Delta T setpoint was not reached.
The analysis confirms that the conclusions of the FSAR remain valid for this event for the RTP program.
- 2. 1.12 Loss of Normal Feedwater Flow The loss-of-normal-feedwater-flow event was analyzed by the licensee to show that the auxiliary feedwater system is capable of removing the stored and decay
- heat, thus preventing overpressurization of the reactor coolant system or uncovering the core, and returning the plant to a safe condition.
The reanalysis was based on a positive moderator temperature coefficient.
A conservative decay heat model based on the ANSI/ANS-5. 1-1979 decay heat standard (Ref. 8) was used.
Pressurizer power operated relief valves and the maximum pressurizer spray flow rate were 'assumed to be available since a lower pressure results in a greater system expansion.
The initial pressurizer water level was assumed to be at the maximum nominal setpoint of 62X narrow range span.
Reactor trip occurred when the low-low steam generator water level trip setpoint was reached.
The results of the analysis show that a loss of normal feedwater does not adversely affect the reactor core, the reactor coolant system, or the steam
- system, and that the auxiliary feedwater system is sufficient to prevent water relief through the pressurizer relief or safety valves.
The pressurizer does
~
~
gI not fill and, therefore, the conclusions of the FSAR remain valid for this event, including RTP conditions.
2.1.13 Excessive Heat Removal Due to Feedwater S stem Malfunctions The excessive-heat-removal event due to feedwater system malfunction was analyzed by the licensee to demonstrate core protection.
This analysis was necessary because of changes in reactor core temperatures and pressure for the RTP program and because
'of changes to the OTDT and OPDT trip setpoints.
This event is an excessive-feedwater-addition event caused by a control system malfunction or an operator error which allows a feedwater control valve to open fully.
The licensee analyzed'oth full power and hot zero power cases.
Both cases assumed a conservatively large negative moderator temperature coefficient.
'The full power case assumed the reactor was in automatic or m'anual control.
The Improved Thermal Design Procedure (ITDP) of Reference 4 was.used in the analysis.
For the accidental full opening of one feedwater control valve with the reactor at hot-zero power conditions, the licensee determined that the maximum reactivity insertion rate is less than the maximum reactivity insertion rate analyzed. in.
the Uncontrolled-Rod-C1uster-Assembly-Bank-Withdrawal-at-Subcritical-Condition event.
Thus, this hot-zero power case is bounded by the results obtained previously for the other event.
In addition, if the event were to occur at a hot-zero power and an exactly critical condition, the power range high neutron f'lux trip (low setting) of about 25K of nominal full power will trip the reactor.
The hot-full power case with the reactor in automatic control is more severe than the case with the reactor in manual control.
For all excessive feedwater
- cases, continuous addition of cold feedwater is prevented by automatic closure of all feedwater isolation valves on steam generator high-high level signal.
A turbine trip's then initiated and a reactor trip on a turbine trip is then assumed.
The results presented by the licensee demonstrate the safe response of Cook Unit 1 to the event, at hot-full power and in automatic control, with the DNBR remaining well above the safety-limit DNBR value.
- 2. 1.14 Excessive Increase in Secondar Steam Flow The excessive-increase-in-secondary-steam-flow event was analyzed by the licensee to demonstrate core protection.
This event is an overpower transient for which the fuel temperature will rise.
It was analyzed because of reactor core temperature and pressure changes for the RTP program and because of changes to the OTDT and OPDT setpoints.
The Cook Unit 1 reactor control system is designed to accommodate a 10X step load increase and a 5X-per-minute ramp load
. increase over the r ange of 15 -to 100 percent of full power.
Load increase in excess of these rates would probably result in a reactor trip.
Four cases were analyzed by the licensee.
These included minimum and maximum reactivity feedback cases with each case analyzed for both manual and automatic reactor control.
For the minimum reactivity feedback cases, a zero moderator temperature coefficient was assumed to bound the positive moderator temperature coefficient.
For all the cases, no credit was taken for the pressurizer heaters.
The analyses used the ITDP of References 4.
The studies show that the reactor reaches a new equilibrium condition for all the cases studied, with DNBR remaining well above the safety-limit DNBR value.
The operators would follow normal plant procedures to reduce power to an acceptable value to conclude the event.
2.1.15 Loss of all AC Power to the Plant Auxiliaries The loss-of-all-AC-power-to-the-plant-auxiliaries event was analyzed to demonstrate the adequacy of the heat removal capability of the auxiliary feedwater system.
This transient is the limiting transient with respect to the possibility of pressurizer overfill.
This event is more severe than the loss-of-load event because the loss of AC power results in a flow coastdown due to the loss of all four reactor coolant pumps.
'This results in a reduced capacity of the primary coolant,to remove heat from the core.
A positive moderator temperature coefficient was assumed in the analysis.
A conservative decay heat model based on the ANSI/ANS-5. 1-1979 decay heat standard (Ref. 8) was used.
No credit is taken for the immediate release of the control rods caused by the loss of offsite power.
Instead a reactor trip is assumed to occur on a steam generator low-low level signal.
Pressurizer power operated relief valves and the maximum pressurizer spray flow rate was assumed to be available since a
lower pressure results in a greater system expansion.
The initial pressurizer water level is assumed to be at the maximum nominal setpoint of 62'arrow range span plus uncertainties of 5X narrow range span.
The results demonstrate that natural circulation flow is sufficient to provide adequate decay heat removal following reactor trip and reactor coolant pump coastdown.
The pressurizer does not fill.
Thus, the loss of AC power does not adversely affect the core, the reactor coolant system, or the steam
- system, and the auxiliary feedwater system is sufficient to prevent water relief through the pressurizer relief or safety valves.
