ML17333A815
| ML17333A815 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 03/13/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17333A814 | List: |
| References | |
| NUDOCS 9703190235 | |
| Download: ML17333A815 (23) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 2055&0001, S
Y EVALUAT ON BY THE OFF CE OF UCL R
EACTOR R GULAT ON ELATED TO AHENDHENT NO.
214 TO FACILITY OPERATING LICENSE NO.
DPR-58 D AH NDHEN O. 199 0
CI Y OPERATI G
ICENSE NO. DPR-74 INDIANA MICHIGAN POWER COHPANY DONA D C.
COOK NUC AR ANT UN NOS.
1 AND 2 OCKET NOS. 50-315 AND 50-316
- 1. 0 INTRODUCTION By letter dated Hay 26,
- 1995, and supplemented September 26,
- 1995, August 2,
- 1996, and February 6, 1997, the Indiana Hichigan Power Company (the licensee) requested amendments to the Technical Specifications (TS) appended to Facility Operating License Nos.
DPR-58 and DPR-74 for the Donald C.
Cook Nuclear Plant, Unit Nos.
1 and 2.
The proposed changes are primarily to support operation of Unit 1 at steam generator tube plugging (SGTP) levels up to 30X.
In addition, the licensee has performed analyses and evaluations to support, increased operating margins for Unit 1.
Some of the margin changes have been proposed for both Unit 1
and Unit 2.
Finally, one miscellaneous
,change has been proposed to make one Unit 1 specification more nearly like the corresponding Unit 2 specification, and one administrative change to maintain consistency of Unit 2 acceptance criteria.
The September 26,
- 1995, August 2,
- 1996, and February 6,
- 1997, supplements provided clarifying information that did not expand the scope of the initial application and did not change the staff's proposed no significant hazards determination.
2.0 EV LUATION 2.1 Pro osed Chan es to the Safet Anal sis The primary purpose of the licensee's submittal
.is to request approval to operate Cook Nuclear Plant Unit 1 with SGTP levels as high as 30X in each steam generator.
Since the analysis needed to support this request involved reanalysis or evaluation of most of the events discussed in Chapter 14 of the Updated Final Safety Analysis Report (UFSAR), the licensee performed the analyses such that additional operating margin was achieved in several areas.
In addition, a number of proposed changes that can be supported for both
- units, a miscellaneous
- change, and an administrative change have also'been proposed.
In addition to addressing an increased SGTP level of 30X, the following increased operating margins were also addressed:
9703i90235 9703i3 PDR ADOCK 050003i5 P
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(1)
Reduction of Safety Injection (SI) and Residual Heat Removal (RHR) discharge pressure on recirculation The RHR and SI minimum safeguards pump head curves were reduced by 15X, an additional 5X reduction from the current analysis degradation of 10X.
The charging pump head curve degradation is maintained at the current value of..lOX.
(2)
The emergency diesel generator (EDG) start time was increased from 10 seconds to 30 seconds.
(3)
To support increased delta-T drift, the margin between the safety analysis limits (SAL) and the nominal values of the Kl and K4 gains of the Unit 1 Over-Temperature Delta-T (OTDT). and Over-Power Delta-T (OPDT) setpoint equations were adjusted.
(4)
An increase in the pressurizer code safety valve (PSV) setpoint tolerance from +/-1X to +/-3X.
(5)
Decreased shutdown margin for'avg greater than 200'F.
The analyses and evaluations as discussed in Reference 1 support all of these
- changes, and provides the necessary documentation to support the TS changes associated with the SGTP program.
The results of the accident analyses and evaluations performed by the licensee for the SGTP program demonstrate that safe operation is maintained.
A summary of the conclusions of the accident analyses performed by the licensee that serve as the basis'or the staff's acceptance is provided below.
a e Break OCA LBLOCA The LBLOCA analysis was reanalyzed for the impact of the increased tube plugging level, reduced Thermal Design Flow (TDF), loop flow asymmetry, revised ECCS flows, and the increased EDG start time.
The LBLOCA analysis was not impacted by the pressurizer code safety valve tolerance
- increase, the revised Kl/K4 values, or the decreased shutdown margin.
The LBLOCA analysis was performed with the 1981 version of the Westinghouse ECCS Evaluation Model using the BASH computer code.
