ML20149E021
| ML20149E021 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/30/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149E026 | List: |
| References | |
| TAC-49798, TAC-53462, NUDOCS 8801130200 | |
| Download: ML20149E021 (13) | |
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NUCLEAR REGULATORY COMMISSION 3
- C WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ISSUANCE OF AMENDMENT N0. 90 TO DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY, EL AL.
MILLSTONE UNIT 2 DOCKET NO. 50-336 1.
Introduction and Background Ir Reference 1, Northeast Nuclear Energy Company (NNECO or licensee) sub-mitted a license amendment request and the Reload Safety Analyses (RSA) in support of the Millstone Unit No. 2, Cycle 6 reload.
As indicated in
- ne submittal, the bases on wnich the Cycle 6 reload was analyzed were accumented in a "Basic Safety Report" (BSR) (Ref. 2).
The BSR, as sup-piemented by Reference 3, serves as the reference fuel assembly and safety analysis report for the use of Westinghouse fuel at Millstone 2 (a Combus-tion Enginearing plant).
Reference 4 documents the NRC staff's review and acceptance of the BSR.
By Reference 5, NNECO infonned the staff that due to the elevated levels of radioactive lodine and other fission products identified during Cycle 5 coeration, NNECO anticipated the discovery of a number of fuel assemblies with leaking fuel rods during the 1983 refueling outags.
Since that time, NNEC0 perfonned fuel sipping identifying 26 fuel assemblies with fa Ged fuel rods.
In addition, visual examinations revealed 15 fuel assemblies to have broken holddown springs.
Further, strucural damage was observed in two assemblies, one of which a~tso had a broken holddown spring.
This damage was reported to the staff in, Licensee Event Reports 50-336/83-25, 83-25/01-T, 03-26, and 83-26/01-T.
Reference 5 provided a detailed discussion of the fuel degradation.
As riiscussed in Reference 6, NNEC0 is replacing all leaking fuel assemblies with a combination of new and previously discharged fuel as:;.amblies. These
&nges have necessitated a revised loading pattern for tycle 6 operation.
In.sddition, assemblies F37 and F73, which sustained some structural danage, are being replaced.
J By Reference 7, NNECO reported damage to the themal shield suppsrt afstem a t MiDstor.e Uni t No. 2.
The extent of this damage resulted in the need for l
removal of the thermal shic1d from the core barrel. Reference d provides details of NNECO's thennal shteld danage recovery program.
l In order to assess the impact of a new loading pattern and the recovsl of the thermal shield, NNEC0 has had its. fuel vendor reevaluate the Reference 1 Reload Safety Analyses in support of Millstone Unit No. 2 Cycle 6 operation.
I The results of this review were provided as a supplement to the Reload Safety Analyses (Reference 6).
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. 1.1 General Description of Reactor The Millstone 2 reactor core is comprised of 217 fuel assemblies.
Each fuel assembly has a skeletal structure consisting of five (5) Zircaloy guide thimble tubes, nine (9) Inconel grids, a stainless steel bottom nozzle, and a stainless steel top nozzle. One hundred seventy-six fuel rods are arranged in the grids to form a 14x14 array.
The fuel _ rods consist of
'slightly enriched uranium dioxida ceramic pellets contained in Zircaloy-4 tubing which is plugged and seal welded at the ends to encapsulate the fu el.
Nominal core design parameters utilized for Cycle 6 are as follows:
Core Power (Mv!t) 2,700 System Pressure (psia) 2,250 Reactor Coolant Flow (GPM) 350,000 Core Inlet Temperature (*F) 549 Average Linear Power Density (kw/f t) 6.065 (based on best estimate hot, densified core average stack height of 136.4 inches)
The core loading pattern for Cycle 6 consists of twenty-four (24) interior feed assemblies containing 2.7 w/o U-235 and forty-eight (48) peripheral feed assemblies containing 3.2 w/o U-235.
These are replacing seventy-two (72)
Combustion Engineering (CE) batch E assemblies.
Due to fuel defects in Cycle 5 and subsequent symmetry consideration, sixteen (16) interior feed assemblies containing 2.70 w/o U-235, twenty (20) CE assemblies from Batch A and one (1) CE assembly from Batch B (these CE assemblies were discharged at the end of Cycle 1) are needed as well.
2.
F uel System Design Tne fuel system design for Millstone Unit 2, Cycle 6 is the same as tnat approved (Ref. 4) for Cycles 4 and 5.
That is, approval of the BSR constituted approval of the use of a mixed core of Combustion Engineering and Westinghouse fabricated fuel assr$ lies.
The replacement of CE fuel with Westinghouse fuel at each reloading would eventually lead to a core with all Westinghouse fuel.