2.1. 16 Steaml ine Break The steamline break accident was analyzed by the licensee to assess the impact of the reduced reactor coolant system pesssure of the RTP program and the low steam pressure setpoint (lowered>>from 600 psig to 500 psig) of the coincidence logic with high-high steam flow for steamline isolation and safety injection actuation.
An end-of-life shutdown margin of 1.6X delta K/K for no load, equilibrium xenon conditions, with the most reactive control rod stuck in its fully withdrawn position, was assumed.
A negative moderator temperature coefficient corresponding to the end-of-line rodded core was assumed.
The licensee evaluated four combinations of break sizes and initial plant conditions to determine the core power transient which can result from large area pipe breaks.
The first case was the complete severance of a pipe downstream of the steam flow restrictor with the plant at no-load conditions and all reactor coolant pumps running.
The second case was the complete severance of a pipe inside the containment at the outlet of the steam generator with the plant at no-load conditions and all reactor coolant pumps running.
The third case is the same as the first case with the loss of offsite power simultaneous with the generation of a Safety Injection Signal (loss of offsite power results in reactor coolant pump coastdown).
The fourth case is the same as the second case with
. loss of offsite power simultaneous with the generation of a Safety Injection Signal.
A fifth case was performed to show that the DNBR remains above the safety-limit DNBR value in the event of the spurious opening of a steam dump or relief valve.
The licensee determined that the first case was the limiting
- case, that is, the double-ended rupture of a main steam pipe located upstream of the flow restrictor with offsite power available and at no-load conditions.
The results indicate that the core becomes critical with the control rods inserted (however, with the most reactive control rod stuck out) before boron solution at 2400 ppm enters the reactor coolant system.
The core power peaks at less than the nominal full core power.
The DNB analysis showed that the
minimum DNBR remained above 'the safety limit DNBR value, even though this event is classified as an accident with fuel rods undergoing DNB not precluded.
The analysis performed by the licensee demonstrates that a steamline break accident will not result in unacceptable consequences.
2.1.17 Ru ture of Control Rod Drive Mechanism Housin Rod E ection Accident The rod ejection accident is analyzed at full power and hot, zero-power conditions for both beginning-of-cycle (BOC) and end-of-cycle (EOC).
The analysis used ejected rod worth and transients peaking factors 'that are conservative.
Reactor protection for,a rod ejection is provided by neutron flux trip, high and low setting, and by the high rate of neutron flux increase trip.
The analysis modeled the high neutron flux trip only.
The maximum fuel temperature and enthalpy occurred for hot, full-power BOC case.
The peak fuel enthalpy was,
- however, below 200 cal/gm for all the cases analyzed.
For the hot, full-power cases, the amount of fuel melting in the hot pellet was less than lOX.
Because fuel and clad temperatures and the fuel enthalpy do not exceed the FSAR limits, the conclusions. of the FSAR remain val.id..
Based on a review of the licensee's evaluation and analysis of the non-LOCA transients and accidents (2. 1.3 through 2. 1.17) for the reduced temperature and pressure operation (the RTP program), the staff concludes that they are acceptable because (1) approved methodologies and computer codes have been
- used, and (2) all applicable safety criteria have been met.
This review is based on (1) a full power vessel average temperature of less than or equal to 567.8 F, (2) a steam generator tube plugging level of lOX with a peak tube plugging level of 15K, and (3) the minimum measured flow requirement of 91,600 gpm per loop is met.
2.1.18 Steam Generator Tube Ru ture SGTR Accident The licensee analyzed the steam generator tube rupture (SGTR) event for Cook Unit 1 using methodology and assumptions consistent with those used for the Cook FSAR SGTR analysis.
The range of parameters associated with a future rerating
.program and the RTP program were used in sensitivity analyses to assess the impact of these programs on the primary-to-secondary break flow and the steam released to the atmosphere by the affected steam generator.
These two factors affect the radiological consequences of an SGTR accident.
In addition, the licensee's evaluation of the radiological doses considers the effect of the noble gas concentrations.
The licensee states that the results of the analyses show that the doses remain within a small fraction (10K) of the 10 CFR Part 100 guidelines for both the thyroid and whole body doses.
Since the worst case doses are within the 10 CFR Part 100 guidelines, the staff concludes that the analysis of the SGTR is acceptable
- 2. l. 19 Fuel Structural Evaluation The fuel assembly lift and buoyancy forces are increased for the RTP program at Cook Unit 1 because a reduction in reactor coolant system temperature of about 20 F will increase the coolant density by about 3X.
The licensee evaluated this force increase against the fuel assembly allowable holddown load.
The results of the evaluation show that the increased force is well within the minimum spring holddown force design margin.
In addition, the licensee determined that the cold-leg break remains the most limiting pipe rupture transient with respect to lateral and vertical hydraulic forces.
Based on the licensee's review, the staff concludes that the 15x15 fuel assembly design remains acceptable.
I I~ The fuel rod design was evaluated to assess the impact of future rerating.
The licensee determined that the rod internal pressure criterion will continue to be the more important factor in fuel burnup capabilities.
The fuel will also undergo more severe fuel duty because of the uprated power.
The licensee plans to perform cycle-specific verification for each reload to assure that all fuel rod design criteria are met.
2.1.20 Justification-for-Pressurizer-Level The purpose of the Pressurizer High Level Limit is to ensure that a steam bubble is present in the pressurizer prior to power operation to minimize the consequences of overpressure transients and the possibility of passing water
~ through the relief and safety valves.
The safety analysis assumes a maximum water volume which corresponds to about 65% indicated level.
This nominal indicated level is maintained during normal operation by the pressurizer level control system.
