ML20042C156
| ML20042C156 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/05/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20042C154 | List: |
| References | |
| TAC-47355, TAC-47389, TAC-47473, NUDOCS 8203300278 | |
| Download: ML20042C156 (19) | |
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s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDriENT NO. 7 4 TO FACILITY OPERATING LICENSE NO.' DPR-65 NORTHEAST NUCLEAR ENERGY _COMPAtil, ET.AL.
MILLSTONE NUCLEAR P0'n'ER STATION, UNIT. NO. 2 00CKET N0. 50-336 o
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TABLE OF CONTENIS Subject Page 1
1.0 Introduction 2.0 Discussion and Evaluation 2
2.1 Fuel System Design 2
2.1.1 Seismic and LOCA Mechanical Reiponse 3
2.1.2 CEA and Fuel Assembly Guide Tebe Wear 3
2.1.3 Cladding Collapse 4
2.1.4 Fuel Manufacturing Problems 4
2.1.5 Miscellaneous 5
2.2 Nuc' lear Design 5
2.2.1 Control Rod Worth 6
2.2.2 Moderator and Doppler Temperature Coefficients 6
2.3 Thermal-Hydraulic Design 6
6 2.3.1 Hydraulic Compatibility 6
2.3.2 Design Power Level
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7 2.3.3 Analytical Methods 2.3.4 Reactor Coolant Flow 7
2.3.5 Limiting Transient-Complete loss of Reactor Coolant Flow 8
2.4 Accident Analyses 8
2.4.1 Boron Dilution Event 8
2.4.2 CEA Ejection 9
2.4.3 CEA Withdrawal from S' bcritical 10' u
2.4.4 CEA Withdrawal at Power 10 2.4.5 Complete Loss of Reactor Coolant Flow 10 2.4.6 Reactor Coolant Pump Seized Rotor 11 2.4.7 Steam Line Rupture Accident 11 2.5 Loss of Coolant Ac.ident 12 2.6 Radiological Consequences of Postulated Accidents 12 3.0 Technical Specification Changes 12 3.1 Shutdoven Margin 13 3.2 Moderator Temperature Coefficient 13 c
3.3 Withdrawn Position of Regu. ting CEAs 13 3.4 Pressurizer Level Control 13 4.0 Environmental Consideration 14 5.0 Conclusion 14
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6.0 References 15 O
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1.0 Introduction By applications dated December 2 and 17,1981 and January 14, 1982 (Ref.
811202, 811217 and 820114)* and supplemental information as listed in the reference sections, Northeast Nuclear Energy Company (NNECO or the licensee) requested an amendment ot Facility Operating License No. DPR-65 for the Millstone Nuclear Power Station, Unit No. 2 (Millstone-2or the facility).
The amendment request consists of:
P G Appendix A (Safety) Technical Specifications (TS) changes resulting from the analyses of the Cycle 5 reload fue.1; e Continued approval to operate with modified (sleeved, reduced flow and insert) Control Element Assembly (CEA) guide tubes; 9 Approval to operate with an additional 704 steam generator tubes pidgged;.and 0 Evaluation of numerous changes partially related-to Cycle 5 operation.
The specific request of the December 2,1981 application, to modify the oper-
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ability requirements for two independent shutdown cooling loops, was issued by Amendment No. 71 (Ref. 811218).
To simplify this reload Safety Evaluation (SE), numerous other changes partially related to Cycle 5 operation were issued by Amendment No. 72 (Ref. 820222).
The steam generator (SG) tube pitting and resultant inspection program and tube plugging is addressed in the SE supporting this Amendment.
The associated specific TS changes are described in Section 3.0 of the following SE.
In early 1977, NMEC0 indicated to the NRC staff their intention to change r
-- fuel assembly vendors from Combustion Engineering, Inc. (CE) to Westinghouse Electric Corporation (W).
In March 1980, NNEC0 submitted the Basic Safety Report (SSR), (Ref. 80U306) authorized by W fo" Millstone-2. This BSR in part supersedes the original FSAR that was prepared by CE. Our evaluation and rpproval of the BSR is given in References 810622, 82Qll2 and 820218.
Our evaluation of the Cycle S reload safety analysis (RSA) will not address those issues (e..g., Westinghouse reload fuel design ba;es, rod bowing analyses, etc.) which were resolved in our above referenced approvals of the BSR.
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'Ref erence nuncer made up of year, month and day in that order.
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2.0 Discussion and Evaluation In this evaluation of the Cycle 5 reload using, for the second cycle, fuel assemblies designed and manufactured by Westinghouse in the Millstone-2
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core,- use is made of our generic review of the BSR and various other topical reports. Some of the topical reports have not received formal NRC staff appr. oval.
In all cases where a topical report has not received such an approval, the report has been examined, its methods judged to be reasonable, and an appraisal has been made that a complete review will not reveal the methodology to be significantly in error. On this basis, all topicals referenced are judged to be acceptable for this reload of Millstone-2 and for operation at the licensed power level of 2700 MWt.
