ML20245K382
| ML20245K382 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/28/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245K380 | List: |
| References | |
| NUDOCS 8907050111 | |
| Download: ML20245K382 (9) | |
Text
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. UNITED STATES 2
NUCLEAR REGULATORY COMMISSION o
l r.
j WASHINGTON, D. C. 20555
/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELAltD TO AmtnuntMI NO. 37 I
i TO FACILITY OPERATING LICENSE NO. NPF-49 NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.
MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 i
DOCKET NO. 50-423 l
1.0 INTRODUCTION
t I
By application for license amendment, Northeast Nuclear Energy Company, et al.
(the licensee), requested changes to Millstone Unit 3 Technical Specifications.
(TS). The proposed amendment would change the Millstone Unit 3 TS to allow-Cycle 3 operation as follows:
(1) TS 3/4.2.2 " Heat Flux Hot Channel Factor.-
Four Loops Operating and Three Loops Operating" would be changed to eliminate the reference to fuel assembly grid locations,-(2) TS 5.3.1., " Fuel Assemblies"-
would be changed to properly describe the Cycle 3 fuel assemblies, (3) TS 5.3.2,
" Control Rod Assemblies" would be changed to allow use of silver indium:-
cadmium control rods and (4) TS 6.9.1.6, " Radial Peaking Factor Limit Report" would be changed to allow submittal of the report prior to each' cycle's initial criticality.
Inresponse3arequestfromtheNRCstaff,thelicenseesubmittedadditional 2
information concerning Cycle 3 design and operation.
2.0 DESCRIPTION
OF MILLSTONE 3, CYCLE 3 The Millstone Unit 3 reactor core is comprised of 193 fuel assemblies. The Cycle 3 core loading configuration features a low leakage pattern. During.
j Cycle 2/3 refueling, 32 fresh Region SA assemblies, 44 fresh Region 5B assemblies and 9 Region 2 assemblies from the spent fuel pool will replace 45. Region 2 fuel assemblies and 40 Region 3 fuel assemblies. A sussiary of the Cycle 3 fuel inventory is show below:
Region 2
3 4A 4B 5A 5B Enrichment (w/oU-235)*
2.899 3.395 3.497 3.808 4.10**
4.50**
g Geometric Density
- 94.965 94.980 95.13 95.17 95 95
(% theoretical) g No L g Number of Assemblies 9
24 56 28 32 44
{
ao 78 Approximate Burnup at 21 24,160 19,470 15,450 0
0 BeginningoCycle3(MWD /NTUl350
- 00
'~U
]
-o
- All values are as-built except Region SA and 5B i
- Enrichment of enriched axial region of. assemblies. Each Region 5 o
.Em assembly also has six inches of 0.74.w/o axial blanket fuel at top and
- ~'
bottom
- Based on actual EOC1 burnup of 18,700 MWO/MTU and nominal EOC2 burnup of 15,800 MWD /MTU.
1
. 2.1 Mechanical Design I
The mechanical design of the Regions SA and SB fuel assemblies is the same as the Region 4 fuel assemblies except that the Region 5 assemblies will incorporate several upgraded fuel design features.
These features include:
(1) Extended Burnup Capability, (2) Reconstitutable Top Nozzles (RTNs), (3)
Debris Filter Bottom Nozzles (DFBNs), (4) Integral Fuel Burnable Absorbers q
(IFBA), (5) Axial Blankets, and (6) Snag-Resistant Grids.
These design improvements are described below.
l l
(1) Extended Burnup Capability The Region 5 fuel assembly design was modified for extended burnups by reducing the thickness of both the top nozzle and bottom nozzle end plates, decreasing the height of the top nozzle and bottom nozzle, and increasing the fuel rod length with a corresponding increase in the length of the fuel rod plenum.
The fuel assembly.overall height was adjusted to be consistent with fuel assembly growth predictions based j
upon accumulated Westinghouse in-core experience.
This experience includes the results of high burnup demonstration programs conducted jointly by Westinghouse and utilities.