The licensee (and the fuel supplier - Westinghouse) recommends the use of
'92K'or the Pressurizer High Level trip limit.
They state that this new trip limit will still ensure the presence of a steam bubble in the pressurizer.
The pressurizer level will, however, be controlled to the nominal value.
For normal operations (Condition I event),
the reactor parameters, including the pressurizer
- level, do no significantly deviate from their nominal values.
The licensee concludes that, for the pressurizer level to exceed the nominal level, a
transient or accident must occur for which protective action is provided by the Reactor Protection System.
Any other possible conditions for which the nominal level would be exceeded before and during a transient would require a transient or transients beyond those usually considered for an FSAR type of analysis.
The staff concludes on the basis of the licensee's evaluation that a Pressurizer High Level Trip of 92% is acceptable.
2.2 BALANCE OF. PLANT SYSTEMS The licensee states that balance of plant (BOP) systems and components were analyzed for the effects of operation at reduced. temperature and pressure conditions.
The secondary side conditions for these analyses were determined using the Performance Evaluation and Power System Efficiencies PEPSE) heat balance data (14.20 E6 lb/hr main steam flow and main feed flow.
The systems reviewed were the non safety-related secondary side power generating and nonpower generating systems.
Included in the licensee's analysis were portions of the main feedwater, main steam, steam generator blowdown (SGBS),
component cooling water (CCWS), auxiliary feedwater (AFS), heating, ventilation, and air conditioning (HVAC), service water, waste disposal, fire protection, radiation monitoring, and spent fuel pool (SFP) cooling and cleanup systems.
The performance of the above BOP systems was evaluated at the reduced temperature and pressure by using the new primary side NSSS data (14.20E6 lb/hr main steam and main feed flow, and 434'F main feed temperature) furnished by Westinghouse.
The licensee states that the impact on containment pressures and temperatures following a postulated design basis main steam line break was evaluated and its
- effect on equipment qualification was verified.
The flooding analysis in safety-related areas of the plant as a result of a postulated pipe break was reevaluated due to the slight increase in flow rates in the main feed, condensate, and main steam systems.
The turbine-generator system was also evaluted to confirm its integrity and performance at the increased steam volumetric flow rate and to verify that the original turbine missi le analysis remains valid.
"12" The licensee's analysis of BOP system performance provided the following findings concerning the RTP conditions at the present licensed power level of 3250 HWt NSSS power:
(a)
(b)
(c)
The capability of the safety-related portion of the main feedwater system will not be affected and will continue to perform its safety function because the proposed RTP conditions 'are bounded by the existing main feedwater system design.
The licensee s analysis of the pressure/temperature rating conditions for the system confirms that pressure boundary integrity will not be affected.
In addition, the main feedwater system isolation valve closure tim'e is not affected by the RTP-imposed conditions.
The capability of the steam generator blowdown system to remove impurities from the secondary side remains essentially the same for the RTP-imposed conditions during normal operation based on the exsisting design.
The reactor makeup water system's (MSW) capability to provide demineralized water for makeup and flushing operations throughout the NSSS auxilliaries, the radwaste
- systems, and fuel pool cooling and cleanup system is not challenged because the existing system design is based on the worst case demand which bounds the RTP conditions.
(d)
The licensee confirmed that safety-related equipment will not be affected by changes in the flooding analysis due to the RTP conditions.
Flooding in the auxiliary building due to failure of nonseismic Class I piping has been reviewed.
The licensee analyzed systems having access to large water volumes and/or potentially large flowrates were considered as discussed in the FSAR.
The only such system is the main feedwater system.
Since the changes in flow in the main feedwater system are still within the design limits, the results concerning flooding discussed in the FSAR are still applicable.
Flooding in the containment is slightly increased due to the larger initial water mass in the reactor coolant system because of the higher density at the reduced temperature.
This change was found to be within the volume margins used to determine the maximum flood-up elevation.
The containment flooding evaluation in the FSAR remains valid at the RTP-induced conditions.
(e)
The adequacy of the AFW system for accident mitigation was demonstrated in the Westinghouse accident analysis performed in support'f the RTP program under the following scenarios:
1.
Loss of main feedwater 2.
- 3. Main steam line rupture Each accident analysis demonstrated acceptance criteria such as system overpressure limits or DNB limits.
The AFW system's ability for design basis accident decay heat removal calculated in the RTP analysis is unaffected.
. ~
(g)
(h)
(k)
As evaluated in the RTP analysis, the heat loads in both the primary and secondary systems due to reactor decay heat remain unchanged.
Therefore, the Component Cooling Water System (CCWS) analysis and service water system (SWS) analysis in the FSAR remain valid.
For main steam line breaks inside the containment structure, the pressure and temperature will remain within the bounds of the peak pressure and temperature used in the evaluation of containment performance.
The initial primary temperatures and secondary steam pressures under the RTP conditions will be lower than those used in the FSAR analysis.
The licensee has confirmed that containment environmental qualification of equipment inside containment is not affected.
The superheated mass and energy release analysis outside containment was evaluated to address equipment qualification...
issues'.
The primary temper'atures and secondary steam pressures resulting from the RTP conditions will be lower than those used in the FSAR analysis.
The mass and energy release will be lower and operation with RTP will result in lower temperatures in the break areas.
As such, the current superheat mass and energy release analysis outside containment remains bounding provided the full power vessel average temperature is restricted to the currently-licensed 567.8'F and below.
The secondary pressure conditions assumed in the high energy steam line break analysis will be lower than those presented in the FSAR.
These bound the proposed RTP conditions and therefore the current analysis is sufficient.
The primary function of the spent fuel pool cooling system (SFPCS) is to remove decay heat that is generated by the elements stored in the pool..
