ML17334A966
| ML17334A966 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/21/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17334A965 | List: |
| References | |
| NUDOCS 8605300659 | |
| Download: ML17334A966 (25) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
82 TO FACILITY OPERATING LICENSE NO.
DPR 74 INDIANA AND MICHIGAN ELECTRIC COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NO.
2 DOCKET NO. 50-316 Introduction By letters dated March 14 and March 27, 1986, the Indiana and Michigan Electric Company (the licensee);
proposed changes to the Donald C.
Cook Nuclear Plant, Unit No. 2.
These changes are grouped and evaluated below by those changes re-lated to the Unit 2 cycle 6 reload and Technical Specification changes.
C cle 6 Reload The reactor core for D. C.
Cook Cycle 6 will contain 191 Exxon fuel assemblies and one Westinghouse fuel assembly each having a
17 x 17 fuel rod array.
Eighty eight of the Exxon fuel assemblies are new.
Cycle 6 burnup has been projected to be 17,790 Mwd/MTU at a core power of 3411 MWT.
The design characteristics of the Exxon fuel aspemblies were reviewed and approved by the staff for the cycle 5 core.
The Exxon fuel in the cycle 6 core will be of'he same design as that of cycle 5.
As additional confirmation of the integrity of Exxon fuel remote visual examinations of irradiated fuel were performed following cycle 4.
No evidence of wear, fretting or other physical damage was noted for this fuel.
The exposure for the examined fuel ranged from 16,440 MWD/MTU to 17,630 MWD/MTU.
In anticipation of steam generator tube plugging the cycle 6 safety analyses were performed with an average steam generator tube plugging of 10%.
The effect of asyometric tube 'plugging was considered in the analyses.
Fuel Thermal-Mechanical Desi n
Cycle 5 contained a mixture of Exxon and Westinghouse fuel elements.
Staff conclusions regarding the thermal-mechanical design of the cycle 5 core remain applicable to cycle 6.
In particular cladding strain, external corrosion (oxidation), fuel rod internal pressure, and fuel rod pellet temperature were analyzed using the RODEX 2 code which has been approved by the NRC staff.
The analytical results satisfied the acceptance criteIia.
Collapse of fuel cladding into a pre-existing axial gap produced by the differential pressure between the reactor coolant pressure and the internal fuel rod pressure was investigated.
It was determined that the cladding would not collapse.
8605300659 Sb052i PDR ADOCK 050003ih P
"2-The UO centerline temperature was evaluated and it was concluded that fuel centerfine melting will not occur.
The structural adequacy of the fuel assemblies from loadings produced by seismic and LOCA events was evaluated and found to be acceptable.
These evaluations for the Exxon fuel in the cycle 5
core apply to the Exxon fuel in the cycle 6 core since it is of the same design and was evaluated using the same methodology.
Thermal H draulic Desi n
Cycle 6 was evaluated for steam generator tube plugging up to an average of 10K for the four steam generators.
The steam generators are responsible for approximately 30K of the frictional pressure drop within:the cooling loops.
Plugging 10K of the steam generator tubes would result in a coolant flow decrease of approximately 3X.
The increased frictional pressure drop and reduced coolant flow rate was utilized by the licensee in evaluation of design basis transients and accidents..
Cycle 5 was a mixed loading of Westinghouse and Exxon fuel.
A 2X penalty was applied to the calculated DNBR to account for uncertainties in the calculation from the mixed loading.
The cycle 6 core is all Exxon fuel except for one Westinghouse fuel assembly which is not in a limiting power location nor is it adjacent to Exxon fuel in a limiting location.
The 2X penalty is therefore removed.
The staff reviewed and approved Exxon methodology for computing the effect of fuel rod bowing on the core heat, transfer which is described in Exxon topical report XN-75-32(P)(A).
Fuel rod bowing was determined to not affect core DNBR for anticipated operational occurrences for fuel burnups less than 43,000 MWD/NTU.
This is a greater burnup than is anticipated for any fuel assembly in cycle 6 and therefore no rod bow penalty is imposed.
Nuclear Desi n
The neutronic characteristics of the cycle 6 core are designed to be similar to that of cycle 5.
The design methods are described in Exxon topical report XN-75-27(A) which has been reviewed and approved by the NRC staff.
The maximum boron concentration required to prevent criticality during cycle 6 was determined to be 1451 ppm.
This is exceeded by the minimum concentration required by the Technical Specifications for the refueling water storage tank.
Power distributions have been obtained with the three-dimensional quarter core XTGPWR code which has been approved by the staff.
The expected total peaking factor, along with values of the moderator, and Doppler temperature coefficients, boron worth, delayed neutron fraction and shutdown margin, presented for beginning and end of cycle 6 at full and zero power conditions, are conservative with respect to those used in the transient and accident analysis.
They are similar to those quantities for cycle 5.
Beginning and end-of-cycle radial power distributions are presented.
These indicate that the values for total peaking factor and maximum relative pin power should remain within limits during cycle 6.
Power distribution control during the cycle will be accomplished by following the approved procedures described in Exxon topical report XN-NF-77-57(A).
We conclude that the nuclear design of the cycle 6 reload is acceptable.
Non LOCA Transients and Accidents The design basis for D.
C.
Cook Unit 2 includes the evaluation of a number of postulated transients and accidents. 'he assumed failures and operable protection equipment for these events were reviewed and approved by the staff as described in the original SER and its supplements.
The safety analyses of the consequences from these events for the first fuel cycle were performed by Westinghouse, the reactor vendor.
The safety analyses for the cycle 4, 5 and 6
cores was performed by Exxon which supplied the fresh fuel assemblies for these loadings.