These design changes allow for an additional distance between the nozzle plates, which is allocated for two purposes:
(1) increased fuel rod growth associated with extended burnup.
and (2) increased fuel rod length to add plenum space for the increased fission gas release that occurs with increased burnup.
As part of this design change, the grid elevations were relocated slightly to standardize the 17x17 fuel assembly design.
Analyses have indicated the
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acceptability of the mechanical integrity of all fuel assembly components l
for extended burnup levels with the above changes.
Themethogsand criteria established for Westinghouse fuel at extended burnup have been I
approved by the NRC.
(2) Reconstitutable Top Nozzle (RTN)
The RTN differs from the current design in two ways:
a groove is j
provided in each thimble thru-hole in the nozzle plate to' facilitate attachment and removal; and the nozzle plate thickness was reduced to I
provide additional space for fuel rod growth.
In conjunction with the l
RTN, a long tapered fuel rod bottom end plug is used to facilitate l
removal and reinsertion of the fuel rods.
Details of the RTN design features, the design basis, and the evaluation of the RTN are given in l
l Section 2.3.2 of Reference 4 which has been approved by the NRC.
4 (3) Debris Filter Bottom Nozzle (DFBN) l Thb bottom nozzle is designed to inhibit debris from entering the active fuel region of the core and thereby improves fuel performance by minimizing debris related fuel failures.
The DFBN utilizes the same i
material, geometry, and welding requirements as its existing bottom nozzle counterpart.
The DFBH is a low profile bottom nozzle design made of stainless steel, with reduced plate thickness and leg height thus providing additional space for fuel rod growth as part of the extended burnup feature.
The DFBN is hydraulically equivalent to the existing bottom nozzle and meets all mechanical design functional requirements.
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)
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.j
.; 4 (4): Integral Fuel Burnable Absorber (IFBA)-
The IFBA coated fuel pellets are' identical to the enriched uranium dioxide pellets except for. the addition of a thin boride coating on the pellet-I cylindrical surface along the central portion of 'the fuel stack length.
IFBAs' provide power peaking and moderator. temperature coefficient control.
Details of the IFBA design are given in Section 2.5 of Reference 4.
(5) Axial Blankets The axial blanket consists:of natural uranium (approximatelyl0'.74 w/o)-
dioxide pellets at each end of the fuel stack-to reduce neutron leakage and to improve uranium utilization. The. axial blanket pellet design ~is.
the same as the. enriched and IFBA pellet designs except for an increase in length. The. length difference'in the axial blanket pellets will help-prevent accidental mixing with the enriched and IFBA pellets.,l Axial blankets are further discussed in Section 2.4 and.3.3 of, Reference 4.
-(6) Snag-ResistantGrg The snag-resistant grids contain outer grid straps that are modified to~
help prevent assembly hangup due to grid strap interference during fuel assembly removal. This was accomplished by changing the grid-strap.
l corner geometry and adding guide tabs on the' outer grid' strap.. Inter-
~'
mediate vanes to the top and tabs to,the bottom-of grids to reduce the potential-of an assembly overlapping (and possibly locking) onto an adjacent fuel assembly. The corner chamfer is formed in the outside-strap punching operation to eliminate grinding.and the resultant sharp edge.
In addition, a weld is placed on the small overlap on the top and bottom of the corners to increase strength and round over the leading.
edge of the corner.
1 In Cycle 3, some of the Hafnium PCCAs may be replaced with Ag-In-Cd Enhanced
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Perfomance Rod Cluster Control Assemblies '(EP-RCCAs). The absorber diameter l
of the EP.RCCA is reduced slightly at the lower extremity of the rodlets in order to acconinodate absorber swelling and minimize cladding interaction.
r-However, the EP-RCCA design for Millstone Unit 3 does not include the wear resistance feature that is typically standard with the EP-RCCA design.- The i
licensee is still evaluating the need for the wear resistance feature. With regard to the remaining hafniunt control rods, the licensee is planning to conduct examinations of the gafnium control rod assemblies using the...
i Westinghouse reconsnendations. during the next refueling outage which is l
scheduled to begin in May 20, 1989. Based on the results of the examinations, NNEC0 may elect to replace certain control rod assemblies with the silver-indium-cadmium rods or will provide a justification for continued operation-with the existing hafnium control rod assemblies.