Decay heat generation is proportional to the amount of radioactive decay in the elements stored in the pool which is proportional to the reactor power history.
Since the plant's rated power level of 3250 HWt remains unchanged, the demand on the SFPCS is not increased.
The purification function is controlled by SFPCS demineralization and filtration rates that are not affected by the RTP conditions.
The fire protection systems and fire hazards are independent of the plant operating characteristics with the exception of the slightly increased current requirements for the electric motor driven pumps in the primary system.
The increased load is due to the more dense water being pumped under the RTP conditions.
The increased current required is small and therefore is not considered to be a fire hazard.
The licensee confirmed that BOP systems have the capability to maintain plant operation under the RTP-induced conditions without modification to the existing design.
The staff has reviewed the FSAR and licensee submittals in order to verify that safety-related BOP system performance capability, as analyzed, bounds the changes in design basis accident assumptions created by the RTP operation.
The staff has confirmed that safety-related BOP system design capability, flooding protection, and equipment qualifications are bounded for the proposed rerating and therefore are considered acceptable as is.
Based on the above, the staff concludes that the proposed license amendment for the D.C.
Cook Nuclear Plant Unit 1 concerning the Reduced Temperature and Pressure is within the existing safety-related BOP system design'capability for design basis accident mitigation and, therefore, the staff s previous approval against the applicable licensing criteria for the main steam
- system, main feed
- system, CCWS,
- SWS, AFS,
- MSW, SGBS,
- SFPCS, flooding protection, containment
'erformance, and equipment qualifications remain valid.
The staff, therefore, finds the BOP systems concerned acceptable for continued operation at 'the proposed reduced temperature and pressure.
2.3 REACTOR VESSEL AND VESSEL INTERNALS The reactor vessel. is designed to the ASME Boiler and Pressure Vessel Code;Section III (1965 Edition with addenda through the winter 1966).
The licensee has determined that the operation of the reactor vessel under the most limiting:
conditions of the RTP rerating is acceptable for its original 40-year design objective.
All of the stress intensity and usage factor limits of the applicable code for the Unit 1 reactor vessel are still satisfied when the RTP is incorporated, with the exception of the 3Sm limit for the Control Rod Drive Motor (CROM) housings and outlet nozzle safe end.
However, the code permits exceeding the 3Sm limit provided plastic or elastic/plastic analysis criteria are met.
The licensee's review of the reactor vessels internals for the RTP program included three seperate areas:
a"thermal/hydraulic assessment, a
RCCA drop time evaluation, and a structural assessment.
Force increases were calculated for the upper core plate, across the core barrel, and in the upper internals near the outlet nozzles.
In these areas the existing.margin was determined to be sufficient to accommodate the increased stresses.
The results of this review indicate that the original reactor internals components remain in compliance with the current design require-ments when operating at the new range of primary temperatures and pressures.
The PTS rule requires that at the end-of-life of the reactor vessel, the projected reference temperature (calculated by the method given in 10 CFR 50.61(b)(2),
RT/pts) value for the materials in the reactor vessel beltline be less than the screening criterion in 10 CFR 50. 61(b)(2).
The RT/pts value is dependent upon the initial reference temperature, margins for uncertainty in the initial reference temperature and calculational procedures, the amounts of nickel and copper in the material, and the neutron fluence at the end-of-life of the reactor vessel.
Of these properties, only neutron fluence is affected by rerating with RTP.
Since the colder coolant in the downcomer region is more dense and thus provides for a more efficient neutron shield for the reactor
- vessel, fluence estimates are lower than those at current operating conditions.
All other properties are independent of the RTP-induced conditions.
The effects of NRC Generic letter 88-11, dated July 12, 1988, regarding Regulatory Guide 1.99 Rev.
2 were evaluated by Westinghouse and determined to not be significant for RTP.
The effect of RTP will be incorporated by the licensee in future PTS submittals.
"15-An evaluation was performed to determine the impact of RTP rerating on the applicability of the PTS screening criteria in terms of vessel failure.
A probabilistic fracture mechanics sensitivity study of limiting PTS transient characteristics, starting from a lower operating temperature, showed that the conditional probability of reactor vessel failure will not be adversely affected.
Therefore, the overall risk of vessel failure will not be adversely
- impacted, meaning that the screening criteria in the PTS Rule are still applicable for the D.C.
Cook Nuclear Plant Unit 1 reactor vessel relative to rerated conditions.
Analysis of the CRDM housings and the outlet nozzle safe end shows the maximum range of primary plus secondary stress intensity exceed the 3Sm limit.
The
- licensee, however, performed a simplified elastic/plastic analysis in 'accordance with paragraph NB-3228.3 of the ASME Boiler and Pressure Vessel Code,Section III (1971 or later edition) and the higher range of stress intensity is justified.
Therefore, based on the licensee's reviews and analysis of the above portions of the reactor vessel and internals, the staff concludes that the conditons imposed on the reactor vessel and.internals by the RTP rerating are acceptable.
2.4 TURBINE MISSILES The FSAR turbine missile analysis is based on a low pressure turbine failure.
The licensee's analysis of the slightly changed steam conditions entering the low pressure turbine shows that the probabilty of a low pressure turbine missile is virtually unaffected.
The factors that directly or indirectly cause stress corrosion cracking in the low pressure turbine wheels are steam pressure and temperature, mass flow rate, steam moisture content, water chemistry, oxygen level, and turbine speed.
The licensee reported that changes fA these factors are negligible due to the RTP-induced conditions.
The only noticeable change that the staff can determine is a 1.0X increase in the steam flow rate.
The staff's conclusion, based on the licensee's review, is that the turbine missile hazard is neglibily affected by the RTP conditons and is, therefore, acceptable.