2.1 Fuel System Design The objectives of the fue'l system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences, (b) fuel system damage is never so
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severe as to prevent control rod insertion When iY'is ~r'equired, (c) the number of fuel rod failures is not underestimated for postulated accidents, and (d) coolability is always maintained. We has a reviewed the information provided in support of Millstone-2, Cycle 5 operation to determine if these objectives have been met.
The Millstone-2, Cycle 5 core will be comprised of (a) 73 fuel assemblies that were manufactured by Combustion Engineering, the original NSSS vendor, and (b) 144 fuel assemblies supplied by Westinghouse, the Cycle 4 and 5 reload fuel vendor. The Cycle 5 core loading inventory is given in the following table.
Millstone, Unit No. 2, Cycle 5 Core Loading Inventory Initial BOC Assembly Assembly Nunber of Enrichment Theoretical Average Exposure Designation Vendor Assemblies (w/o U235)
Density (%)
(MWD /MTU)
B+
CE 1
2.336 95 17,450 El CE 24 2.730 94.75 24,650 E2 CE 48 3.235 94.75 22,600 F1 W
24 2.697 94.54 13,470 F2 7
48 3.297
' 94.87 9,650 G1 7
24 2.70 95*
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48 3.20 95*
0 217 total
- Region G1 and G2 densities are nominal. Average densities of 94.5% we.re used in the safety analysis.
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3-The fuel management pattern was developed to accommodate a Cycle 4 burnup range of 10,650 MWD /MTU to 12,000 MWD /MTU. Af ter 'the core reload, the beginning-of-cycle core-average exposure will be about 11,430 MWD /MTU making the predicted end-of-cycle core-average exposure about 21,830 MWD /MTU (Ref. 811221).
The Westinghouse reload fuel was designed to be geometrically similar to and compatible with the Combustion Engineering reference fuel. Table 1 of Reference 820218 provides a comparison of the fuel mechanical designs.
2.1.1 Seismic-and-LOCA Mechanical Response As discussed in the Millstone-2 Cycle 4 reload SER (Ref. 801006) and the BSR SER (Ref. 820218), both CE and W performed analyses of the fuel response to combined seismic-and-LOCA loadings.
Each of those analyses was per-formed for a homogeneous core of one type of fuel (e.g., CE or W).
Because Cycle 5 operation of Millstone-2 will involve a h terogeneous core of both 9
CE and W fuel, a mixed-core seismic-and LOCA analysis was required.
The licensee submitted a mixed-core analysis (Ref.liO5'0'1 and 810608), which shows that (a) the maximum deformation occurs in peripheral W assemblies and (b) this deformation does not invalidate the results of the current LOCA analysis.
It should be noted that mechanical response analyses have not been completely reviewed at this time, but that the Task Action Plan for this generic issue (Ref. 781100) provides a basis for continued operation while the issue is being fully resolved.
Because of (a) the present unreviewed status of the underlying primary systems asymmetric loads analysis, (b-) the temporary existence of a mixed core in !1illstone-2, (c) our previous approval of the W analytical methods for the fuel assembly response, and (d) the favorable ana'ytical result reported by the licensee, we consider the nixed-core issue to be adequately resolved for Cycle 5 operation without further review.
2.1.2 CEA and Fuel Assembly Guide Tube Wear Sackground information on fretting wear of CEA (control element assembly) cledding and fuel assembly guide tubes can be found in th'e safety evaluation reports (Ref. 820218 and 801006) that were written on the ilillstone-2, BSR and the Cycle 4 safety analysis report, respectively.
In order to provide continuing assurance of both CEA and guide tube integrity, NHECO has outlined (Ref. 810928) a proposed surveillance program to be per-formed following Cycle 4 operation. The program yill involve a combination of visual, profilometric, and eddy current examinations of 2 to 6 CEAs and 6 to 16 fuel assemblies. The fuel assemblies to be examined will include both CE and W standard sleeved assemblies and demonstration asse;mblies that were
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all positioned in rodded core locations during Cycle 4 operation. The demon-stration assemblies employed in Cycle 4 consisted of 4 CE assemblies of the reduced guide tube flow design and 4 W assemblies having guide tube inserts.
m We believe that the HNECO surveillance techniquesiand pr.oposed program will be adequate for establishing CEA and guide tube integrity.
NNECO has agreed to formally submit the examination results for NRC review within 90 days following MGlstone-2 restart.
In that submittal, we recommend that NNECO describe plans for continuing CEA and guide tube surveillance or provide justification for discontinuing those specific examinations.
On the basis cf an anticipation of acceptable wear measurements from the surveillance program and the fact that all r' dded Cycle 5 fuel assemblies o
(except 4 W assemblies with guide tube inserts) will be sleeved, we conclude that NNECO has provided sufficient justification for Cycle 5 operation.
2.1.3 Cladding Collapse As described in the safety evaluation on the BSR, individual reload applica-tions referencing the cladding creep-collapse analysis of W reload fuel should confirm that the collapse analysis was performed in accordance wi.th the condition of approval placed on the W generic analytical method. That condition involves the use of specified Tnput curves (e.g., initial ovality) for the analysis.