In the staff safety evaluations for the cycle 4 and 5 cores deficiencies in Exxon methodology for calculating increases in reactor system pressure and core DNBR during transients and accidents were noted.
These have now been corrected.
Exxon has provided reanalysis for the design basis transients and accidents for D.C.
Cook Unit 2 Cycle 6 with the exception of steam line break.
Return to criticality following a postulated main steam line break at D.
C.
Cook Unit 2 is prevented by the flushing of the Boron Injection Tank (BIT) into the reactor system.
The minimum DNBR was evaluated in detail for Cycle 2 with Westinghouse fuel and for Cycle 4 with Exxon fuel.
The minimum DNBR remained above the acceptance criterion for fuel failure.
The licensee did not reanalyze main steam line break for initial Cycle 6 operation.
This is acceptable based on similarity to previous cores and the commitment by the licensee to provide analyses of steam line breaks during mid cycle.
Steam line breaks are overcooling events for which limiting conditions are not reached until the end of core life. Exxon utilized the following methodology for analysis of non LOCA <ransients and accidents.
a.
b.
The PTSPWR2 computer code is described in topical report XN-NF-74-5(P)
Revision 2 "Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Hater Reactors (PTS-PWR)"
and supplements I through 6 of the topical report.
The PTSPWR2 code is utilized to predict reactor
'ystem temperatures and pressures following anticipated transients and accidents which do not produce steam within the coolant loops.
Staff review of the code is complete and documentation granting approval is being developed.
The SLOTRAX-ML computer code is described in Topical Report XN-NF-85-24(P)
" SLOTRAX-ML: A Computer Code for Analysis of Slow Transients in PWRs".
The SLOTRAX-ML code is utilized for long term mass and energy balance calculations following loss of feedwater/feedwater line break.
It is also utilized to calculate the pressurizer surge line flow rate for these events.
Staff review of the code is complete and documentation granting approval is being developed.
C.
The XCOBRA-IIIC computer code is described in topical report XN-NF-82-21(P)(A) "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations."
The XCOBRA-IIIC code is utilized by Exxon to calculate DNBR within the reactor core.
The code utilizes reactor system temperatures and pressures from anqlyses using PTSPWR2.
The code was reviewed and approved by the NRC staff.
d.
The XTGPWR computer code is described in topical report XN-CC-28, Rev.
5 "XTG-A Two Group Three Dimensional Reactor Simulation Utilizing Coarse Mesh Spacing".
The XTGPWR code is utilized to calculate power distributions for transients involving spatial reactivity abnormalities within the reactor core, including control rod ejection.
The code was reviewed and approved by the NRC staff.
In the Standard Review Plan NUREG-800 acceptance of the predicted consequences from design basis events is based on the anticipated likelihood of occurrence.
Two types of events are required to be analyzed:
(1)
Those incidents which might be expected to occur during the lifetime of the reactor (anticipated transients) and (2)
Those incidents not expected to occur which have potential to result in significant radioactive material release (accidents).
Analysis of abnormal anticipated transients have been submitted which show that the integrity of the reactor coolant boundary is maintained and that the minimum DNBR is above the applicable limit.
Thjs limit is 1.17 using Exxon methodology.
Anticipated transients can be classified as reactivity insertions, increased cooling, under-cooling, loss of forced flow and changes in coolant inventory.
There are a number of postulated accidents which are not expected during the life of the plant. Within this group are events which form the design basis for the various barriers and safety systems.
Some degree of fuel damage may be calculated for accidents; however the offsite dose consequences must be demonstrated to be below the guideline doses of 10CFR 100.
The following is the staff evaluation of the abnormal anticipated transients and postulated accidents which form the design basis for D.C.
Cook Unit 2.
Increased Coolin Transients The licensee evaluated the following events that produced increased primary system cooling.
(1)
(2)
(3)
(4) decrease in feedwater temperature increase in feedwater flow increase in steam flow (excess load) inadvertent opening of a steam generator relief or safety valve.
The most severe of the above transients in terms of DNBR are excessive load increase events.
The minimum DNBR was calculated to be 1.547 for an excess load event at the beginning of the core cycle with the control rods in automatic.
The PTSPWR2 code was used to evaluate core flows and pressure and the XCOBRA-IIIC code was used to predict minimum DNBR for the event.
The analyses were made conservative by biasing the initial reactor system temperatures and pressures to reflect instrument error so that the minimum value of DNBR would be produced.
Since the minimum value of DNBR is greater than the acceptance limit the analytical result is acceptable.
Under-Coolin Transients The licensee evaluated the following events which result in a decrease in heat removal by the secondary system.
1) loss of external load, turbine trip 2) loss of normal feedwater flow 3) loss of condenser vacuum 4) closure of main steam isolation valve 5
loss of non emergency AC power The loss of external load event was determined to produce the highest reactor system pressure.
The maximum reactor system pressure was determined to be 2620 psia using the PTSPWR2 code.
Control systems which would act to limit overpressure were not assumed to operate.
The maximum pressure from this event is acceptable since the staff's acceptance limit of 110$ of design pressure is not exceeded.
The minimum DNBR was determined to occur for a loss of external load event if the action of control grade systems to reduce reactor system pressure were maximized.
The positive moderator coefficient was biased upward to increase reactor power during the event.
Core power reached 1205 and the minimum DNBR reached 1.263 which is above the acceptance limit of 1. 17.
In order to determine the ability of the auxiliary feedwater system to provide adequate removal of the heat generated by the core for an extended period of time, the licensee analyzed a loss of main feedwater event for 10,000 seconds using the SLOTRAX-NL code.