2.5 PLANT STRUCTURAL AND THERMAL DESIGN The NSSS review consisted of comparing the existing NSSS design with the performance requirements at the rerated RTP conditions.
The current components of the Cook Unit 1/model 51 steam generators continue to satisfy the requirements of the ASME B8PV Code,Section III,(the code applicable for the design of the Cook Nuclear Plant Unit 1), for this program.
In addition, thermal hydraulic evaluations of the steam generators show acceptable stability and circulation ratios at the RTP rerated conditions.
Circulation ratio is primarily a function of power, which is unchanged, therefore is itself virtually unchanged.
The dampening factor characterizes the thermal and hydraulic stability of the steam generator.
Westinghouse has determined that all dampening factors are negative at nearly the same value as the current operating conditions.
A negative dampening factor indicates a
stable device.
Since the code requirements continue to be satisfied, and since stability and circulation ratios have been determined by Westinghouse to be within the design criteria, the staff concludes that RTP operation is acceptable for the Model 51 steam generators.
The pressurizer structural analysis was performed by modifying the original D.C.
Cook Nuclear Plant Pressurizer analysis ("Model 51 Series Pressurizer Report" ).
The analysis was performed to the requirements of the ASME Code 1968 Edition, which is the design basis for. the D.C.
Cook Nuclear Units.
.The only ASME Code requirement affected by the transient modifications was fatigue.
The limiting components for fatigue usage factors are 'the upper shell and the spray
- nozzle, which are calculated to be 0.97 and 0.99 respectively.
These
- remain, however, within the ASME acceptance criteria of 1.0 and are, therefore, acceptable to the staff.
Reactor coolant pump hydraulics and motor adequacy were reviewed for the proposed RTP conditions by Westinghouse.
The increased hot horsepower and stator temperature conditions are within the NEMA Class 8 limits.
A review of generic Reactor Coolant Pump stress reports for model 93A pumps by Westinghouse finds that all the design requirements provide adequate boundi.ng. of. the..
RTP-induced conditions and, therefore, the staff finds this acceptable.
Due to lower temperatures from the RTP program, the RCS will not expand as much as currently designed.
This will result in support gaps being present in locations that were previously zero.
The small gaps in the support structure may result in increased dynamic loading (both seismic and LOCA) in localized areas.
The overall LOCA loadings on the
- RCS, however, remain approximately the same for the following reasons:
1.
The lower RCS temperatures yield 1ower thermal loadings.
2.
The D.
C.
Cook Nuclear" Plant has a leak before break design methodology which allows the faulted condition evaluation to proceed without having to consider loadings from postulated breaks in the primary loop piping.
.The seismic margin available for this plant is also significant which means that there are no components in the system which are close to their allowable stresses.
Based on the above, the temperatures associated with the RTP rerating are, therefore, acceptable to the staff for the loop piping, the loop
- supports, and the primary equipment nozzles.
The effects of the D.C.
Cook Nuclear Plant RTP rerating on the operability and design basis analysis of the CRDM's of Unit 1 were reviewed.
The RTP rerating does not affect the operability or service duration of the CRDM latch assembly, drive rod, or coil stack.
The CRDM latch assembly and drive rod were originally designed for 650'F, and the design basis stress and fatigue calculations remain representative, for these components since the components are exposed to the hot leg temperature, which has not increased.
The coil stack is located on the outside of the pressure housing which is subject to ambient containment temperatures, which have not changed.
An evaluation was performed on the impact of the RTP rerated operating conditions on the structural analysis of the CROM pressure housing.
The component of the pressure housing which experiences the greatest stress range and has the highest fatigue usage factor is the upper canopy.
This is the pressure housing seal weld between the rod travel housing and the cap.
Westinghouse provided a review on the impact of the differences
-17" between the original normal and upset condition transients and those of the RTP on the code allowable stress levels and fatigue usage factors.
The results of the evaluation are:
1.
The maximum stress intensity r'ange is equal to 109,960 psi, which is less than the maximum allowable range of thermal stress of 127,105 psi'hich was previously found to be-acceptable.
2.
The total fatigue usage factor is equal to 0.672, which is less than the allowable limit of 1. 0 (ASME Section III, 1971 Edition).
The staff concludes, based on licensee evaluations, that the impact of the RTP program on the CRDM's is within design criteria and, therefore, is found to be acceptable.
- 2. 6 CONTAINMENT EVALUATION Short-Term Containment Res onse As part of the analysis to support RTP operation, the reactor cavity and loop subcompartments short-term pressurization in the event of a break of large coolant piping or a steam line was reanalyzed by Westinghouse.
In some of those
- areas, the analyzed pressure exceeded the structural limits as expressed in the FSAR.
These structures were reevaluated using the peak pressures obtained from the RTP analysis, WCAP 11902 (ref.2), to confirm that the acceptance criteria of Section 5.2.2.3 of the updated FSAR, titled "Containment Design Stress Criteria,"
were met.
The original design of the containment included a number of considerations of which the subcompartment pressure'es were but one.
For example, radiation shielding requirements may have dictated a thicker concrete slab than was necessary from a structural perspective.
The actual capacity is generally greater than the design pressures stated in the
- FSAR, and is further increased due to the fact that the materials used are stronger than the required minimum design strengths.
In the RTP structural review, advantage was taken of these greater capacities by performing manual or finite element evaluations of the affected structural elements.
The greater material strengths were used in the analysis where appropriate.
Loo Subcom artments The containment building subcompartments are the fully or partially enclosed spaces within the containment which contain high energy piping.
The subcompartments are designed to limit the adverse effects of a postulated high energy pipe rupture.
The results of the short term containment analyses and evaluations for the D. C.