The licensee has stated (Ref. 820204) that the input curves were used as specified by the SER and that the W reload fuel is not predicted to collapse during Cycle 5 operation.
Hence, this issue is satisfied.
The licensee has completed (Ref. 820223) the Cycle 5 cladding col. lapse analysis for the CE fuel.
CE fuel is pressurized to preclude cladding collapse and analyzed.with conservative methods that demonstrate free-standing cladding beyond a 34,500 EFPH exposure, which. bounds the lead fuel rod exposure for Cycle 5.
Therefore, we conclude that no cladding collapse will occur during Cycle 5.
2.1.4 Fuel Manufacturing Problems Dinensional checks to ensure that the reload fuel assemblies are compatible with other core components are a part of the new-fuel receipt inspection progran performed by NUEC0 at Millstone-2.
The inspection of Cycle 5 rei.ad fuel revealed 2 conditions which required a more thorough exanination and which resulted in the need to ship sone fuel &ssemblies back to the Columbia, South Carolina fuel fabrication facility f.or modifications (see Ref. 820108).
The required modifications to the Cycle 5 reload fuel assemblies were not as extensive and unlike those that were previously required for all of the W Cycle 4 reload fuel assemblies (see Ref. 801006).
The first condition which stad a problem was discovered during the envelope inspection of fuel assembi; Lp nozzles.
An onsite upper gauge block (UGB),
which is designed.to seat on top nozzle posts, is used to verify that fuel
.m ycm assemblies will align properly under the upper core plate. The UGB would not seat on some fuel assemblies due to one or both of the following:
(1) top nozzle plates were not parallel to bottom nozzle plates or (2). top nozzle posts were misaligned or irregularly spaced.
Since the UGB is built to require closer seating tolerance than the core plate, W designed and built a gauge block which more closely represented the tolerance needed for the Millstone-2 core plate.
All but 4 fuel assemblies passed the inspection with the W gauge block. These 4 assemblies were subsequently modified at the fuel Tabrication facility.
The second condition that indicated a problem was encountered during the CEA free-path and end-clearance checks. The licensee reported. that most fuel assemblies failed this examination; consequently, measurements of available CEA lateral clearance at the bottom of ' guide tubes were taken on each fuel assembly.
The reduction in clearances in the lowermost portion of the guide tubes is attributable to the use of guide tube end plugs which were'left.over from the Cycle 4 fuel production lot.
Nhen these'end plugs were swaged to the bottom of guide tubes, the diameter of the guide tube walls was locally crimped.
Fortunately, NNEC0's measurements revealed that adequate clearance for CEA operation remained'for all fuel assemblies
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and for all instances of worst-case conditions. The Cycle 6 reload fuel will enploy a different end plug design.
W is considering fabrication modifications that will preclude similar occur-rences in the future, and NRC's Office of Inspection and Enforcement is reviewing quality assurance controls that are used at the fuel fabrication facility.
We conclude that both Cy.cle 4 and Cycle 5 reload fuel fabrication problems arose because of W's inaccessibility to CE proprietary information on the design of MillstoneT2 fuel.
Inasmuch as W has now supplied 2 reloads for Millstone-2, we would not anticipate further problems of this nature in the future.
., 2.1.5 Miscellaneous Analyses We asked NNECO about 2 issues that were not addressed in the reload safety -
analysis report.
Those issues were supplemental ECCS calcula.tions with the cladding models of NUREG-0630. and fuel rod bowing analyses for CE fuel.
The licensee stated (Ref. 820204) that the analyses of these' issues that were performed for Cycle 4 operation are bounding relative to those for the planned Cycle 5 operation of Millstone-2. We accept th,is response without further question.
2.2 Nuclear Design The nuclear design procedures and models used for'the analysis of the Mill-stone-2, Cycle 5 reload core are the same as those used for Cycle 4.
These are documented f a the Millstone-2 BSR and have been approved for the analysis of the Millstone-2 core using W reload fuel beginning with Cycle 4.
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2.2.1 Control R'd Worth o
The control rod worths and shutdown requirements for both beginning and end-of-cycle (E0C) 5 are presented and compared with previous Cycle 4 values.
At E0C 5, the reactivity worth with all control rods inserted assuming the highest worth rod is stuck out of the core is 5.93% Ap assuming a 10% uncer-tainty reduction. The reactivity worth required for shutdown, including the contribution required to cont.rol the steam line break even at EOC 5 is 5.90% Ap. Therefore, sufficient control rod ~ worth is available to accommo-date the reactivity effects of the steam 1.ine br eak at the worst time in core life allowing for the most reactive control rod stuck in the fully withdrawn position and also allowing for calculational uncertainties in these worths based upon comparison of calculations with experiments presented in the BSR and in previous W reports.
On the basis of our review, we have concluded that MEC0's assessment of reactivity control is suitably conser-vative and that adequate negative reactivity worth has been provided by the control system to assure shutdown capability assuming the next mosti reactive control rod is stuck in the fully withdrawn positjgn.
2,2.2 Moderator and Doppler Temperature Coeffici_ents.