Only the lowest capacity of the 3
AFW pumps was assumed to be operable.
.The single pump was determined to provide adequate cooling.
Long term decay heat generation was determined using the ANS-5. 1 decay heat standard including the contribution from decay of elements in the Actinide group.
The peak reactor system pressure following loss of feedwater was determined by inputting the surge line flow calcualted by SLOTRAX-ML into the PTSPWR2 pressurizer model.
This methodology conservatively maximizes the calculated result.
The resulting reactor system pressure was calculated to be less than llOX of design and is therefore acceptable.
"6-Loss of Forced Coolant Flow Transients The licensee evaluated various loss of reactor coolant system forced flow events and determined that the limiting loss of flow transient would result for the tripping of all four reactor coolant pumps while the reactor, was at full power.
The licensee analyzed the minimum DNBR that could occur from this event using the PTSPWR2 code to evaluate the'ore fluid conditions and the XCOBRA-IIIC code to evaluate fuel pin heat flux. Initial conditions and calculational assumptions were selected so as to be conservative for the DNBR calculation.
The DNBR was calculated to not decrease below 1.27 which is acceptable.
Below the P-7 interlock setpoint the reactor does not automatically trip on loss of forced coolant flow.
The P-7 setpoint can be as high as 20X power when instrument error is taken into account.
The licensee evaluated the consequences of loss of flow below the P-7 setpoint and determined that the reactor would be tripped either by a non-safety grade trip or by operator action shortly after forced flow was lost.
The non-safety grade trip would occur since loss of power to the coolant pumps would also cause power to be lost to the control rod drives.
The control rods would then drop into the core by gravity.
In addition plant procedures require the, reactor to be tripped on loss of forced coolant flow.
In the elapsed time before reactor trip, the minimum DNBR calculated was 1.8, which is acceptable.
Complete coastdown of the coolant pumps was assumed so that only natural circulation would exist in the reactor system.
Control Rod Withdrawal Events The licensee analyzed inadvertent control rod bank withdrawal at initial power conditions ranging from just critical to full power.
The control rod drive mechanisms are wired into preselected bank configurations which are always moved in a programmed sequence.
The location of the individual control rod assemblies throughout the core and the sequential movement of the banks is designed to limit core power peaking during control rod movement.
Control rod bank withdrawal transients were analyzed using the PTSPWR2 and the XCOBRA-III computer codes.
For control rod withdrawal when the reactor is just critical the licensee concluded that the core DNBR would remain above the 1. 17 design limit for Exxon fuel provided that the in'itial primary system pressure was above 2175 psia and the initial reactor system temperature was below 549 F.
The calculation was made conservative by utilizing a low value of assumed initial power at the just critical initial condition and by performing a sensitivity study to select the most conservative value of delayed neutron fraction throughout core life.
The analyses assumed a minimum of 3 reactor coolant pumps in operation and that the reactor would not be tripped by the source or 'intermediate neutron flux channels.
The power range channels (low range) were assumed to trip the reactor and terminate the transient.
The low power range trip setting was assumed to be at 35K power which is conservatively high.
The technical specifications are being updated to ensure that the reactor system temperature, pressure and coolant pump operation are consistent with the assumptions of this
"7" analysis.
Either a minimum of three reactor coolant pumps will be operating or the control rod drives will be deenergized so that control rod withdrawal cannot occur during hot standby or shutdown operation.
For inadvertent control rod withdrawal at power the reactor would trip on overtemperature delta-T or high power.
The high power range trfp was assumed to be at 118K power.
A spectrum of control rod bank withdrawal events were analyzed at 9X, 60X and 100K, of rated power.
The minimum calculated DNBR was
- 1. 196 which is above the minimum design limit of 1. 17 and is therefore acceptable.
The licensee also evaluated other reactivity transients involving a dropped control rod assembly, a dropped control rod bank, a misaligned control rod assembly and single control rod assembly withdrawal.
These events were evaluated using the XTGPWR code to compute the perturbed three dimensional power distribution.
The DNB calculations were performed using the XCOBRA-IIIC code.
With the exception of the single control rod withdrawal event none of these occurrences would produce DNBR of less than the minimum acceptance criterion of 1. 17.
No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single rod assembly of an inserted bank.
The event could only occur as the result of multiple wiring failures or deliberate operator disregard for procedures and event indication.
The staff has determined that single control rod withdrawal events are not anticipated operational occurrences but should be reviewed under the acceptance criteria for design basis accidents which allow for some fuel damage.
The analysis of a single control rod withdrawal at power was performed using the XTGPWR code to evaluate radial power distribution which was coupled with the design axial power distribution to evaluate DNBR.
This approach is conservative since rod withdrawal would reduce axial peaking.
This reduction was not credited in the analysis.
The percent of fuel failure was calculated using the XNB critical heat flux correlation statistics by which the failure probability of the fuel rods is a function of the DNBR.
Using this methodology 3.7X of the fuel rods were assumed to fail.
The offsite radiation dose from this amount of assumed fuel failure was calculated using assumptions consistent with those of the FSAR.
The resulting offsite dose was determined to be well within the guidelines of 10CFR100.
The staff has not reviewed Exxon methodology for calculating fuel failure by statistical methods.
The amount of fuel failure which could occur from single control rod withdrawal would be limited to the fuel in that immediate vicinity of the affected control rod and would be less than the 10.4X of fuel failure calculated for a rod ejection accident using approved methodology.
The staff determined that the offsite dose consequences'would remain within the 10 CFR Part 100 guidelines even if the larger fraction of fuel failure was assumed.