Cook Nuclear Plant Unit 1 demonstrate that, for the pressurizer enclosure, the fan accumulator
- room, and the steam generator enclosure, the resulting peak pressures remain below the allowable design peak pressures.
For the loop compartments, the peak calculated pressures at the RTP rerated conditions are higher than the FSAR design allowables.
For these
- areas, structural evaluations were performed as discussed above for the revised peak pressures, and the structural adequacy of the containment subcompartments have been confirmed (Ref. 10) as follows:
Differential Pressure Node 1 or 6 to Node 25 This is the differential pressure from the reactor coolant loop compartments adjacent to the refueling canal nodes 1 or 6 across the operating deck to the upper containment.
Original Design pressure Original Calculated pressure New Calculated pressure II The licensee demonstrated the increased by review of existing computer analysis and reevaluation of the operating deck 1
3.6.6 psi 14.1 psi 18.7 psi differential pressure to be acceptable of the reactor coolant pump hatch covers oad carrying capacity.
Differential Pressure Node 2 or 5 to Node 25 This is the differential pressure across the operating deck from the reactor coolant loop compartments located 90 degrees from.the, refueling canal to the upper containment.
Original Design pressure Original Calculated pressure New Calculated pressure 12.0 psi 10.6 psi 13.0 psi The licensee demonstrates the increased differential pressure to be acceptable by comparison to Node 1 and Node 6 areas.
The slabs in both areas are the same.
Peak Shell Pressure This is the differential pressure across the containment shell to the outside, for,nodes located in the ice condenser inlet areas closest to the refueling canal.
Original Design pressure Original Calculated pressure New Calculated pressure 12.0 psi 10.8 psi 14.0 psi The licensee demonstrates the increased pressure to be acceptable by evaluation on a localized basis.
The containment shell can handle pressures well in excess of the overall 12 psi design pressure.
The average pressure over the structurally significant portion of the containment shell surrounding and including these nodes is smaller than the 12 psi containment shell design pressure.
Reactor Cavit The reactor cavity is the structure surrounding the reactor with penetrations for the main coolant piping.
This structure is designed to limit the adverse effects of the initial pressure response to a loss of coolant accident.
The results of the reactor cavity analysis and evaluations for the D.
C.
Cook Nuclear Plant Unit 1 demonstrate that, for the reactor vessel annulus and pipe
- annulus, the resulting peak pressures at the RTP rerated conditions are within the FSAR design allowables.
For the upper and lower reactor cavities the peak calculated pressures under RTP conditions exceeded the structural design pressures (Ref. 2, Sections 3.7.2 and 3.7.3) as stated in the FSAR.
For these
- areas, structural evaluations were performed for the revised peak pressures, and the structural adequacy of the containment subcompartment has been confirmed (Ref. 10) as follows:
Missile Shield Refuelin Canal Bulkhead Blocks and U
er Reactor Cavit Wall Differential Pressures The upper reactor cavity walls surround the reactor head.
The missile shields and the refueling canal bulkheads are blocks separating the upper reactor cavity from upper containment.
The missile shield is bolted down during operation, and is removable for refueling.
The refueling canal bulkheads fit snugly in grooves in the upper reactor cavity walls.
~Ci ll 1i 'li ii SM 1d and Bulkheads Original Design pressure 48.0 psi Original Calculated pressure 44.1 psi New Calculated pressure '8.4 psi 48.0 psi 44.1 psi,,
54.3 psi The licensee demonstrates the increased pressure for the cavity wall to be acceptable by finite element analysis of the entire upper reactor cavity wall.
The licensee has demonstrated the increased pressure for the missile shields and the bulkheads to be acceptable by manual calculation.
The test cylinder break strength of the concrete, which is higher than the design strength, was also taken into consideration.
Peak Lower Cavit Pressure This is the cavity located under the reactor vessel.
The peak pressure is used in the structural analysis rather than the differential pressure since most of the cavity walls are in the foundation mat.
Original Design pressure Original Calculated pressure New Calculated pressure 15.0 psi 13.8 psi 18.5 psi The licensee demonstrated that the increased pressures are acceptable by manual calulation.
The staff concludes, based on the licensee's demonstration, that the D.
C.
Cook Nuclear Plant's design basis pertaining to containment short term response, as stated in Chapter 5.2.7.3 of the FSAR, is adequate for RTP operation, and therefoi e, is acceptable.
The licensee must update the FSAR to reflect the higher structural design values.
Lon Term Containment Pressure The long term peak containment pressure analysis supports operation with the RHR crosstie valves closed at a power level of 3425 MWt for both Units 1 and 2 containment structure.
This analysis contained additional justification for operation under the RTP conditions (Ref. 11) and was approved by the staff Safety Evaluation dated January 30, 1989 (Ref. 12).
2.7 NUCLEAR, PROCESS AND POST-ACCIDENT SAMPLING SYSTEMS The Nuclear Sampling System (NSS) is designed to provide representative samples for laboratory.-analyses used to guide the operation of various primary and secondary systems throughout the plant during normal operation.
Since reduction of. sample pressure and temperature, when necessary, is already being done by heat exchangers and needle valves, the parameters associated with the RTP program do not affect the performance of the NSS.
With no power uprating, the source term remains unchanged.
Therefore, the staff concludes that operation under RTP conditions is acceptable for the NSS.
The staff finds that, since no power uprating is being proposed at this time, there is an insignificant effect on the post-accident containment thermal conditions and therefore the existing post-accident sampling system remains adequate and is acceptable.
Operation under RTP conditions results in slight reductions in secondary side, temperatures and pressures'ith no change in the source term.