The most positive moderator temperature coefficient between 70% to 100%
power has increased to +0.4x10-" Ap/*F from the Cycle 4 value of +0.2x10-4 Ap/ F.
The Doppler coefficient has been extended to -1.92x10-s ap/*F com-pared to the Cycle 4 most negative value of -1.87x10-5 Ap/*F. The maximum delayed neutron fraction has also increased slightly from the previous cycle value.
The maximum differential rod worth of two CEA groups moving together (at hot zero power) has increased from 24.3x10-5 Ap/in to 36.6x10' Ap /in.
These changes, as well as changes in the total trip reactivity as a function of position and the Doppler power coefficient as a function of power, exceed the liriiting range of values established by the Cycle 4 and BSR safety analy-sis.
Therefore, reanalyses of.those transients which are affected by these kinetics parameters were performed (see Section 2.4 of this SE).
2.3 ' Thermal-hydraulic Design The thermal-hydraulic design for Millstone-2 is presented in the BSR (Ref.
800306).
2.3.1 Hydraulic Compatibility As discussed in the BSR, the W Cycle 5. reload fuel assemblies for Millstone-2 are designed, and shown through testing, to be hydraulically compatible with the CE Cycle-3 reference fuel assemblies.
2.3.2 Design Power Level The design power level for Millstone-2, Cycle 5 remains 2700 MWt (the same as for Cycle 4).
The' safety analysis uses a power level of 2754.MWt (102%
power) to allow for measurement uncertainties.
A summary of our evaluation follows.
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s 2.3.3 Analytical Methods 7
The steady state DNB analysis for Cycle 5 was performed using the THINC-I code in conjur.ction with the W-3 correlation (Refs. 7803, 72, 6901 and 7201).
For the W-3 correlation, the 95/95 confidence / probability limit for not suffering _ departure from nucleate boiling is a DNBR greater than 1.30.
In the analysis,. uncertainties in various measured parameteYs were factored in as biases for LC0 and LSSS setpoints.
This biasing of the measurements' uncertainties in the analysis is equivalent to adding the absolute power uncertainties in the various measured parameters and applying the total power uncertainty to the best estimate calculation.
The specific uncertainties along with their equivalent power uncertainties for Cycles 4 and 5, as -deter-mined with the THINC-1 code in conjunction with the W-3 correlation (grid spacer correction = 1.0), and for Cycle 3, as determined with the TORC thermal hydraulic code in conjunction with the CE-1 correlation, are as follows.
Percent Uncertainties Measured Measured Parameter Equiva. lent Power Uncertainty - %
Parameter Uncer tainty Cycle 3 Cycle 4 Cycle 5 Axial Shape Index (ASI) 0.06 ASIU 2.2%
3.0%
3.0%
Pressure 22 psi 0.8%
0.5 0.5 Temperature 2F 0.9 1.0 1.0 F1ow 4%
5.0 2.0 2.0 Power (LCO) 2%
1.4 2.0 2.0
. Pewer (LSSS) 5%
3.5 5.0 5.0 00TE:
Cycle 3 determined with TORC code in conjunction with CE-1 correlation.
Cycle 4 determined with THINC ' code in conjunction wi-th W-3 correlation.
LCO = Limiting Conditions for Operation LSSS = Limiting Safety Systems Settings The following parameters related to LCO and LSSS are the same for Cycles 3, 4 and 5:
power level (2754 MWt), maximun steady state core inlet temperature (551*F), minimum reactor coolant flow (133.7 x 10 8 lb/hr), and maximum allowed initial peak linear heat rate (16.0 kw/ft).
NMEC0 agreed.to provide justification for the measurement uncertainty values
( Axial Shape Index ( ASI), Pressure, Temperature, Flow Power (LCO) and Power
-(LSSS)) for further review of the Cycle 4 and Cycle 5 power uncertainties.
This will'be supplied by March 1,1982.
While our review of measurement uncertainties continues, LC0 and LSSS limits for Cycle 5 will be maintained at the values used for Cycle 3 (Ref. 800603).
We find this acceptable.
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2.3.4-Reactor Coolant Flow The design flow for the Cycle 5 analysis is 370,000 gpm (133.7 x 108 lb/hr at 2200 psi and 551*F) and is the same as the low flow limit included in the TS and analysis for Cycle 4.
2.3.5 Limiting Transient-Complete Loss of Reactor Coolant Flow t
-The loss of flow accident was reanalyzed for. Cycle 5.
The results show that the reactor trip protection provided by the reactor coolant pump. speed sensing system is sufficient to prevent cladding and fuel damage. The DNBR aproaches but does not decrease below 1.30 during the transient.
This is the same result as for Cycle 4.
We find this acceptable.
2.4 Accident Analyses The licensee's analysis of accidents for Cycle 4 was provided in the BSR (Ref. 800306) and the Cycle 4 RSA (Ref. 800603).
Cur approval for-Cycle 4 operation (Ref. 801006) found these accident analyses acceptable.