Fuel misloading errors were evaluated in three categories A) exchanging an exposed assembly with another exposed assembly.
B) exchanging a fresh assembly with an exposed assembly and C) exchanging a fresh assembly with another fresh assembly with burnable absorber rods.
Core power distributions were calculated using the XTGPWR code.
The more significant misloading events would be detectable from deviations in the incore detector readings during power ascension.
Those events which could not be detected would not produce power maldistributions sufficient to produce fuel damage.
Inadvertent Boron Dilution Dilution of the boric acid concentration within the reactor core adds reactivity to the reactor core.
During power operation inadvertent dilution would lead to reactor trip in a manner similar to withdrawal of a control rod.
During shutdown and hot standby inadvertent dilution could eventually produce criticality.
To protect against inadvertent criticality from boron dilution D.C.
Cook Unit 2 is designed so that at least 15 minutes would be available to alert the operator following initiation of the event.
The occurrence of inadvertent dilution would be made known to the operator by alarms indicating deviation from the preset dilution flowrate, CVCS valve and pump status lights, pressurizer and volume control tank levels.
In addition a neutron level alarm is available from the source range channels except during startup operations.
The licensee evaluated boron dilution both with the reactor coolant pumps running and during RHR system operation without the reactor coolant pumps running.
When the coolant pumps are not running the core is more sensitive to boron dilution since a smaller volume of water is in circulation.
To ensure that the 15 minutes criterion is maintained, the required shutdown margin is increased for RHR operation.
Since the reactor is more sensitive to boron dilution at higher boron concentrations, the minimum shutdown margin is a function of boron concentration.
A curve providing the minimum required shutdown margin as a function of boron concentration will be added to the Plant Technical Specifications.
The staff concludes that the protection from inadvertent boron dilution is consistent with the original approved design of the plant and is acceptable.
Feedwater S stem Pi e Break Accidents The principal concern from rupture of a main feedwater pipe outside of containment is overpressure of the reactor system.
Overpressure is caused by the mismatch between reactor core heat generation and steam generator heat removal.
The licensee calculated the maximum pressurizer pressure that would occur following the rupture of a main feedwater pipe using the pressurizer model from the PTSPWR2 code.
Surge flow into the pressurizer was calculated using the SLOTRAX-ML computer code.
This methodology is conservative since the surge flow is maximized in SLOTRAX-ML and the pressurizer pressure is maximized in PTSPWR2.
The maximum pressurizer pressure of 2600 psia, was calculated to occur 6.4 seconds following initiation of the event.
The reactor vessel pressure would be approximately 50 psi higher than the pressurizer but would still be below the acceptance criteria of 110K of design pressure.
The auxiliary feedwater pumps are protected against runout by automatic valves in the discharge lines which would throttle the flow to within design limits.
The SLOTRAX-ML analysis of the feedwater line break event predicts that if the operator took action to direct auxiliary feedwater flow to the intact steam generators within 10 minutes that adequate feedwater would be provided even
'assuming single failure of the largest pump.
The staff believes that adequate time is available for this action. Offsite dose consequences following a main feedwater line break would be well within the Part 100 guidelines since no fuel failure would occur.
The staff concludes that the predicted consequences from the postulated feedwater line break accident are acceptable.
Reactor Coolant Pum Rotor Seizure Accident Seizure of a reactor coolant pump rotor would cause a rapid reduction in reactor coolant system flow.
The reactor would be tripped by a low reactor coolant system flow signal.
Before the heat transfer within the core could be reduced by the reactor trip, core heat up and damage from boiling transition is of concern.
The staff assumes that any fuel which is calculated to have a
DNBR of less than the acceptance criteria of 1. 17 fails and releases fission products to the reactor coolant.
The licensee calculated the consequences from a locked rotor event using the PTSPWR2 computer code to describe the reactor system fluid transient and the XCOBRA-IIIC code to calculate DNBR.
The calculation was made conservative by assuming pump failure in a coolant loop with no steam generator tube plugging.
Plugging was assumed in the other loops to maximize the fraction of coolant flow that would be affected.
Assumptions were made to minimize reactor system pressure which is conservative since DNBR is also minimized.
The minimum DNBR for this postulated accident was calculated to remain above
- 1. 17. Therefore no fuel failure would occur.
Control Rod E ection The mechanical failure of a control rod mechanism pressure housing would result in the ejection of a control rod cluster control assembly.
For a fully inserted control rod assembly the consequences would be a rapid reactivity insertion and possibly local fuel rod damage.
The licensee analyzed the consequences from a postulated control rod ejection accident using methodology which has been approved by the NRC staff including the multi-dimensional XTRAN code to solve the space and time dependent neutron diffusion equation.
The maximum energy deposition in the UO fuel was calculated to be 161.9 cal/gm.
This is less than the staff makimum acceptance criterion of 280 cal/gm and is therefore acceptable.
Energy depositions below the staff's criterion ensure that gross core damage will not occur.
The reactor system pressure during the rod ejection event was calculated using the PTSPWR2 code.
The energy generation calculated by the point-kinetics model of PTSPWR2 was benchmarked against that calculated by the two dimensional XTRAN code and determined to be conservative.
The maximum pressurizer pressure was calculated for both beginning of cycle and end of cycle.
The beginning of cycle result was determined to be bounding.
The pressurizer pressure at beginning of cycle was adjusted to the maximum reactor system pressure location at the pump discharge and determined to be 2747.6 psia which is less than 110%
of design.
Regulatory Guide 1.77 recomnends that the maximum pressure be less than that which would cause stresses to exceed "Service Limit C'l of the ASME Code.