The staff concludes that the change can be accommodated by the process sampling system without causing degradation of their performance, and is, therefore, acceptable.
2.8 ELECTRIC SYSTEMS DESIGN Operation under RTP conditions results in minor changes to the heat balance.
The only impact noted on the electrical systems is the slight increase in motor current for the motors used as prime movers of primary coolant.
The required power is increased by the higher densities encountered idue to the RTP program.
The licensee has reviewed cable penetrations,
- busses, and motor ratings to conclude that there fs sufficient design margin to handle the increased load.
The staff finds, based on the licensee's evaluation, that the proposed RTP program minimally affects the electric power system and associated loads and is therefore, acceptable.
3.0 TECHNICAL SPECIFICATIONS 1.
Definition 1.38 on design thermal power is being deleted on page 1-7 of the Technical Specifications (TS's) because there is no longer a single design thermal power at which all the transient and accident analyses have been performed.
The licensed power level for Cook 1
remains 3,250 MWt.
This change is acceptable.
2.
3.
Table 1-3 on page 1-10 is being deleted because it previously gave information on the analyses performed at the design thermal power.
This change is acceptable because the definition of design thermal power is being deleted also.
C Figure 2.1-1 on page 2-2 is being revised to reflect the revised DNBR safety limit of 1.45.
This change is acceptable because it is supported by the safety analysis.
4.
The pressurizer pressure low setpoint (Item 9 of Table 2.2-1 on page 2-5) is increased by 10 psig.
This is acceptable because it was assumed in the large-and small-break LOCA analyses.
- 3. 0 TECHNICAL SPECIFICATIONS 2,'.
4.
5.
6.
7.
Definition 1.38 on design thermal power is being deleted on page 1-7 of the Technical Specifications (TS's) because there is no longer a single design thermal power at which all the transient and accident analyses have been performed.
The licensed power level for Cook j.
remains 3,250 HMt.
This change is acceptable.
Table 1-3 on page 1-10 is being deleted because it previously gave information on the analyses performed at the design thermal power.
This change is acceptable because the definition of design thermal power is being deleted also.
Figure 2.1-1 on page 2-2 is being revised to reflect the revised DNBR safety limit of 1.45.
This change is acceptable because it is supported by the safety analysis.
The pressurizer pressure low setpoint (Item 9 of Table 2.2-1 on page 2-5) is increased by 10 psig.
This is acceptable because it was assumed in the large-and small-break LOCA analyses.
The Overtemperature-Delta T trip setpoint equation (pages 2-7 and 2-8) i.s being revised in terms of rated 'thermal power rather than design thermal power.
In addition, this revised OTDT trip setpoint protects the core safety limits of Figure 2. 1-1.
This change is acceptable because it is supported by the non-LOCA safety analyses.
The Overpower-Delta T trip setpoint equation'page 2-9) is being revised to reflect the revised core safety limits of Figure 2.1-1.
This equation is also being defined in terms of the indicated T
at rated thermal power.
These changes are acceptable because th) are supported by the safety analysis for the RTP program.
Technical Specification 3.2.2 on page 3/4 2-5 is being revised from a maximum F of 2.10 to 2.15.
This change is acceptable because it is supportetl by the large-break LOCA analysis.
The F
values for Exxon fuel are being deleted because this fuel will n3 longer be used at Cook Unit 1.
8.
The K(Z) curve applicable to Exxon fuel (page 3/4 2-7) is being deleted.
This is acceptable because Exxon fuel will no longer be used at Cook Unit l.
9.
10.
The K(Z) curve for Mestinghouse fuel (page 3/4 2-8) is being revised.
This is acceptable because it is supported by the new LOCA analysis for Cook Unit 1.
The F-Delta H limit applicable to Exxon fuel (page 3/4 2-9) is being deleted.
This is acceptable because Exxon fuel will no longer be used at Cook Unit l.
Table 3.2-1 on page 3/4 2-14 on DNB parameters is being revised.
T must be less than or equal to 570.9'F, the pressurizer
"22-pressure must be less than or equal to 2050 psig, and the reactor coolant system total flow rate must be greater than or equal to 366,400 gpm.
These changes are acceptable because they reflect the safety analysis for the RTP program.
Technical Specification 3.2.6 on page3/4 2-15 is being revised to change F
in the APL limit to 2.15.
This change is acceptable because
$t reflects the new F limit of Specification 3.2.2.
The limits on APL applicable to Exxon fuel are being deleted because Exxon fuel will no longer be used at Cook Unit 1.
Functional Units 2 and 11 of Table 3.2-2 on page 3/4 3-10 are being changed.
Functional Unit 2 incorporates an editorial change to indicate that the response time is applicable to both the high and low setpoints of the Power Range Neutron Flux trip.
This change is acceptable because it is editorial in nature.
Functional Unit ll is being changed from a response time of "not applicable" to "equal to or less than 2 seconds.".
This is acceptable because this trip on pressurizer water level-high was modeled in the analysis of the control,rod withdrawal-at-power event.
Functional Units 1.f and 4.d of Table 3.3-4 on pages 3/4 3-24 and 3/4 3-26 are being changed to decrease the steamline pressure low setpoint by 100 psig.
These changes are acceptable because they are supported by the steamline break analysis and the steamline break mass and energy evaluations.
Technical Specification 3.4.4 on page 3/4 4-'6 is being revised to 92K of span.
This change is acceptable because it is supported by the safety analysis.
Technical Specification 3.5.l.b on page 3/4 5-1 is being revised from an accumulator borated minimum water volume of 929 to 921 cubic feet.
This change is acceptable because it is consistent with the LOCA analysis for Cook Unit l.
Surveillance Requirement
- 4. 5. 2.f is being revised to reduce the discharge pressure of the safety injection pump and the residual heat removal pump.