For Cycle 5', NNEC0 has reanalyzed the boron dilution dvent-t& des ~2', 3 and 4, the CEA eject. ion event, the CEA withdrawal event-!1 odes 1 and 2, the complete loss of reactor coolant flow event and the RCP seized rotor event. The Cycle 5 RSA states that this reanalysis was necessary because of changes in cycle-specific parameters in the area of kinetic characteristics, CEA worths, and core peaking factors (Ref. 81111/). We find that the correct reanalyses of.
accidents have been performed.
By References 820204 and 820301, NNEC0 provided the results of their review to determine the acceptability of transient and accident analyses considering
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the increase of plugged steam generator tubes from 500 to 750 per generator.
They concluded that the resultant change in RCS flow and heat transfer charac-teristics do not change the previously docketed non-LOCA transient and accident analyses.
We have reviewed their presentation and concur with their findings.
The reanalyzed accidents are evaluated as follows.
2.4.1 Boron Dilution Event An inadvertent boron dilution will reduce the boron concentration in the primary coolant which in turn will increase the reactor core positive reac-d tivity.
During power operation, the resulting reactivity insertion will increase the reactor power and automatic safety systems will act to sNt down the reactor and meintain the plant within safety limits. However, a boron dilution event during shutdown will not be m,itigated by any automatic safety systems.
It may ' continue and result in reactor criticality if the operator does not take the appropriate corrective action Nithin the necessary time period.
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-g-In Reference 811117, NNECO indicates that the shutdown margin requirements for Cycle 5 are more limiting than those for Cycle 4 for Modes 1, 2, 3 and 4.
Therefore, the operator action time available before a complete loss of shutdown margin occurs during a boron dilution event is less in these modes for Cycle 5.
The transient was reanalyzed for these hot modes (startup, hot standby, 'and hot shutdown) only.
It is not necessary for power operation
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(Mode 1) as discussed above.
The results of the reanalyses were 64, 24 and 24 minutes to-lose shutdown margin for the startup, hot shutdown and hot standby modes, respectively.
The reduction in these times is due to the decreased Modes 1, 2, 3 and 4 shutdown margin from 3.20 to 2.90.
This decrease results from the Cycle 5 specific analyses. We find these results exceed our 15 minute criteria and' are, therefore, acceptable.
Since the, shutdown margin for Mode 5 (cold shutdown) and Mode 6 -(refueling) has not changed, NNEC0 did not reanalyze the boron-dilution event for these modes.
The staff did, however, request additional information regarding the ability of the installed instrunentation charrnels to detect and alert the operator of a boron dilution event and the resultant operator action time available (Ref. 811224). The information has been provided (Ref.
820204).
Our finding in the Amendment 61 Cycle 4 reload SE (Ref. 801006) was, "The limiting dilution event for the Cycle 4 operation is for the refueling mode with a calculated time to criticality of 34 minutes which is more than the required 30 minutes. Therefore, we find this analysis and its results acceptable for all cases when the reactor is subcritical.
We now conclude that this finding is also applicabTe to Cycle 5 operations.
However, the licensee should be aware that the staff is presently evaluating the need for all operating PWRs to provide additional protection 'against uncontrolled boron dilution events during the shutdown modes.
Pending the outcome of this evaluation, it may be nect:ssary to require addi-
' tional instrumentation to alert the operator of a boron dilution event.
NNEC0 will be notified if any such action is necessary.
2.4.2 CEA Ejection Incident Th'e more. positive moderator temperature coef ficent, between 70*; and 100$
power, required a reanalysis of the CEA ejection incident initiated from hot full power conditions.
The higher total peaking factor after ejection (compared to the BSR value) required a reanalysis of the hot zero power CEA ejection incident.
The results indicated that the Regulatory Guide 1.77 limiting criterion of 280 cal /gm is not exceeded for either case.
We have reviewed the analysis assumptions including the Doppler and moderator coef-fic.ients, delayed neutron fractions, ejected rod worths, hot channel factors and trip reacti'vity insertion and find the analysis to be conservative and the predicted consequences accetable.
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Changes in the Cyc)e 5 trip reactivity curve, delayed neutron fraction, and Doppler power coefficient, as well as in the maximum differential rod worth of two CEA groups moving together at hot zero power, required a reanalysis of the CEA withdrawal incident from a subcritical condition. The results show that the DNBR is greater than the limiting value of 1.30 and, therefore, no cladding damage or fission product release to the reactor coolant system-will result.
2.4.4 CEA Withdrawal at Power Changes in the Cycle 5 trip reactivity curve, delayed neutron fraction, and Doppler power coefficient, as well as in the maximum differential rod worth of two CEA groups moving together, required a reanalysis of the CEA withdrawal incident from power. The results show that the thermal margin low pressure trip provides protection over the full range of r_eactivity insertion rates from 0 to 2.44x10'" ap/sec so that the minimum DNBR remains above 1.30.
We find the CEA withdrawal analyses and conseqyences..a,ccceptable.
2.4.5 Complete Loss of Reactor Coolant Flow A loss of reactor coolant flow could result from mechanical or electrical failure in one or more of the reactor coolant pumps. The immediate effect of reduced coolant flow is a rapid increase in coolant temperature. This heat up in coolant temperature could lead to DNB and subsequent fuel damage if proper protection were not provided.