"Service Limit C" corresponds to approximately 120$ of design pressure.
The licensee's pressure calculation, is therefore acceptable.
The number of fuel rods experiencing DNBR less than the
- 1. 17 design limit for EXXON fuel was calculated to determine the offsite dose consequences from the event.
This was accomplished using the XCOBRA-IIIC code to determine the local power level at which the minimum DNBR would be 1. 17.
A flow penalty was in-cluded to account for non-synmetric effects on core flow from rod ejection.
Using the calculated radial power distribution for the XTRAN calculations the fraction of the core with local power levels greater than that which would produce a
DNBR of less
- 1. 17 was determined.
From this calculation 10.7% of the fuel was predicted to penetrate the DNBR limit and to fail.
The offsite dose consequences from the event were calculated to be well within the exposure guidelines of 10CFR100 and are therefore acceptable.
Loss of Coolant Accidents Large break LOCA/ECCS analysis were performed in 1982 (1, 2) to support operation of the D.C.
Cook Unit 2 reactor at 3425 MWT with ENC fuel.
The limiting break was identified as the 1.0 double ended cold leg guillotine (DECLG) break as developed in reference 1.
The results of calculations with one and two 'LPS pumps operating were presented in reference 2 which indicated that a higher PCT occurred with two LPS pumps operation.
Reference 3 documents the results of LOCA/ECCS analysis performed in support of cycle 6 and future cycles with all ENC fuel at a thermal power rating of 3425 MWt, with up to 10% of the steam generator tubes plugged.
Calculations were performed for the previously identified 1.0 DECLG break, with full ECCS flow.
Three exposures using a center peaked axial power shape were studied to determine exposure dependence.
The exposures range from 2 MWD/kg to 47 MWD/kg peak rod average burnup.
The axial dependence of the peaking factor limit is denoted K(Z) and is defined as (K(Z) = FQ(Z)/MAX FQ(z) where FQ(Z) is the maximum peaking factor allowed at any elevation Z.
The topmost segment of the K(Z) curve is limited by the small break LOCA linear heat generation rate (LHGR) limits presented in the technical specifications.
Confirmation of the axial dependence is based on three power distributions:
a center peak chopped cosine power distribution and two conservative top skewed power shapes as presented in Figure 3.2*.
The power distributions are analyzed at the limiting exposure, 2 MWD/kg, where the peak stored energy occurs.
A summary of these results and the exposure study is presented in Table 3.5*.
The normalized K(Z) curve verses core axial height is presented in Figure 2. 1*.
The calculations were performed using the EXEM/PWR LOCA/ECCS models, including Reference 3, Exxon Report XN-NF-85-68(P) Revision 1.
fuel properties calculated at the start of the LOCA transient with the NRC generically approved RODEX2 code.(4).
The quench time, quench velocity and "CRF correlations in REFLEX and the heat transfer correlation in TOODEE2 are based on ENC's 17x17 Fuel Cooling Test Facility (FCTF) data (5, 6, 7).
Reference 3
reflects the revisions in the correlations based on the FCTF daga which are documented in References 6 and 7 and to reflect a
- 1. 1 multiplier on the peak power parameter used in the FCTF correlations in the REFLEX code.
This documents the NRC acceptance of the
- 1. 1 multiplier as applied and agreed to with EXXON Nuclear Company, Inc.
The NRC finds that the analysis results and methods suamarized above and as presented in reference 3 supports operation of the D.C.
Cook Unit 2 reactor for cycle 6, and future cycles with ENC fuel, at a total power peaking factor limit (Fg) of 2.10.
Considering the results of the analysis as summarized in Table 3.5*, peak clad temperature is less than 2200'F, local oxidation is less than 17K and core wide metal-water reaction is less than 1.0$.
Therefore we find that the criteria of 10CFR50.46 have been satisfied.
REFERENCES Loss of Coolant Accidents 1.
XN-XF-82-35, "Donald C.
Cook Unit 2 LOCA/ECCS Analysis Using EXEM/PWR Large Break Results,"
Exxon Nuclear Company, Inc., Richland, WA
- 99352, April 1982.
2.
XN-NF-82-35, Supplement 1, "Donald C.
Cook Unit 2 Cycle 4 Limiting Break LOCA/ECCS Analysis Using EXEM/PWR," Exxon Nuclear Company, Inc., Richland, WA 99352, November 1982.
3.
"Donald C.
Cook Unit 2 Limiting Break LOCA/ECCS Analysis, Revision 1, 10$
Steam Generator Tube Plugging, and K(z) Curve," XN-NF-85-68(P) Revision 1, Exxon Nuclear Company, Inc., Richland, WA 99352, April, 1986.'.
XN-NF-81-58(P)(A), Rev. 2, and Rev.
2 Supplements 1 and 2, "RODEX2: Fuel Rod Thermal-Mechanical
Response
Evaluation Model," Exxon Nuclear Company, Inc., Richaldn, WA 99352, February 1983.
5.
XN-NF-85-16(P), Volume 2; PWR 17x17 Fuel Cooling Test Program Reflood quench Carryover, and Heat Transfer Correlations,"
Exxon Nuclear Company, Inc., Richland, WA 99352, May 1985.
6.
7.
"PWR 17xl7 Fuel Cooling Test Program Reflood quench, Carryover, and Heat Transfer Correlations,"
XN-NF-86-16(P), Revision 1, and all supplements, Exxon Nuclear Company, Inc., Richland, WA 99352, January 1986.