The failed fuel assemblies at Millstone necessitated a revision to the reload plan such that a mixed core, as described in Section 1.1, results. The reload redesign is a result of tne following:
1.
Fuel rod failures in 26 assemblies 2.
Removal of two damaged fuel assemblies 3.
Removal of the thermal shield i
4.
Failure of holddown springs in 15 fuel assemblies,
' As described in Reference 9, the reload redesign utilizes a combination of ne,4 and previously discharged fuel assemblies to replace the leaking and broken fuel assemblies.
Since this redesign uses previously approved, fuel assembly types nd since the redesign and the reinserted CE assemblies will not receive c
greater than design exposure, the redesign is acceptable from tne ruel system point of view.
The NRC was infonned of the broken holddown springs identified on 15 fuel assemblies by Reference 10. A summary of information discussed at a meeting on October 12, 1983 on the broken holddown springs was presented in Reference 6.
At this meeting, NNEC0 documented plans to evaluate the replacement of the broken holddown springs.
A repair procedure and tooling was developed te effect the replacement of the holddown springs on irradiated fuel assemblies.
This procedure was utilized successfully on one fuel assembly.
However, NNECO decided that the irradiated fuel repair procedure involved a high risk with the potential for damaging fuel assemblies, particularly fuel pins, during the repair.
NNELO thereiure reacned the conclusion and provided supporting analysis (Ref.
- 11) that operation of Cycle 6 with 9 fuel assemblies, each with a single broken holddown spring, is acceptable and prudent.
The analysis provided by NNECO characterizes the breaks to the noidoown springs, provides justification that the breaks were caused by excessive vibratory motion during reactor operation, discusses fretting wear, loose parts, control rod,iammina and the probability of multiple fractures, anu cunauues t9at operation of Cycle 5 with the 9 assemblies having broken holddown springs would be acceptable.
Thi is primarily because the number of active turns of the springs is only siim,y ceneased by the types of breaks observed.
Future new fuel will am e.'ewiv designed springs.
We have reviewed the material provided by NNEC0 and agree with the conclusion that operation of Cycle 6 with 9 assemblies containing broken holddov.n springs will not pose a significant reduction in satety oi cue power piant.
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- 3. 3 Nuclear Desian Tne nuclear design procedures and models used for the analysis of the itfilstone Unit 2 Cycle 6 reload core (Reference 1) are the same as those used for Cycle $.
These are documented in tne bli sstune unit 2 nasic Safety Report (85R) 1 (Reference 2) and have been approved (Reference 4) for the analysis of the i
tiillstone Unit 2 core using Westinghouse reload feel Seginning with Cycle 4.
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Tne licensee proviaeu a taouiar summary (Table 2, Reference 1) of the changes in the Cycle 6 kinetics characteristics compared with the current limits based on the most limiting BSR safety analysis and the Cycle 4 and 5 analyses.
All of the Cycle 6 values fell within the current limits. The kinetics parafteters were, therefore, acceptable for use in the Cycle 6 accident analysis because they are calculated with approved methods, and they are within the bounds of values previously approved.
The reanalysis of the reload per1onned as a result of toe tuee idliures (Reference 9) and removal of the thennal shield was performed witn tne saae approved techniques discussed above.
In Reference 9, Table 2 the kinetics paro.aeters Tor the Cycle 6 reload redesign are given.
These are all within toe current sialts witn a small exception in the least negat ye and above 30%
N power doppler temperature coefficients and the maximum delayed neutron fraction.
The licensee examned the effects of these changes on accident analyses in Reference 9, pages 7 and 8, with the conclusion that tne potential etrects were small, and no reanalyses were necessary.
We reviewed these evaluations and agree that the small changes in these parameters do not lead to a need for reanalyses of any accidents, and that the revised fuel loading and removal of the thennal shield is acceptable with respect to nuclear design.
The control rod worths and shutdown requirements for the Cycle 6 redesign and the initial Cycle 6 design are presented in Table 3 of Reference 9 and compared with previous Cycle 5 values.
At E0C 6, the reactivity worth with all control rods inserted assuning the highest worth rod is stuck out of the core is 6.00% assuming a 10% reduction to allow for uncertainty.
The reactivity worth required for shutdown, including the contribution required to control the steanline break event at E0C 6 is 5.92%
Therefore, sufficient control rod worth is available to accommodate the reactivity effects of the steamline break 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.
We have reviewed the calculated control rod worths and the uncertainties in these worths based upon comparison of calculations with experinents presented in the BSR and in previous Westinghouse reports.
On the basis of our review, we have concluded that the NNEC0's assessment of reactivity control is suitably conservative and that adequate negative reactivity worth has %en provided by the control system to assure shutdown capability assuning the most reactive control rod is stuck in the fully withdrawn position.