The lose of flow accident was reanalyzed for Cycle 5 because of the change in the trip reactivity curve for Cycle 5 and changes in the delayed neutron fraction and Doppler and temperature coefficient.
Millstone-2 has provided the following protection against this event:
l.
Reactor coolant pump speed sensing system.
2.
Low reactor coolant loop flow trip.
The reactor coolant speed sensing systen is provided to protect.against loss of power to all punps.
The low reactor coolant loop flow trip is provided to protect for loss of one or two reactor coolant punps.
The licensee has analyzed the transient with three digital computer codes, i.e., LOFTRAN, FRACTRAN AND THINC.
The acceptability of these codes is discussed in Reference 820112.
The results provided in Reference 811117 indicate that for the most limiting loss of flow event the DNBR decreases to a minimum value of 1.31 at approximately 3.7 seconds into the transient.
Core flow at. the time of minimum DNBR is approximately 65% of normal full fl ow.
Although there is a turbine-generator assist feature which would provide a slower coastdown, it was not considered for this analysis.
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.- m c,. - me The licensee has performed an analysis of complete loss of coolant flow tran-sient to determine its impact on the DNBR. The re'sults of the analysis indicate that the DNBR does not decrease below 1.30 during the transient. The results also confirm that the analysis as presented in Reference 800306, Millstone Basic Safety Report, continues to bound Cycle 5 plant operation.
We, therefore, conclude that the results of this analysis are acceptable.
2.4.6 Reactor Coolant Pump Seized Rotor The seized rotor transient was reanalyzed for Cycle 5 because lyf the change in the trip reactivity curve as a function of rod position and other changes in the delay neutron fraction and Doppler and temperature coefficients. The
. accident postulated is an instantaneous seizure of a reactor coolant pump rotor. Flcw through the affected reactor coolant loop is rapidly reduced,-
leading to an intiation of a reactor trip on a low flow signal.
In Reference 811117, the licensee provided the results of the analysis to
-demonstrate that the integrity of the primary coolant system would not be endangered since the peak reactor coolant system pretsure, approximately 2500 psia, is less than 110% of the RCS design pressurere The peak clad temperature of approximately 1960*F is much less tha'n 2700*F (Millstone-2 fuel design temperature limit) which guarantees that the core will' remain intact with no loss of core cooling capabiltiy following the accident.
The results also indicate that less than 2 percent of the fuel rods are predicted to experience departure from nucleate boiling.
Since the Cycle 5 plant response to a reactor coolant pump seized roter tran-sient is within the reactor coolant' system pressure and fuel limits, we conclude the results are acceptable.
2.4.7 Steam Line Rupture Accident The steam line rupture accident was reanalyzed for Cycle 5 because of the change in shatdown margin, trip reactivity curve, and kinetics coefficient.
This transient is the most limiting case which assumes the steam line rupture inside the containnent at the outlet of the steam generator.
The plant initially it at no load condition with offsite power available.
The analysis was performed with the assumption that auxiliary feedwater (AFW) flow would be initiated autonatically during the transient.
It was assuned that 2800 gpq of AFW, 355 more than the maximun runout flow, would be delivered to the affected steam generator at three minutes after the beginning of the transient. This is conservative with respect to the expected time of AFW initiation since auto-matic actuation of the AFW system would occur on a low steam gene'rator water level trip signal. The assumption was also made that the minimum capabi.ltiy for injection of boric acid solution (1720 ppm) corresponds to the most restrictive single failure in the safety injection pump and one low pressure O
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safety injection pump delivering full fic./ to cold leg header. Results provided in Reference 811217 show that the reactoi core returns to critical i
after CEA insertion (assuming the most reactive CEA is stuck in the with-drawn position). This is due to the high cooldown rate, resulting from the steam discharge and auxiliary feedwater addition in the presence cf a negative moderator temperature coefficient. However the addition of boron.from the high pressure safety injection pump brings the core suboritical again.
The peak heat flux attained during th'is transient is small, approximately 3 per-cent, and the DNBR margin design basis of.1.30 will not be violated. The maximum pressure within the reactor coolant boundary and the main steam system would nut exceed 110 percent of the design pressure.
We conclude that appropriate analysis has been provided for this transient and the results of the analysis are acceptable.
2.5 Loss' of Coolant Accident i
By letter dated February 19, 1982 (Ref. 820219), NNEC0 provided the LOCA analysis with additional plugged steam generator' tubes. The analysis was erformed with the approved version of the W evaluation model (1981) assuming p'02% licensed core power rating and with 9.T% steTigenerator tubes (800 l
tubes per steam generator) plugged. Modification to computer input included reduction in the primary steam generator flow area and volume. One large break calculation is appropriate for this type of reanalysis. The model changes properly reflected changes in plant conditions.
We find this re-analysis has been performed in accordance with 10 CFR 50 Appendix K and l*
related staff positions and, therefore, the revised large break LOCA analysis is acceptable.
NNEC0 states (Ref. 820204-2) that the small break LOCA analysis was reviewed i.
and it has been determined that input parameters were assuned for each steam generator which were equivalent to having approximately 1000 plugged tubec per stean generator.