"PWR 17x17 Fuel Cooling Test Program Sensitivity Studies,"
XN-NF-85-16(P),
Volume 1, and all supplements, Exxon Nuclear Company, Inc., Richland, WA 93352, January 1986.
Reference 3, Exxon Report XN-NF-85-68(P) Revision 1.
"12-TECHNICAL SPECIFICATION CHANGES Technical S ecifications A large number of Technical Specification changes have been proposed by the licensee.
Several of the changes are proposed as a result of the safety analysis for Cycle 6.
Other motivations for changes included:
1.
Correction of typographical errors and other editorial changes (clarification, etc).
2.
Removal of references to three-loop operation from the specifications.
Such operation is not permitted for the Cook reactors and the specifications are somewhat more lucid when such references are removed.
3.
Removal of the Axial Power Distribution Monitoring System (APDMS) from use and its replacement with Allowable Power Level (APL) control.
4.
Changes to bring the specifications into consistency with Rev.
4 of the Standard Technical Specifications.
5.
Changes due to incorporation of RDF RTDs and Foxboro transmitters.
6.
Separation of flow and
- FhH, Specifications.
7.
Clarification of the Differential Pressure Between Steam-Line-High ESF Actuation Signal.
8.
Revisions to the Bases The evaluation of Categories 1, 2, 4, 7 and 8 is organized by category.
The evaluation of the other Specifications is organized by specification number.
EDITORIAL CHANGES The licensee has proposed a number of charges as purely editorial in nature which have been reviewed and found to include changes for clarity, changes to remove unnecessary redundancy of requirements, correction of errors, and grammatical changes.
These changes are listed in the March 14, 1986 and the March 27, 1986 letters in'ufficient detail that listing them again in this evaluation is not deemed necessary.
In our review of the changes, two additional oversights came to our attention.
In the proposed change to Table 3.3-1, Item 20.B, the Standard Technical Specifications provide more clarity in the title than was proposed by the licensee.
We proposed to change the title for this clarity and the licensee agreed.
The title is changed from "Above P-7" to "Above P-7 and Below P-8".
There is no change to the intent or application of this technical specification by this change.
On page 3/4 3-5, of the technical specification in Action 2.b.,
we determined tha%-an additional clarification was overlooked in License Amendment No. 75, issued on August 26, 1985, which clarified Action 6.b.
These two statements are meant to be equal in requirements and as applied in plant operation, however, only Action 6.b was changed by Amendment 75.
We have proposed to update Action 2.b, and the license has agreed to the clarification.
With this change',
the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of the other channels per Specification 4.3. 1. l. 1.
These words are also consistent with the Standard Technical Specification.
I We have reviewed the proposed editorial changes to the Technical Specifications and together with the two clarifications as agreed to by the licensee, we find the revised Technical Specification acceptable.
REMOVAL OF 3-LOOP TECHNICAL SPECIFICATION The licensee has proposed to simplify the Technical Specifications by removing requirements for 3-loop operation.
D.C.
Cook is a 4-loop plant, i.e.,
there are four reactor coolant loops used in plant operation to transfer heat from the core to four steam generators.
During the plant design it was anticipated that licensees may have reason to need 3 loop operation (I loop inoperable) but at a reduced power level.
Technical Specifications were included in the D.C.
Cook Facility Operating License but because of the state of review and approval of 3-loop operation, a license condition was also included which prohibits 3-loop operation pending further review by the
- NRC, The licensee now proposes to remove the 3-loop Technical Specifications in order to simplify the documentation available to the operators.
Removal of these technical specifications will not affect operation or lessen any requirement for approved operation.
We have reviewed each of the proposed changes in this category and find them acceptable.
DIFFERENTIAL PRESSURE BETWEEN STEAM LINE HIGH ESF ACTUATION SIGNAL The licensee proposed to change a footnote on Table 3.3-3 to clarify the actions to be taken during Mode 3 when only three steam generators are operable.
The Differential Pressure Between Steam Line-High actuation of the Engineered Safety Feature (ESF) depends upon the comparison of pressures between steam lines. - In the above changes to remove 3-loop (reactor coolant loops) Technical Specification, the licensee also removed 3 Steam Generator Loop operation from Modes I and 2.
Three steam generator loop operation is allowed in Mode 3; however, with one of these loops out, the comparison of pressures between steam lines will produce half of the required ESF signal.
The current footnote can be misinterpreted to mean that under the condition of 3 steam loop operation, all the channels on the inoperable loop should be placed in the tripped mode.
This would create an ESF signal when none was required.
The licensee proposed to delete the footnote and add a new provision which requires that the channel on each of the operable loops which indicates high differential pressure with respect to the idle loops be placed in the tripped mode.
This action reduces the ESF actuation logic for the active loop differential pressures from 2 out of 3 to I out of 2 and thus permits 3-loop operation in Mode 3 since 2 channels per steam line are necessary for a trip.
In further discussions with the licensee, it was agreed to modify the language in the proposed footnote for clarity.
We have reviewed the proposed change and find that it does provide clarity to the intent of the Technical Specification and is therefore acceptable.
CHANGES FOR CONSISTENCY WITH STANDARD TECHNICAL SPECIFICATIONS The licensee has proposed three changes which are classified as consistent with the Standard Technical Specifications and although this may be correct, each of the changes was reviewed on its own merit.
The first change was to Table 3.3-1, Items 7 and 8.
The Action statements for Overtemperature and Overpower delta T channels inoperable has been changed from Action 2 to Action 6.
The only difference in these two actions is that Action 2 also has a requirement to either adjust power and neutronics instrumentation or to periodically monitor quadrant Power Tilt Ratio.
Neither of these actions is directly related to the Overtemperature or Overpower delta T.