Analysis assumptions included ECCS flow with the RHR cross-tie valves closed; a total peaking factor of 2. 15, a hot channel enthalpy rise peaking factor of 1.55, and an accumulator temperature of 100 F.
A full spectrum break analysis was performed at the nominal RCS conditions (initial RCS pressure of 2250 psia and initial hot leg temperature of 609. 1'F) from which the limiting break discharge coefficient was determined.
The limiting break was then reanalyzed at the reduced hot leg temperature and nominal RCS pressure of 2250 psia, and also at nominal hot leg temperature and an initial RCS pressure of 2100 psia.
The above cases were all analyzed with minimum safety injection flow, which was determined to be limiting.
The limiting break was determined to be Cd = 0.4 at the nominal hot leg temperature (Thot 609. 1'F) and a pressure of 2100 psia with minimum safety injection flow.
The peak cladding temperature (PCT) was calculated to be 2164'F.
The LBLOCA analysis for 30X SGTP. level, as discussed
- above, was performed using a cosine axial power distribution based on the Power Shape Sensitivity Model (PSSM).
After the licensee completed their analysis, effective October 30, 1995, the PSSM was replaced by an alternate axial power shape methodology (ESHAPE) which is based on explicit analysis of a set of skewed axial power shapes.
Replacement of PSSM by ESHAPE was approved by NRC, and.
the use of ESHAPE methodology, in general, results in a more conservative PCT.
- Hence, the licensee performed an assessment (Reference
- 2) to determine the impact of this change on the magnitude of the PCT (2164 F) as submitted.
In their assessment, the licensee included a "compensatory benefit" which was developed by Westinghouse to reduce or eliminate the PCT penalty associated with the change from PSSM to ESHAPE.
This compensatory benefit results from the incorporation of the effects of leakage through the hot leg nozzle gap (HLNG) into the BASH methodology.
Westinghouse submitted the HLNG methodology for staff review on July 26, 1995 (Reference 3), but the review has not yet been initiated.
The Table below presents the results of the licensee's assessment:
BETII I L I PCT PSSM 2164 F
ESHAPE 2266 F
ESHAPE in conjunction with HLNG 2029 F
The above table shows that replacing PSSM by ESHAPE results in a PCT increase of 102 F.
If, however, the HLNG model is used in conjunction with ESHAPE, the PCT is 2029 'F.
Since the change in PCT is significant due to the change of evaluation model from PSSM to ESHAPE, and the HLNG is not currently an NRC approved model, the licensee is required by Section 50.46(a)(3)(ii) of Title 10 of the Code of Federal Regulations (10 CFR 50.46(a)(3)(ii)) to submit a
schedule for providing a reanalysis or take other action as may be needed to show compliance with 10 CFR 50.46 requirements.
Use of the HLNG model without prior,NRC review is permitted by 10 CFR 50.46;
- however, the staff reserves the right to review HLNG in the future.
In response to the staff's request, the licensee informed NRC by letter dated February 6,
- 1997, (Reference
- 4) that a revised submittal with an approved evaluation model will be made prior to the start up of cycle 18, which is currently scheduled to begin on April 20, 2000.
At present, Unit 1 is coasting down in cycle 15.
The proposed schedule for reanalysis appears reasonable.
However, the staff may review the HLNG methodology.
If the methodology is found to be unacceptable for inclusion in the LBLOCA, the licensee will be required to take immediate action to bring the plant into compliance with 10 CFR 50.46 requirements.
An accelerated schedule for reanalysis of the plant with an acceptable evaluation model may also be required.
~BB LII The small break LOCA analysis was reanalyzed for the impact of the increased tube plugging level, reduced TDF, loop flow asymmetry, revised ECCS flows, and the increased EDG start time.
The small break LOCA analysis was not impacted by the pressurizer code safety valve tolerance
- increase, the revised Kl/K4
- values, or the decreased shutdown margin.
The small break LOCA analysis was performed with the Westinghouse small break LOCA ECCS Evaluation Model using
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the NOTRUHP code (including the recent model changes submitted in
,WCAP-10054-P, Addendum 2, Revision I, which was approved by staff on August 12, 1996).
The key analysis input assumptions include ECCS flows with the High Head Safety Injection (HHSI) cross-tie discharge valves closed, a
total peaking factor of 2.32 and hot channel enthalpy rise peaking factor of 1.55.