1 TheRegionsSAand58{uelhasbeendesignedutilizingthe' late 9 westinghouse fuel performance model, the Westinghouse clad flattening model, and the Westinghouse extended burnup methodology. The Westinghouse fuel is designed L
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1 t
- 4..
andoperatedsothatcladflatteningwillnotoccurforitsplanngd-residence time in the reactor. The fuel rod internal pressure design basis is satisfied for all regions.
Westinghouse'sexperiencewithZirca;cycladfugisdescribedinWCAP-8183, i
" Operational Experience with Westinghouse Cores This report is updated I
annually.
The principal design features of the Regions SA and 58 fuel assemblies have
.been generically addressed by Westinghouse and: accepted by the NRC staff.
Based upon the above, we conclude that the use'of the Regions 5A and 58 fuel assemblies is acceptable for Cycle 3 operation.
2.2 Nuclear Design The nuclear design of the Cycle 3 core used Westinghouse codes approved by the NRCandthestandardcalculationalmghedsdescribedintheWestinghouse Reload Safety Evaluation Methodology This methodology is not affected by changes to the maximum uranium enrichment used in the fuel. The changes in physics cieracteristics for Cycle 3 are typical of the normal variations seen from cycle to cycle.
The Cycle 3. core loading is designed to meet a F K(Z)* for four loop operation and 2 A0 x K(Z)*0for three loop oper{ation.
x P ECCS limit of 2.32 x The flux difference (AI) bandwidth ~(duHng normal operation conditions +3,
-12%
for four loop operation and +5,-5% for three loop operation.
The Cycle 3 kinetics characteristics values fall within the current limits 1
with the exception cf the least negative Doppler temperature coefficient._
)
There is no significant impact of this change in the least negative Doppler temperature coefficient, on the accident analysis (see Section 3.0, herein).
The licensee has addressed the control rod worths and requirements at.the most limiting condition during the cycle _for a core of 61 Hafnium rod cluster.
controlassemblies(RCCAs). Therequigdshutdownmarginisbasedon previously submitted accident analysis The available shutdown margin 1
exceeds the minimum required.
For Cycle 3 some Hafnium RCCAs may be replaced by Ag-In-Cd EP-RCCAs. This change has been evaluated to allow the exchange of the Ag-In-Cd EP-RCCAs for any number of Hf RCCAs provided that any control or shutdown bank consists entirely of only one type of absorber material. 'This is possible since both RCCA designs have similar neutronic characteristics. The largest change'in 4
total rod worth during the cycle is less than 100 pcm**.
Core peaking factors change by less than 15.. As a result, the core performance characteristics of the Ag-In-Cd EP-RCCAs remain essentially the same. The available shutdown margin will exceed the minimum required shutdown margin.and all other Technical..
- pcm=10~gFigures2and3ofReference3.
"MZ) - se
_ _- --_ ______ L_a
1
-1 0
. )
Specification limits related to nuclear design will be met for any combination of Hf and Ag-In-Cd RCCAs in the configuration (s) described above.
- 2. 3 Thermal and Hydraulic Design No significant variations-in thermal margins will result from the Cycle 3 i
reload.gjugicientDNBmarginexistsfuralleventstomeetthedesign criteria for the Cycle 3 reload core.
The DNB core limits and safety analysis used for Cycle 3 are based on i
conditions given in Sections 1.0 and 3 g.
Fuel temperatures were calculated
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1 using the revised thermal safety model and include the effects of standardized pellets.
3.0 ACCIDENT ANALYSIS l
l The licensee has evaluated the impact of the Cycle 3 reactor core design on the Millstone Unit 3, FSAR, Chapter 15 events.