Therefore, they find the current small break LOCA
... analysis results remain valid for Cycle 5.
We concur in this finding.
1 2.6 Radiolooical Consequences of Postulated Accidents i
We have reviewed the BSR, RSA and the other submittals suppo' ting Cycle 5 r
operation and find the potential radiological consequences of design basis accidents to be appropriately bounded by the original May 10, 1974 Safety Evaluation or by the Cycle 3 Reload Safety Evalution.
Since the guidelines
.of 10 CFR Part 100 continue to be met, we find the potential consequences acceptable.
3.0 Technical Specification Changes NNEC0 proposed the TS changes necessary for Cycle 5 operation in References 811217 and 820114.
A large number of the proposed changes not specifically related to the reload'were. issued by Amendment No. 72 (Ref. 820222). As stated in earlier portions of this SE,. the majority of the Cycle 5 analysis using W_ fuel is, as was the case for Cycle 4, bounded by the Cycle 3 analysis A
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where CE fuel was reloaded.
The necessary TS changes are as follows.
3.1 Shutdown Margin NNECO proposed a change-in the Modes 1 through 4 shutdown margin from 3.2%
t.K/K (Cycle 4) to 2.9% LK/K for Cycle 5 (Ref. 820114). This value is a direct result of the cycle specific analyzed core charac.teristic. We find s
this change justified by the analysis.
The TS pages affected are 3/41-1 and B 3/4 1-1.
3.2 Moderator Temperature Coefficient
. Reference 811217 Item Ho. 7 proposes to' change the moderator temperature coefficient (MTC) limit in TS 3.1.1.4 from less positive than 0.2 x 10-"
to 0.4 x 10-" aK/K/*F whenever thermal power is > 70*; of rated. This change is the direct result of the Cycle 5 core characteristics and is supported by normal and accident analyses.
Therefore, we find this proposed change
- to Page 3/4 1-5 acceptable.
Itam 8 of the same reference requests removal of TSesurveillance require-ment (SR) 4.1.1.4.2c.
This SR is to perform a MTC determination at mid-cycle.
NNEC0 stated that they have successfully demonstrated their capability to predict the MTC through four fuel cycles with two fuel vendors (CE and W).
They also contend that:
(1) the MTC measurement at beginning of cycle --
ensures that no unforeseen core characteristics exist; (2) MTC testing is a high risk plant test involving significant CEA novement and axial shape index shifts; and (3) the MTC test requires a loss of production of' about 6% for approximately 4 days.
We find that the MTC determination at mid.-cycle is of marginal value and therefore, because of the adverse aspects of such a test, should be deleted from Page 3/41-6.
3.3 Withdrawn Position of Regulating CEAs
- The current TS 3.1.3.6 requires that r3gulating CEAs shall be limited to the withdrawal sequence show on Figure 3.1-2, the CEA insertion limit.
NNECO's.
Reference 811217 proposal is to clarify this specification by indicating that the fully withdrawn position is greater than or equal to'176 steps.
This is the approved fully withdrawn position of the shutdown CEAs (see TS 3.1.3.5).
We find the proposed change to TS Page 3/41-28 will have no adverse effect on reactivity insartion or peaking factors and is, therefore, acceptable.
3.4 Pressurizer level Control Amendnent No. 66 (Ref. oiU407) changed numerous TS, pages including Page 3/4 4-4 to implement TMI Category A requirements.
The change to Page 3/4 4-4 was to requ' ire at least 130kW of pressurizer heater capacity and level within +5%
of its programmed value. This last change has proved to be impractical-
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mu.twmm
- o since. pressurizer level cor. trol during transient operation (startup, power
- level changes, trips, etc.) is not that precise.
NNECO has proposed (Ref.
811217 Item 11) to footnote an exception for level control during transient operations. Although this is one way to correct the TS, we do not believe it is the best.
Current STS give only a minimum acceptable level (enough water. to prevent heater damage) and a maximum level (enough steam to prevent solid conditions).
In discussions with NNECO, they agree. that the new STS coul,d provide a better basis, and have initiated their procedure to propose.
such a TS change including the appropriate STS surveillance requirements.
However, since this action will take time and TS violations may occur during this time period, both the licensee and the staff agree to the proposed TS nodification to Section 3.4.4 until NNECO can make another proposal.
We find the proposed changes to TS Page 3/4 4-4 acceptable.
Environmental Consideration WC have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environnental impact and, pursuant to 10 CFR $51.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion Ue have concluded, based on the consideratiuns discussed above, that: (1) because the amendment does not involve a significant increase in the proba-bility or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable ' assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not 5e inimical to the common defense and security or to the health and safety of the public.
Date: MAR 5 1932 Principal Contributors:
Dale Powers Vince Leung Harry Balukjian Norm Lauben Larry Kopp Monte Conner e
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6.0 References The following references are listed in the chronological order.of the date of the transmittal letter. The first two digits are the year, the next two the month and the last two the day of the month.
Reference No.
Descr'iption 6901 Westinghouse Report, WCAP-7015, Subchannel Thermal Analysis of Rod Bundle Cores, H. Chelemer, J. Weisman and L. Tong, January 1969.