Other analysis input assumptions incorporated in the small break LOCA analysis are reduced hot assembly average power (Pha) and a power shape based on a reduced axial offset of +20X.
A single break size analysis was performed at the previously-limiting break size of three inches.
The calculation used the reduced temperature, reduced pressure operating condition.
An evaluation of the break spectrum and the range of operating conditions concluded that the analyzed case would remain bounding with respect to peak clad temperature.
The calculation was performed with minimum safety injection flow, which was limiting.
The peak cladding temperature was calculated to be 1443'F.
OC H d aulic o cin unctions LOCA hydraulic forces are relatively insensitive to specific SGTP levels.
The D.C. Cook Nuclear Plant LOCA hydraulic forces were most recently analyzed for the Rerating Program.
The RCS parameters used in the existing analysis conservatively bound the conditions at 30X tube plugging.
Therefore, the existing LOCA forces analyses remain conservative relative to the SGTP Program.
Non-LOCA Anal ses The non-LOCA events were addressed by a combination of evaluations and analyses for the impact of the increased tube plugging level, reduced
- TDF, loop flow asymmetry, revised ECCS flows, pressurizer code safety valve tolerance
- increase, increased EDG start time, revised Kl/K4 values, and decreased sh'utdown margin.
The non-LOCA safety analyses were reviewed on the basis of both DNB and non-DNB acceptance criteria.
All DNB event reanalyses were found to yield a minimum DNBR which remains above the limit value.
The analyses demonstrate that all licensing basis criteria continue to be met and the conclusions presented in the UFSAR remain valid.
Steam Generator Tube Ru ture SGTR The SGTR event was analyzed for the impact of the increased tube plugging level and associated reduced TDF and loop flow asymmetry.
The SGTR analysis was not impacted by any of the SGTP Program increased operating margins.
The thyroid and whole body doses estimated for the Unit I, based on the 30X SGTP evaluation, remain within a "small fraction" (lOX) of the 10 CFR 100 exposure limit guidelines.
Therefore, the conclusions of the UFSAR remain valid.
ost-LOC ot Le Recirculation Time The hot leg switchover to preclude boron precipitation and post-LOCA long term cooling are not adversely affected by the 30X SGTP Program.
The proposed changes do not significantly affect the normal plant operating parameters, the safeguards systems actuation, the accident mitigation capabilities important to these
- events, o} the assumptions used in the analysis of these events.
The
proposed changes do not create conditions more limiting than those assumed in the LOCA-related analyses.
Radiolo ical Doses A reanalysis of the offsite doses following a large break LOCA was performed for the incr'ease in emergency diesel generator start time to 30 seconds.
,While there was a slight increase in the offsite thyroid doses, the doses are within the applicable limits.
The source terms for LOCA and the fuel handling accident are unaffected by the increase in SGTP level or any of the other SGTP Program increased operating margins.
Post Accident H dro en Production The licensee examined the effects of the Steam Generator Tube Plugging program on post-LOCA hydrogen generation and concluded that the values employed in the Rerating Program analysis remain bounding for the Steam Generator Tube Plugging (SGTP)
Program.
Containment Anal ses The design pressure for the D.
C.
Cook primary containment is 12.0 psig as given in TS Bases for LCO 3.6. 1.4 and 3.6. 1.5 and discussed in FSAR Section 5.2.2.2 "Design Load Criteria".
The previous analysis of peak containment pressure (licensee letter to NRC dated August 22, 1988) resulted in a value of 11.89 psig.
As part of the Steam Generator Tube Plugging (SGTP)
- Program, the licensee recalculated the peak containment pressure.
The revised value, based on calculations for the SGTP Program, is 11.49 psig.
TS Bases for LCO 3.6. 1.4 and 3.6. 1.5 were revised to reflect this value.
The analyses assumed a power level of 3413 (plus 2X) Hwt, which bounds the operation of both units.
The licensee reanalyzed containment integrity following a LOCA and a Hain Steam Line Break to consider the impact of the increased level of tube plugging, the reduced thermal design flow, loop asymmetry, revised ECCS flows and the increased EDG start time.
The licensee also evaluated the short term containment analysis done for the Rerating Program and concluded that it bounds the case of 30% tube plugging.