These changes are acceptable because they are consistent with the LOCA analyses.
Surveillance Requirement
- 4. 5. 2. h is'eing revised by adding a
requirement to verify that the charging pump discharge coefficient is within a specified range following ECCS modifications.
The footnote is broken into four parts for clarity.
This change is acceptable because it ensures that the flow delivered to the core by the charging pumps in the event of a LOCA is within the analyzed values.
V Surveillance Requirement 4.7.1.2 on page 3/4 7-5 is being revised to change the discharge pressure requirements of the motor and turbine driven auxiliary feedwater pumps to 1375 psig and 1285 psig, respectively.
This corresponds to a 5X degradation of the pumps from the manufacturer's pump head curve.
These changes are acceptable because they are consistent with the changes for the RTP program.
20.
Basis page B 2-l(a) is being changed to incorporate the design limit and safety analysis limit DNBR values.
The DNB limits for Exxon fuel are being deleted since Exxon fuel is no longer used at Cook Unit 1.
The design limit and safety analysis limit DNBR values are acceptable because they are consistent with the RTP program.
21.
Basis page B 2-2 is being revised to delete reference to F-Delta H
for Exxon fuel and to design thermal power.
These changes are acceptable because references to both items have been deleted in the Specifications.
22.
Bases page B 2-4 is being revised to reflect the changes to the Overtemperature-Delta T trip function.
The changes are acceptable, because they reflect'changes made to the Specifications.
23.
Bases page B 2-5 is being revised to reflect the changes to the Overpower-Delta T trip function and the pressurizer water level-high trip.
These changes are acceptable because they reflect changes to the Specifications.
24.
Bases page B 3/4 2-1 is being revised to replace the minimum DNBR value of 1.69 by the words "the safety limit DNBR".
This change is acceptable because it will avoid changes to 'the Bases if the safety limit DNBR value is changed.
25.
Surveillance Requirement 4.1.1.5.b is being changed to require T determination of T every 30 minutes when the reactor is critiBF and T is less tQII 545'F:
This change is supported by Reference 9 and Fflows a full power T of 550'F for Cook Unit 1 Cycle 11 without requiring a monitor8) every 30 minutes while at full power, which the previous value of 551'F would have required.
This change is acceptable because the intent of maintaining the minimum coolant temperature for criticality of Specification 3.1.1.5 is preserved.
4.0 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
~Re later on June 9, 1989(
Accordin917, based
~upon t e environmental assessment, we have determined that the issuance of he amendment will not have a significant effect on the quality of the human environment.
5.0 CONCLUSION
The staff has reviewed the request by the Indiana and Michigan Power Company to operate the Donald C.
Cook Nuclear Plant Unit 1 at the reduced temperatures and pressures of the RTP program.
Reactor operation is restricted to an upper limit on T of 567.8'F because the steamline break mass and energy release inside conPNnment was not reanalyzed as part of the RTP program.
Although the
safety analysis was performed at power ratings which would support a possible power uprating for Cook Unit 1, power uprating is not addressed in the staff's review.
The power of D.C.
Cook Nuclear Plant Unit 1 is limited to the present rated thermal power of 3250 MWt.
Based on its review, the staff concludes that appropriate material was submitted and that normal operation and the transients and accidents that were evaluated and analyzed are acceptable.
The Technical Specifications submitted for this license amendment suitably reflect the necessary modifications for the operation of Cook Unit l.
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 operation in the proposed
- manner, and (2) such activities will be conducted in compliance with the Commission s regulations, and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: June 9,
1989 Principal Contributors:
Dan Fieno John Stang, NRR Anthony Gody, NRR
4
6.0 REFERENCES
I 1.
Letter (AEP:NRC:1067) from M.
P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated October 14, 1988.
2.
"Reduced Temperature "and Pressure Operation for Donald C.
Cook Nuclear Plant Unit 1 - Licensing Report,"
D. L. Cecchett and D.
B. Augustine, MCAP-11902, October 1988.
3.
Ellenberger S.L., et al., "Design Bases for the Thermal Overpower-Delta T
and Thermal Overtemperature-Delta T Trip Functions,"
MCAP-8746, March 1977.
4.
- Chelemer, H.; Boman, L. H.; Sharp, D. R., "Improved Thermal Design Procedures,"
MCAP-8567, July 1975.
5.
Butler, J. C.,
and Love, D.S., "Steamline Break Mass/Energy Releases for Equipment qualification Outside Containment,"
MCAP-10961, Rev.
1 (proprietary) and WCAP-lll84 (nonproprietary),.October 1985..
6.
Morita, T., et al., "Dropped Rod Methodology for Negative Flux Rate Trip Plants,"
WCAP-10297-P-A (proprietary) and MCAP-10298-A (nonproprietary),
June 1983.
7.
Letter (AEP:NRC: 1067B) from M.
P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated February 6, 1989.
8.
"American National Standard for Decay Heat Power in Light Mater Reactors,"
ANSI/ANS-5.1-1979, August 1979.
9.
Letter (AEP:NRC: 1067A)'rom M.
P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated December 30, 1988.
10.
Letter (AEP:NRC:1067C) from M.
P. Alexich (Indiana and Michigan Power Company) to the USNRC, dated March 14, 1989.
ll.
Letter (AEP:NRC:1024D) from M.
P. Alexich to T.
E. Murley (NRC), dated August 22, 1988.
Includes MCAP-11908, "Containment Integr ity Analysis for Donald C.
Cook Nuclear Plants, Units 1 and 2."
12.
Letter, J.
F. Stang (NRC) to M.
P. Alexich (IMECo), dated January 30, 1989.
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