This is correctly reflected in the Standard Technical Specifications.
Therefore, the licensee s proposed change is corrective of an error or an oversight which is correctly displayed in the Standard Technical Specifications.
The second change is to Table 3.3-1, Item 20.b and will add an exemption to Section 3.0.4 (prohibits changing mode if inoperable) for the Reactor Coolant Pump Breaker Position Trip channels.
The Reactor Coolant Pump Breaker Position Trip provides protection against DNB at reactor coolant flow rates above the P-7 interlock.
This interlock is enabled between 0 and illrated thermal power.
Technical Specification 3.4. 1. 1 requires all reactor coolant loops be in operation for Modes 1 and 2.
With all coolant loops in operation, there is more than. enough flow for DNB protection up to the P-7 interlock (13K rated thermal power) and the ESF actuation for DNB protection is not needed in Mode 1 until after the P-7 is enabled.
At that point, the Reactor Coolant Pump Breaker Position Trip channel must be in operation.
The proposed change to exempt Section 3.0.4 will allow entry into Mode 1 without these channels required operable but will not allow operation above P-7 interlocks without meeting the appropriate action statements.
This proposed change was also recognized in later. revisions to the Standard Technical Specificatiuns.
We find the proposed change to add an exemption to Section 3.0.4 for the Reactor Coolant Pump Breaker Position Trip to be acceptable.
The third change proposed by the licensee is to delete a footnote in the Technical Specification for Relief Valves, Section 3.4.11, which provides that no PORV report pursuant to Section 6.9. 1.9 is required if the PORV is closed due to an inoperable block valve.
Section 6.9. 1.9 was changed by Amendment 51 issued on February 7, 1983 and is no longer associated with PORV reports.
Amendment 58 issued November 23, 1983 added challenges to the PORV as part of the Annual Report and the reporting requirements of 10 CFR 50.72 and 50.73 are sufficient to cover equipment failure.
We find the proposed change to delete the footnote acceptable.
CHANGES TO BASES SECTION The licensee has proposed changes to the Bases section to support the proposed Technical Specification changes.
Although not part of the Technical Specifications, we have reviewed the Bases changes and find them acceptable.
OTHER SPECIFICATIONS Figure 2. 1-1 for four loop operation has been revised to contain new core safety limits based on safety analyses for lOX steam generator tube plugging.
The new limits are consistent with the analyses and are acceptable.
2.
Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints The overtemperature hT and overpower hT trips have been altered as a
result of the analyses with 10'team generator tube plugging and to account for the changeover to RdF RTDs.
The allowable values for the trip setpoints were revised to reflect the changes in uncertainties associated with the new RTDs.
This is acceptable.
The value of the T at rated power was altered to account for the lOX plugging analyses.
Phe new value is consistent with the analyses and is acceptable.
The values of the constants K
K
, and K
in the algorithm for the overtemperature bT trip were hltefed to r)fleet the 10K plugging analyses and the conversion to the new RTDs.
The new values are consistent with the analyses and are acceptable.
The definition for f(AI) has been revised in both the overtemperature bT and overpower hT trips.
This was required as a result of the cycle 6
safety analysis.
The new definition is consistent with the safety analysis and is acceptable.
The design flow value has been changed to reflect the analysis with 10K plugging and the conversion to new RTDs.
The new value is consistent with the analysis results and is acceptable.
The allowable values for the loss of flow and steam generator low-low level trips are changed to reflect a new Westinghouse analysis of allowances for these trips and are acceptable.
3.
Specification
- 3. 1.1. 1, Shutdown Margin The applicability of this specification has been restricted to Modes 1, 2, and 3.
The Mode 4 shutdown margin requirement is addressed in Specification
- 3. 1.1. 2.
An additional surveillance requirement,
- 4. 1. 1. 1. 3, has been imposed in order to be consistent with safety analyses.
This is acceptable.
"16-Specification 3.1.1.2 This specification has been expanded to include Mode 4 (see above).
This change was made as a result of a new analysis of the boron dilution event for Modes 4 and 5.
In both modes the requirements on operqbility of Reactor Coolant Loops and Residual Heat Removal Loops were altered to be consistent with the new analyses.
This is acceptable.
Specification 3.1.1.3 A footnote is added to this specification which exempts the RWST, as a
source of water for the Boron Dilution event provided it is maintained with a value and boron concentration which maintains the shutdown margin in the event that water from the RWST is inadvertently pumped into the reactor.
This effectively precludes a boron dilution from the RWST and is acceptable.
Specifications
- 3. 1.2.3 and 3. 1.2.5 A similar footnote (see 5 above) has been added to these specifications and is similarly acceptable.
Speci ficati ons 3.1.2. 7 The required volumes of borated water in the Refueling Water Storage Tank and Boric Acid Storage Tank have been altered to assure that the core may be maintained in a subcritical condition during Modes 5 and 6.
The required volumes are consistent with the analyses and are acceptable.
Specification 3/4.2. 1 Actions relating to operation of the APDMS have been removed and replaced by actions relating to APL.
Siqce the APOMS will not be used this is acceptable.
Surveillance to F
(z) is replaced with APL surveillance.
This is consistent with the re lacement of the APOFiS with APL and is acceptable.
Specification 3.2.2 The F (z) limit has been changed to 2.10 for Exxon fuel to reflect the results of the latest LOCA analysis.
This is acceptable.
Action a. 1 has been altered to remove the requirement that any necessary modification of the overpower bT trip be performed in Hot Standby.
The licensee has performed an evaluation which concludes that the modification may be done in Mode 1.
We find this to be acceptable.