For the LOCA, the licensee used previously approved methods to calculate the mass and energy released by a LOCA into containment and the resulting containment pressure.
The LOCA mass and energy release calculations used the 1979 model which consists of the SATAN VI, WREFLOOD and FROTH computer codes.
The results were then input to the LOTIC-1 computer program to perform the containment integrity peak pressure calculations.
In general, the peak containment pressure resulting from a LOCA decreases with increased steam generator tube plugging since the reactor coolant volume and fluid released are reduced, heat transfer across the steam generator tubes is reduced because the area is reduced (which decreases the energy of the escaping fluid) and the pressure differential upstream of the break is increased, resulting in a decrease in the break flow rate.
However, for
1 I
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conservatism, the licensee assumed OX tube plugging for the mass.
and energy release calculations in order to maximize energy released to the containment.
The core rated thermal power assumed in these calculations was 3413 Hwt.
The core rate thermal power assumed in the Rerating Program was 3425 Mwt.
In a August 2,
- 1996, response to a staff question, the licensee stated that the calculated decrease, in containment pressure (from 11.89 psig to 11.49 psig) is due to several reductions in margin included in the calculations performed for the Rerating Program.
The revised calculations were performed with plant specific rather than generic values for the steam generator metal and the amount of fuel stored energy.
In addition, more accurate data for mass and energy release during the post reflood period contributed to the decrease in margin which resulted in a lower v'alue of peak containment pressure.
The calculations still retain sufficient coriservatism including the assumption of minimum safeguards and assumptions to maximize the melting of the ice bed since this depletes the ice in the shortest time.
The licensee also calculated the consequences of both a large and small Hain Steam Line Break.
The containment pressure calculated for a LOCA is more severe than that for the Main Steam Line Break.
However, the Main Steam Line Break results in higher containment temperatures.
The calculated peak containment temperature is 326'F which is within the Environmental Acceptance Criteria.
2.2 Pro osed Chan es to the TS The proposed changes in the TS are discussed below.
TS
. 2-F u
e 2. 1-1:
Reactor Core Safety Limits were revised.
This change was proposed to increase Unit 1 operating margin.
The revised OTDT and OPDT setpoints are based upon new core thermal safety limits, which account for the effects of the Reactor Coolant System (RCS) parameter changes associated with the increased level of SGTP, using the methodology described in Reference 5.
These setpoints were revised to increase the available margin between the safety analysis setpoint values and the nominal, or TS values, such that more delta-T drift could be accommodated between instrumentation calibrations during the fuel cycle.
Presently, the power margin associated with the Rerating Program is being utilized to offset the delta-T drift that is being experienced during core burnup (i.e., the core power of 3411 HWt is supported by the analyses, but the plant is actually operated with a core full-power value of 3250 HWt).
However, since the 30X SGTP parameters do not have this power margin available, there was a need to revise the OTDT and OPDT setpoints as part of the SGTP Program.
TS
. 2-5 Table 2.2-1 Footnote:
Design flow in footnote of Table 2.2-1 was redefined to 1/4 Minimum Measured Flow (HHF).
This change was directly related to increased SGTP level.
TS HMF is 1.025 times thermal design flow (TDF).
The HMF employed in the licensee's DNB analysis is 1.019 times TDF.
This was done to support a range of HMF's from 1.019 to 1.025 times TDF.
Design flow in current TS Table 2.2-1 is MMF/4.
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Table 2.2-1:
The upper limit on T'(Indicated Tavg at Rated Thermal Power) in OTDT equation was increased to reflect licensee's analyses.
This change was proposed to increase Unit 1 operating margin.
TS
. 2-8 Table 2.2-1:
The "numerical value of Kl in OTDT equation was reduced from 1.32 to 1. 17.
This change and the change 'to f(delta-I) was requested to optimize Unit 1 operating margin.
Some load rejection capability was sacrificed for instrumentation
- margin, increased allowance for core burndown effects on hot leg streaming, and an increase in the positive delta-I break point for the f(delta-I) penalty.
TS
. 2-8 ble 2.2-f(delta-I) was changed to increase the region of positive delta-I which is without penalty.
This change was requested to optimize Unit 1 operating margin.
The previous discussion on reduction of Kl is applicable for this change also.