The following conclusions relate to the accident evaluation:
I For both large and small break LOCAs the Cycle 3 core configuration assures that the analysis presented in the FSAR remains bounding.
Thus, operation'during Cycle 3 meets the requirements of 10 CFR 50.46 and Appendix K to 10 CFR Part 50.
For non-LOCA events, with the exception of three events, the Chapter 15 FSAR analyses were determined to be bounding.
The effects of Ag-In-Cd EP-RCCAs and the least negative Doppler Temperature Coefficient (inadvertent actuation of ECCS) were evaluated with regard to the Chapter 15 FSAR events and were found to be acceptable.
The three events that were reanalyzed by the licensee were the RCCA ejection, the steam system piping failure, and the reactor coolant pump shaft seizure (locked rotor).
3.1 RCCA Ejection Accidents The RCCA ejection accident initiated from hot zero power (HZP) conditions at end-of-life (EOL) was reanalyzed for Cycle 3.
The HZP EOL RCCA ejection case was the only case reanalyzed because the. existing assumptions and results for the other cases continue to bound Cycle 3 operation.
It should be noted the the HZP RCCA ejection event is only analyzed for N-loop operation since this analysis bounds the N-1 loop cases.
The results of the reanalysis demonstrated that the conclusions of the FSAR for the RCCA ejection event remain valid.
3.2 Steam System Piping Failure The main steam line rupture event for N-loop operation was reanalyzed for Cycle 3 using revised core kinetics parameters.. Both the limiting case that assumes the availability of offsite power throughout the event and the less
severe case that includes a loss of offsite pswer were reanalyzed. No reanalysis was required for the N-1 loop main steamline rupture, since the existing analysis for that case continues to bound Cycle 3.
A DNB analysis was perfomed for the limiting case and it was determined that l
the conclusions of the FSAR for the main steam line rupture event remain
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valid. The DNB design basis, thus continues to be met for this event.
3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor)
Reanalysis of the locked rotor event was performed for Cycle 3 to predict the number of fuel rods that undergo DNB for N loop and M-1 loop operation. The analysis for N-1 loop operation assumed an initial nominal power level of 65%
Rated Thermal Power (RTP). Previous analysis for this event, along with all other non-LOCA events, had assumed an initial nominal N-1 loop power level of 75% RTP. The actual licensed N-1 loop nominal power level is 65% RTP so that the use of 75% RTP represented a conservatism in the previous analysis.
For Cycle 3, continued use of 75% RTP would have resulted in the predicted number of fuel rods that undergo DNB for this event exceeding the current limit
- value, j
The result of the locked rotor reanalysis verified that less than 8.0% of the fuel rods were predicted to undergo DNB for the N-1 loop locked rotor event with the 65% RTP assumption. The reanalysis of the N loop locked rotor event verified that less than 6% of the fuel rods were predicted to undergo DNB.
The radiological dose release evaluation for Cycle 3 was performed by the licensee. Doses were determined by adjusting the FSAR analysis doses to account for a higher number of fuel rods exceeding DN8R and to account for a shorter.
period of time to isolate the effected steam generator (from 30 minutes to 20 minutes). We find this approach to be acceptable.
The Cycle 3 locked rotor reanalysis described above was limited to the issue of detemining the number of rods in DNB. The limiting cases of the current locked rotor licensing basis analysis intended to predict other transient conditions such as maximum RCS pressure, maximum clad temperature, and the magnitude of the zirconium steam reaction remain valid for Cycle 3.
It should be noted that for N-1 Loop, use of the current locked rotor analysis means continuing to use an initial N-1 loop nominal power of 75% RTP. The reduced N-1 loop nominal power level of 65% RTP was only used for the rods-in DNB analysis.
.l The results of the locked rotor analysis are acceptable and do not exceed the consequences of previous analyses.
j 4.0 DESIGN BASIS ACCIDENT ANALYSIS RELATIVE TO EXTENDED FUEL BURNUP.