7200 Commission Report Boil'ing Crisis and Criticai Heat Flux, L. Tong,1972.
7201 Westinghouse Report, WCAP-7838, Application of the THINC Program to PWR Design, J. Shefch,eck, January 1972.
Westinghouse Report, WCAP-9272, Y778.~ Reload Safety Evaluation 7803 tiethodology, F. Bordelon, Aarch 7811 NRC Report, NUREG-0371, Task Action Plan for Generic Activi-ties - Category A, November 1978.
l 800306 NflECO Letter transmitting Westinghouse Basic Safety Report (BSR), W. Counsil to R. Reid, March 6,1980.
i 800603 tu!ECO letter transmitting Cycle 4 Refueling Safety Analysis, W. Counsil to R. Clark, June 3,1980.
801036
!!RC letter transmitting Amendment No. 61, Cycle 4 Reload l
Evaluation, R. Clark to W. Counsil, October 6,1980.
810108 NNEC0 letter, Resolution of Cycle 4 Startup Commitment on 11easurement Uncertainity Values, W. Counsil to R. Clark, January 8,1981.
l 810401 N!!ECO letter, Esclution to Cycle 4 Startup Commitment on LOCA Asymmetric Blowdown Loads,~ W. Count!1 to R. Clark, April 1, 1981.
810407 NRC lssues Amendment No. 66, TMI Category A TS,.R. Clark to W. Counsil, April 7, 1981.
810608 NNEC0 letter, Results of LOCA As'ymmetric Blowdown Loads for Mixed Core, W. Counsil to R. Clark, June 8,1981.
810622 NRC ' letter, Acceptance of Physics Portion of BSR, R. Clark to W. Counsil, June 22, 1981.
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Reference No.
Description 810720 Westinghouse letter, large Break LOCA Results for Millstone-2, E. Rahe to N. Lauben, July 20, 1981.
810917 NNECO letter, CEA Guide Tube Wear, W. Counsil to R. Clark, September 11, 1981.
810928 NNECO letter, Resolution of. Cycle 4 Startup Commitment on CEA Guide Tube Wear Evaluation Program, W. Counsil to R. Clark, September 28, 1981.
811015 NNEC0 letter, Resolution of Cycle 4 Startup Commitment on Cycle 5 Reload Outage Steam Generator Inspection Program, W. Counsil to R. Clark, October 15, 1981.
811016 NNEC0 letter, Resolution of Cycle,4 Startup Commitment on Worst Large Break LOCA Burnup of Westinghouse Fuel in CE Designed Core, W. Counsil to R. QJgrk,,0ctober 16, 1981.
- 811117 NNECO letter, Cycle 5 Reload Safety Analysis (RSA), J. Cag-netta to R. Clark, November 17, 1981.
811201 NRC Approval of Westinghouse Appendix K Evaluation Model (Ref. 810515), J. Miller to E. Rahe, December 1,1981.
811202 NNECO Application for Shutdown Cooling an'd Coolant Circula-tion During Refueling, R. Werner to R. Clark, December 2, 1981.
811317 NNECO Application for Cycle 5 Technical Specification Changes, J. Cagnetta to R. Clark, December 17, 1981.
811218 NRC Issues Amedment No. 71, Shutdown Cooling and Coolant Circulation During Refueling, E. Conner to W. Counsil, December 18, 1981.
811224 NRC letter, Request for Additional Information on RSA, C. Trammell to W. Counsil, December 24, 1981.
820108 NNEC0 letter, Cycle 5 Reload Fuel, W. Counsil to R. Clark, January 8,1982.
820112 NRC letter, Acceptance of Section 5.'3.2 through 5.3.9, 5.3.13 and 5.3.15 through 5.3.17 'of the BSR, R. Clark to W. Counsil, January 12, 1982.
820114 NNECO Application for Additional Cycle 5 Technical Specifi-cation Change, W. Counsil to R. Clark, January 14, 1982.
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17 -
Reference No.
Description 820204-1 NNECO. letter, Additional Information on Cycle 5 Reload,.
2 W. Counsil to R. Clark, February 4,1982.
820204-2 NNECO letter, Transient and Accident Analyses with Addi-tional Steam Generator Tubes Plugged, W. Counsil to R. Clark, February 4,1982.
820218 NRC letter, Acceptance of Remaining Sections of BSR, R. Cla'rk to U. Counsil, February 18, 1982.
820219 NNEC0 letter, Large Break LOCA/ECCS Performance Results with Additional Steam Generator Tubes Plugged, W. Counsil to R. Clark, February 19, 1982.
820222 NRC Issues Amendment No. 72, Cycle 5 Miscellaneous Tech-nical Specification Changes, E. Conner to W. Counsil, February 22, 1982.
820223 NNECO letter, fission Gas / Clad Collapse Considerations During Cycle 5, W. Counsil to R. Clark, February 23, 1982.
820301 NNECO letter, Supplemental Information on Plugged Steam Generator Tubes, W. Counsil to R. Clark, March 1,1982.
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