S
. 2-9 able 2.2-1:
The upper limit on T" (Indicated Tavg at Rated Thermal Power) in OPDT equation was decreased to reflect the licensee's analyses.
This change was proposed to increase Unit 1 operating margin.
Unit 1 is operated in a low temperature, low pressure mode to extend the life of the steam generators.
Therefore, the licensee's analysis of the OPDT setpoint was analyzed with a low upper limit on T" to convert unused margin to operating margin.
TS
. 2-9 Table 2.2-1:
The allowable values in notes 3 and 4 were changed based on the licensee's calculation.
These changes were proposed to increase Unit 1 operating margin.
S 3
-1 Sectio s 3. 1.
.1
& 4. 1.
. 1:
The required shutdown margin was reduced.
These changes were proposed to increase the operating margin of both the units.
The new values were supported by the licensee's
- analyses, which included core response steam break (CRSB), steamline mass and energy release (SM&E) inside containment, and SNE outside containment.
S 3 4 1-11 Section
- 4. 1.2.3. 1:
The, Centrifugal Charging Pump (CCP) surveil,lance was changed to be consistent with 10X degradation.
The pump surveillance requirements were changed from "discharge" to "differential" pressure.
These changes were proposed to increase the operating margin of both the units.
The new surveillance criterion is supported by the licensee's analyses which included Loss of Coolant Accident (LOCA), CRSB, SH&E inside and outside containment.
The value given for CCP applies to both units.
TS 3 4 1-1
. Sect'on
- 4. 1.2.4:
The CCP surveillance was changed to be consistent with 10X degradation.
The pump surveillance requirements were changed from "discharge" to "differential" pressure.
These changes were proposed to increase the operating margin of both the units.
See previous discussion for TS pg. 3/4 1-11.
k S
3
-15 Section 3.1.2.7:
The minimum Refueling Water Storage Tank (RWST) temperature was reduced to 70'F.
This change was proposed to increase the operating margin of both the units.
The minimum RWST temperature for mode
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S 3 4
-16 Section 3. 1.2.8:
The minimum RWST temperature was reduced to 70'F.
This change was proposed to increase the operating margin of both the units.
The new value was supported by the core response large break LOCA performed by the licensee.
TS 3 4 2-14. Table 3.2-1:
The DNB temperature limit was increased from 570.9'F to 579.3'F.
In addition, the NHF limit was reduced from 361,600 to 341, 100 gpm which includes a 2.5X instrument uncertainty.
These changes are directly related to increased SGTP level.
Also, see the discussion for TS pg. 2-5.
TS 3 4 3-17 Table 3.3-3:
The Engineered Safety Feature (ESF) actuation logic to support 12X Auxiliary Feedwater (AFW) pump degradation was changed.
This change was proposed to increase Unit 1 operating margin.
The revised part of Table 3.3-3 incorporates the safeguards logic used in Unit 2.
This will allow for the use of 12X AFW head degradation in Unit l.
All analyses, other than an "information only" feedline break analyses, have been performed by the licensee using the flow from an AFW pump with 12X head degradation.
The safeguards logic itself will be modified via design change prior to implementation of these revised TS pages (i.e. before Unit 1, cycle 16).
After this modification, the Unit 2 feedline break analysis using 12X degraded flow will bound Unit l.
TS s.
3 4 3-1 23a 24 26 31 nd 33 Tables 3.3-3 3.3-4 and 4.3-2:
Change ESF actuation logic to support 12X AFW pump degradation.
These changes were proposed to increase Unit 1 operating margin.
See the discussion above for the change in TS pg. 3/4 3-17.
TS s.
3 4 4-4 and 5. Sections 3.4.2
& 3.4.3:
The pressurizer valve tolerance was increased.
These changes were proposed to increase Unit 1
operating margin.
The non-LOCA accidents were reanalyzed or reevaluated by the licensee based on a pressurizer valve setpoint'tolerance of 3X.
The licensee's analyses included loss of load, turbine trip, locked rotor/shaft break events, loss of normal feedwater, feedwater line break, and loss of all power to station auxiliaries.
TS 3 4 5-5 Section 4.5.2.f.2 and 3:
The RHR/SI pump surveillance requirements were changed to be consistent with 15X degradation, and the word "discharge" was replaced by "differential".