We have evaluated the potential impact of the. radiological assessment of the design basis accidents (DBA), which were previously analyzed in the licensing of Millstone Unit 3.
An NRC publication entitled, " Assessment of the Use of Extended Burnup Fuel in Light Water Reactors," NUREG/CR 5009, February 1988, examined the changes that could result in the DBA assumptions, described in the various appropriate SRP sections and/or Regulatory Guides, that could result from the use of extended burnup fuel (up to 60,000 MWD /MT). The only DBA consequence that could be-affected by the use of extended burnup fuel, even in a minor way, would be the potential thyroid doses that could result from a fuel handling accident.
j L
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NUREG-CR/5009 estimates that I-131 fuel gap activity in the peak fuel rod with
)
60,000 MWD /MT burnup could be as high as 12%. This value is approximately 20%
higher than the 10% I-132 fuel gap activities value nomally used in evaluating fuel handling accidents-(Regulatory Guide.l.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facilities for Boiling'and Pressurized Water Reactors").
We, therefore, reevaluated the fuel handling accident for Millstone Unit 3'with
}
an increase in iodine gap activity in the fuel damage in a fuel handling 1
accident. Listed below are the fuel handling ~ accident thyroid doses presented in the operating licensing Safety Evaluation Report (NUREG-1031)15 and the increased thyroid doses (by 20%) resulting from extended burnup fuel.
Exclusion Area low Population Zone
}
Thyroid Dose (rem)
Thyroid Dose (rem)
]
NUREG-1031 20% Increase NUREG-1031 20% Increase l
1.8
'2.2 0.1 0.1 The increased doses are still within our acceptance criterion stated in Standard Review Plan Section 15.7.4, i.e., well within the 300-rem thyroid exposure i
guidelines values set forth in 10 CFR Part 100.
5.0 TECHNICAL SPECIFICATIONS l
The proposed amendmeot would change the Millstone Unit 3 Technical
{
Specifications (TS) to allow Cycle 3 operation as follows:
(1) TS 3/4.2.2 i
" Heat Flux Hot Charnel Factor - Four Loops Operating and Three Loops Operating" would be change to eliminate the reference to fuel assembly grid locations, (2) TS 5.3.1., " Fuel. Assemblies" would be changed to properly described the Cycle 3 fuel assemblies, (3) TS 5.3.2, " Control Rod Assemblies" would be change to allow use of silver - indium - cadmium control rods and (4) TS 6.9.1.6, " Radial Peaking Factor Limit Report" would be changed to allow submittal of the report prior to each cycle's initial criticality.
5.1 Fuel Assembly Grids With regard to fuel assembly grids, TS 3/4.2.2 allows the licensee to exempt the measured values of the planar radial peaking factor (F ) when measured at the specified elevations where the grids are located. Th1Fisduetothe effects of the grids which make the flux measurement inaccurate. Since the new fuel assemblies will have slightly off-set grid locations, the existing TS is not applicable. The licensee has proposed that the specific grid elevations in TS 4.2.2.1.2f.3 and a.2.2.2.2f.3 be eliminated and replaced with a more general requirement that up to 20% of the core height, between 5% and 85% core height, can be eliminated from F measurement.
xy As indicated in the licensee's application,I "The proposed change to delete i
specific grid plane centerline will not affect the total percent of the core monitored for F The previous requirement to monitor the core between 15%
and 85% of core he.ight except within *2% of grid centerlines is essentially x
equal to the new requirement. There are 5 grids in the region to be monitored i
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. which amounts to 20% being excluded due to the effect of the grids and another l
30% excluded at the top and bottom of the core.
The total excluded area with the proposed requirement is also 20% due to grids and 30% at the top and bottom of the core". We concur with this assessment and conclude that the proposed changes to TS 4.2.2.1.2f.3 and 4.2.2.2.2f.3 are acceptable.
l 5.2 Fuel and Control Rod Design The licensee has proposed changes to Section 5 of the TS. Changes to TS 5.3.1 and 5.3.2 would describe the Cycle 3 fuel assemblies and the use of silver-l indium-cadmium control rods, respectively. While the proposed changes to l
TS 5.3.1 and 5.3.2 would permit the subject modifications to be used, any impact on reactor operation due to revised safety analysis results would be reflected in changes to the Limiting Conditions for Operation and/or the Surveillance Requirements; no such changes have been proposed for Cycle 3 except as described in Section 5.1.