Action Statement 3.2.2.a.2 which related to the APDMS has been removed.
Since the APDMS will not be used, we find this to be acceptable.
The F (z) surveillance requirements of 4.2.2.2 has been removed to APL Speci)ication 3.2.6.
This is consistent with the Exxon PDC-II requirements and is acceptable.
Figure 3. 2'. a has been renormalized to account for the revised F~(z) limit.
This is acceptable.
Specification 3.2.3 - Nuclear Enthalpy Hot Channel Factor (F>H)
The flow rate limit has been removed from the Specification to Specification 3.2.5.
The trade-off between flow rate and F
has been eliminated.
This change has been previously approved for UNt 1 and is acceptable tor Unit 2.
Specification 3.2.5.1 DNB Parameters
- Mode 1 The limit on Reactor Coolant System Total Flow Rate has been added to this specification and the former Specification
- 3. 2.5 has been split into two Specifications 3.2.5. 1 and 3.2.5.2.
Surveillance requirements have been added for the flow rate limit.
This is acceptable.
The value of the limit on Reactor Coolant System T v has been changed to reflect the assumption of 10K system generatoPlube plugging and the installation of the RdF RTDs.
The new value is consistent with the analyses and is acceptable.
Specification 3'.5.2 DNB Parameters
- Modes 2, 3, 4 and 5
This specification has been added to establish initial conditions for the rod bank withdrawal analysis for these modes.
The Specification is consistent with the assumption used in the rod withdrawal analysis and is acceptable.
Specification 3.2.6 This Specification is added to satisfy the requirements of the PDC-II power distribution control techniques.
This technique has been generically approved (letter, R. Tedesco, NRC to G. Owsley, Exxon, dated March 18, 1981) and its use for D.C.
Cook is acceptable.
The specification as written is consistent with the generically approved technique and is acceptable.
The F (z) value has 'been changed to 2. 10 for Exxon fuel to reflect the revis3d LOCA analysis.
This is acceptable.
Table 3.3-1 The Power Range Neutron Flux trip unit has an additional operability requirement imposed.
In order to satisfy the assumption used in the zero power rod withdrawal event safety analysis this trip is required to be operable whenever the rod control system is capable of withdrawing rods.
The proposed specification changes accomplishes this and is acceptable.
-18" 15.
Table 3.3-2 A conservative value of the response time for the Pressurizer Pressure High trip was added to this Table to satisfy the requirements of the safety analysis and to provide consistency with Unit 1 speqifications.
This is acceptable.
Table 4. 3-1 Requirement for surveillance of the additional operability of the Power
- Range, Neutron Flux-High trip was added (see Table 3.8. 1 above)..
Exemptions from the requirements of Specification 4.0.4 were added for those functional units for which surveillance must be performed in the applicable mode.
This is acceptable.
17.
Table 3.3-3 An additional requirement for operability has been added to the Steam Generator Water Level - High-High trip to require it to be operable when the generator is being supplied with the main feedpumps in Modes 3 and 4.
This requirement is consistent with the safety analysis and is acceptable.
Conditions for blocking safeguards action in Mode 3 are added.
These are required to satisfy conditions assumed in the safety analyses and are acceptable.
18.
Table 3.3-4 Allowable value for low-low T and steam generator low-low water level trips are changed to comply wf6 the new analysis of trip allowances.
This is acceptable.
19.
Table 4.3 ~ 2 20.
A requirement for surveillance of the additional operability of the Steam Generator Water Level-High-High trip is imposed (see Table 3.3-3 above).
Specification 3.4. 1.2 and 3.4.1.3 Additional requirements on the operability of reactor coolant loops in Mode 3, 4 and 5 are imposed in order to satisfy the assumptions made in the analysis of the zero power rod bank withdrawal event.
This is acceptable.
A footnote is added which exempts consideration of transfer of water from the RWST to the core from consideration as a source of dilution water.
This is acceptable (see discussion under Specification 3.1.2.7 and
- 3. 1. 2. 8).
-19" 21.
Specification 3/4.4.1.4 This Specification has been deleted.
Its requirements have been transferred to other Specification or deleted as not required (e.g.,
three-loop operation actions).
This is acceptable.
22.
Specification 3.4.2 An additional action statement was proposed which satisfies the assumptions of the safety analyses.
This is acceptable.
23.
Specification 3.4.11 The action statement of this specification has been removed to assure that no more than one PORV is inoperable.
If more than one PORV is inoperable orderly shutdown is required.
This action is taken to assure that PORVs are available to mitigate against a steam generator tube rupture without offsite power.
We find the revised specification to be acceptable.
24.
Specification 4.7.1.2 The surveillance requirements on auxiliary feedwater pumps are changed to provide assurance that these pumps can deliver sufficient flow to fulfill the requirements of the safety analyses.
This is acceptable.
25.
Specifi.cation 3/4.7.1.5 The action statements are altered to prohibit continued operation in Mode 1 with an inoperable steam generator stop valve.
Entry into Mode 2 or 3 is mandated.
This is more conservative than the present specification and is acceptable.
The surveillance requirements are clarified with respect to exemptions from Specification 4.04.
This is acceptable.
Environmental Consideration This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR'Sec 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
Conclusion The Commission made a proposed determination that the amendment involves no significant hazards consideration which was published in the Federal Re ister (51 FR 12249) on April 9, 1986, and consulted with the state Uo 'FicCcigan.
o public comments were received, and the state of Michigan did not have any comments.
We have 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Princi al Contributors:
D. Wigginton W. Jensen J.
Watt W. Brooks F. Burrows Dated:
May 21, 1986
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