These changes were proposed to increase Unit 1 operating margin.
The new surveillance criteria are supported by analyses performed by the licensee.
S 3 4 5-5 Sections 4.5.2.f.2 and 3:
The RHR/SI pump surveillance requirement was changed from "discharge" to "differential" pressure.
This is an administrative change for Unit 2.
The discharge pressure criteria in the current TS correspond to the same pump performance characteristics as the proposed differential pressure criteria.
The change ensures that surveillance criteria use similar acceptance criteria.
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4 5-5. Section 4.5.2.f. 1:
The CCP surveillance requirements were changed to be consistent with 10% degradation, and the word "discharge" was replaced by "differential" pressure.
These changes were proposed to increase the operating margin of both the units.
Refer to the discussion given for TS pg. 3/4 1-11 for information concerning the 10%
CCP degradation.
/
TS 3 4 5-11. Section 3.5.5:
The minimum RMST temperature was reduced to 70 'F.
This change was proposed to increase the operating margin of both the units.
See discussion for TS pg. 3/4 1-16.
TS 3
4 7-6 Sections 4.7. 1.2.a
& b:
The AFW pump surveillance.was changed to be consistent with 12% degradation.
This change was proposed to increase Unit 1 operating margin.
See discussion for TS pg. 3/4 3-17.
The proposed surveillance criteria is identical to the criteri a in the Unit 2 TS.
These criteria correspond to the auxiliary feedwater flows used in all analyses for both units, except the "information only" Unit 1 feedwater line break.
As noted in the discussion for TS pg. 3/4 3-17, after the changes to the Unit 1
safeguards actuation logic, Unit 1 will be bounded by the Unit 2 feedwater line break.
TS
. 5-5 Section 5.4.2:
The system volume was reduced to account for plugged steam generator tubes.
This change is directly related to increased SGTP.
Since the actual level of.tube plugging may change each
- outage, a range of RCS volume corresponding to the range 0% to 30% plugging was specified:
approximately 12,466 to 11,551 ft~.
TS B 2-1 a
Bases Section
- 2. 1. 1:
The DNB values for fuel were changed, This change was proposed to increase Unit 1 operating margin.
The values for DNBR for typical and thimble cells were revised.
This change was related to the new thermal design and the new OTDT and OPDT protection trip setpoints.
See also discussions for TS pages 2-2, 2-7, 2-8, and 2-9.
TS B 2-4 Bases Section 2.2.1:
The detail from the discussion of the OTDT protection trip was removed.
This change was proposed to increase Unit 1
operating margin.
The discussion of the proper normalization of T'nd P'as removed.
This information will be controlled administratively.
TS B 2-5 Bases Section 2.2. 1:
The detail from the discussion of the OPDT protection trip was removed.
This change was proposed to increase Unit 1
operating margin.
The discussion of the proper normalization of T" was removed.
This information will be controlled administratively.
TS B 3 4 1-1 Bases Section 3 4. 1. 1. 1
& 2:
The required shut down margin was reduced.
This change was proposed to increase the operating margin of both the units.
See discussion for TS pg. 3/4 1-1.
TS B 3 4 4-1 Bases Section 3 4.4. 1:
The DNB values for fuel were changed.
The numerical value of "1.69" was replaced by the phrase, "the safety analysis limit".
These changes were proposed to increase Unit 1
operating margin.
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TS B 3 4 5-3 Bases Section 3 4.5.5:
The minimum RWST temperature was reduced to 70'F.
This change was proposed to increase the operating margin of both the units.
The conditions under which the reactor will remain subcritical were clarified.
Specifically, large break LOCA was called out as the initiating condition and the control rods were assumed to be out instead of being inserted, except for the most reactive assembly.
In addition, the explanation that a conservatively'high value of the RWST temperature was included in the TS for Unit 1 was removed because the proposed value of 70 F
was based on the licensee's analyses.
TS B 3 4 6-2.
Bases Sections 3 4.6. 1.4
& 5:
The peak containment pressure was changed to reflect licensee's analysis result.
This change was proposed to increase the operating margin of both the units.
3.0
SUMMARY
The staff has completed its review of the documentation submitted by the licensee in support of the proposed changes in the TS and the associated Bases for the Units 1
8 2.
On the basis of the evaluation presented
- above, the staff concludes that the proposed changes are acceptable.