It should be noted that for proposed TS 5.3.1, although the TS and the safety analysis assume a maximum of 5.0 w/o fuel enrichment, the actual maximum Cycle 3 enrichment is 4.5 w/o enrichment.
l The changes described in proposed TS 5.3.1 and 5.3.2 represent refinements, rather than substantial changes, whose efficacy has been previously demonstrated in other operating facilities. We conclude that the proposed changes to TS 5.3.1 and 5.3.2 are acceptable.
5.3 Radial Peaking Factor Limit Report The licensee has proposed a change to TS 6.9.1.6 to allow submittal of the Radial Peaking Factor Limit Report prior to initial-cycle criticality rather than 60 days prior to the time that the applicable limits become effective.
The proposed change would allow the completion of cycle-specific calculations during the refueling outage when changes in core configuration, due to discovery of fuel leakage, may result. We conclude that the proposed change to TS 6.9.1.6 is acceptable.
6.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact have been prepared and published in the Federal Register on June 27, 1989 (54 FR 27082). Accordingly, based upon the environ-mental assessment, we have determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.
7.0 CONCLUSION
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 (3) the issuance of the amendment will not be inimical to the coninon defense and security or to the health and safety of the public.
1
8.0 REFERENCES
(1)
E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Proposed -
Revision to Technical' Specifications," dated January 24, 1989.
(2)
E. J. Mroczka letter to U.S. Nuclear Regulatory Connission, " Proposed Revision to Technical Specifications," dated February 23, 1989.
(3)
E. J. Mroczka letter to U.S. Nuclear: Regulatory Commission, " Cycle 3 i
l Proposed Revision.to Technical Specifications," dated April.12, 1989.
I (4) Davidson, S.L., et al. " Reference Core Report VANTAGE 5 Fuel Assembly,"
3 WCAP-10444-P-A, September 1985.
(5) Johnson,W.J.,"HafniumRCCAExaminationGuidelines,"NS-NRC-89-3417, March 7, 1989 (6) Weiner, R. A. et al., " Improved Fuel Performance Models for Westinghouse.
Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August 1988.
I (7) George, R. A., et al., " Revised Clad Flattening Model," WCAP-8381, July 1974.
(8) Davidson, S. L. and Kramer, W. R., " Extended Burnup Evaluation of l
Westinghouse Fuel," WCAP-10125-R-A, December 1985.
1 (9) Risher, D. H., et al., " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964-A, August 1978.
(10). Foley, J., and Skaritka, J., " Operational Experience with Westinghouse Cores,"(throughDecember31,1987).WCAP-8183, Revision 16, August'1988.
1 1
(11)Davidson,S.L.,etal.,"WestinghouseReloadSafetyEvaluation Methodology," WCAP-9273-A, July 1985.
(12) " Final Safety Acalysis Report Millstone Generating Station, Unit 3," USNRC.
Docket No. 50-423, December 1988.
(13) Letter from A. C. Thadani (NRC) to W. J. Johnson (Westinghouse),
January 31, 1989,
Subject:
Acceptance for Referencing of Licensing.
1 Topical Report, WCAP-9226-P/9227-NP, " Reactor Core Response to Excessive-Secondary Steam Releases."
(14) Leech, W. J., et al., " Revised PAD Code Thermal Safety Model," WCAP-8720, Addenda 2, October 1982.
l (15) " Safety Evaluation Report Related ~to the Operation of Millstone Nuclear Power Station, Unit No. 3," Docket No.- 50-423, July 1984.
ll Dated:
June 28, 1989 l
Principal Contributor:
1 D. Jaffe l
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