The staff concurs that post-LOCA hydrogen generation has been adequately addressed and finds the new containment analyses to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendments.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIOERATION The amendments change the requirements with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements.
The staff has determined that the amendments involve 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 the amendments involve no significant hazards consideration and there has been no public comment on such finding (60 FR 37095).
Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 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 the amendments.
- 6. 0 CONCLUSION 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, (2) such activities will be conducted in compliance with the Commission's regulations,
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- 11 0
and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
1.
"Donald C.
Cook Nuclear Plant Unit 1 Steam Generator Tube Plugging Program Licensing Report,"
WCAP 14285, Revision 1,
May 1995.
2.
Letter from E.E. Fitzpatrick to NRC, "Clarification of Power Shape used in LOCA Analysis," September 26, 1995.
3.
Westinghouse Letter NTD-NRC-95-4477, Transmittal of Topical Reports WCAP-14404-P and WCAP-14405-NP, "Methodology for Incorporating Hot Leg Nozzle Gaps into BASH," N.J. Liparulo to USNRC Document Control Desk, July 26, 1995.
4.
Letter from E.E. Fitzpatrick to NRC, "Request for Additional Information,"
AEP:NRC: 1207C, February 6, 1997.
5.
Ellenberger S.L. et al.,"Design Bases for the Thermal Overpower delta-T and Thermal Overtemperature delta-T Trip Functions,"
- March, 1977.
Principal Contributors:
M. Razzaque, SRXB R. Lobel, SCSB J.
- Hickman, PD33 Date:
March 13, 1997
Li
<r
March 13, 1997 Hr.
E.
E. Fitzpatrick, Vice President Indiana Michigan Power Nuclear Generation Group 500 Circle Drive
- Buchanan, MI 49107
SUBJECT:
DONALD C.
COOK NUCLEAR PLANT, UNIT NOS.
1 AND 2 ISSUANCE OF AMENDMENTS RE:
INCREASED STEAM GENERATOR PLUGGING LIMIT (TAC NOS.
M92587 AND H92588)
Dear Hr. Fitzpatrick:
The Commission has issued the enclosed Amendment'o. 214'o Facility Operating License No.
DPR-58 and Amendment No. 199 to Facility Operating License No.
DPR-74 for the Donald C.
Cook Nuclear Plant, Unit Nos.
1 and 2., The amendments consist of changes to the Technical Specifications (TS) in response to your application dated May 26,,1995, and supplemented September. 26,'995, August 2,
- 1996, and February 6, 1997.-
The amendments revise the TS to allow operatiop of Cobk Uni.t '1 at steam generator tube plugging levels up to 30/C'.
Additional'hanges to increase operating margins for both Unit 1 and Unit' are also included.
o s
I A copy of our related Safety Evaluatio'n is also enclosed.,'Notice of Issuance will be included in the Commission's biweekly Federal Re<eister notice. '
o I
e Sincerely,,
Original Signed by John,B.
Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos.
50-315 and 50-316
Enclosures:
l.
Amendment No.2"4 to DPR-58 2.
Amendment No.199 to DPR-74 3.
Safety Evaluation cc w/encl:
See next page DISTRIBUTION:
See attached list DOCUMENT NAME:
G: II,DCCOOKiC092587. AHD To teceIre ~ copy of tNe docssnent. Indcete In the boxt 'C' Copy without ettechment/endostwe E
~ Copy with ettechment/endostme
'N' No copy OFFICE PH:PD33 E
LA:PD33 E
OGC g NAME JHickma DATE 03/j< /
EBarnhil 1 gJ 03/I /97 03/
/97 OFFICIAL RECORD COPY
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DATED:
March 13 1997 AMENDMENT NO. 2" 4 TO FACILITY OPERATING LICENSE NO. DPR-58-0.
C.
COOK-UNIT I M."NDMENT NO. "99 TO FACILITY OPERATING LICENSE NO. DPR-74-D.
C.
COOK-UNIT 2
-Docket F)le PUBLIC PDIII-3 Reading J.
Roe E. Ba'"nhill J.
Hickman (2)
OGC G. Hill, IfN (4)
C. Grimes, 0-llF23 M. Razzaque R. Lobel ACRS J. Caldwell, RIII G. Marcus
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