The total trip reactivity as a function of position calculated for Cycle 6 was more limiting than that calculated for Cycle 5.
TLa Cycle 6 curve was there-fore used in all accident reanalysis.
4.0 Thermal-Hvdraulic Desian W11stons 2 Cycle 6 utilized the Basic Safety Report (Ref. 2) which was approved by the staff in Reference 4.
The Basic Safety Report was also used as the basis for Cycles 4 and 5 operation.
As discussed in the BSR, the Westinghouse fuel assemblies have been designed and shown through testing to be hydraulically conpatible with all resident fiillstone 2 fuel assc,blies.
A detailed discussion is given in the staff SER of Cycle 4 dated October 6, 1980 (Ref. 12).
The DNB analysis fnr Cycle 6 was performed for a minimum reactor coolant flow rate of 350,000 gpm and a radial peaking factor, F, of 1.565. A reduction in flow from 370,000 gpm to 362,600 gpmandaconservEtivereductioninF II
- r 1.63 to 1.597 was previously implenented during Cycle 5 operation. As indicated by the power and flow sensitivities reported in the Cycle 4 Reload Safety Evaluation Report (Ref.13) a flow reduction can be offset by a power (or F ) reducticn in a 2:1 ratio to maintain a constant DNBR.
Thus the reducfion in flow has been more than offset by the reduction in radial peaking factor and this has been confinned by the licensee in their Cycle 6 analysis.
The Cycle 6 analysis takes a partial credit of 3.0% of the net conservatism which exists between convoluting and summing the uncertainties of various measured plant power parameters in terms of power. This p$rtial credit was applied in previous cycles and its approval is discussed in more detail in
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the Cycle 4 Reload Safety Evaluation Report (Ref 13); therefore, we find operation of Cycle 6 acceptable.
5.0 Accident Analysis 5.1 CEA Withdrawal at Power The CEA withdrawal'at power accident was reanalyzed for Cycle 6 to assess the impact of increased steam generator. tube plugging'and the corresponding reduction in flow. The results of this analysis show that the thermal margin low pressure trip maintains the minimum DNBR above 1.30 over the full range of reactivity insertion rates, which is acceptable.
5.2 LOCA Reanalysis with 15.3% Steam Generator Tube Pluggi,rLq Millstone Unit 2 is currently licensed to operate at a peak core linear power of 15.6 kw/f t with up to 9.4% of the steam generator tubes plugged (reference 14)
In anticipation of further degradation in the steam generators, the licensee has submitted a revised large break and small break LOCA analysis for 15.3% tube plugging (Reference 15).
The limiti. g large break (C d=0,6) was calculated with the currently approved n
Westinghouse evaluation model (EM), and found to conform to the acceptance cri-teria of 10 CFR 50, Appendix K.
The in. crease from 9.4% plugging to 15.3% plug-ging produced a 100F increase in the calculated peak clad temperature (from 20450F to 20550F). We asked the licensee to explain this surprisingly small increase in PCT, in view of previous Westinghouse EM calculations showing much higher sensitivity of PCT to tube plugging beference 3).
The licensee responded that Combustion Engineering plants, n as Millstone 2, differ from Westinghouse plants in seveal important respe..s.
First, the worst case large LOCA for Millstone 2 is with the primary coolant pumps running, and the blow-down is less sensitive to the resistance of the steam generator.
- Further, the C-E containment pressure is a few PSI higher than for W plants, and the locked rotor resistance is lower. Because of these two facts, the calculated reflood rate never falls below 1 inch /second, and the rem;irement to use steam 0
cooling i: not invoked. The resulting calculated PCT ar 2055 F is considerably lower than the value used for the sensitivity study in reference 3.
At the lower PCT, there is considerably less zirconium-water reaction, and peak clad temperature is less sensitive to system changes such as tube plugging.
The effect of additional tube plugging (up to 18%) on the limiting small break was also reevaluated and shown to be minimal.
Based on the LOCA analysis, we conclude that operation with 15.3% steam genera-tor tube plugging does not require a reduction in the technical specification limit of 15.6 KW/f t peak linear power.
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. 5.3 LOCA Consecuences After Removal of Thermal Shield By letter dated November 17, 1983, (Reference 17) NNEC0 submitted a supple-ment to its Cycle-6 reload application with regard to removal of the thermal shield from the core barrel. The licensee presented qualitative arguments to support the validity of previous LOCA analyses which did not account for the removal of the thermal shield, The following are those arguments.
Removal 0f the thermal shield increases the downcomer coolant volume by 120 ft.3 For a small break LOCA event, the added inventory is beneficial since additional margin to core uncovery is provided.
For those break sizes resulting in core uncovery, the added margin would lead to a lower calculated peak clad temperature.
The staff therefore does not require reanalysis of small break LOCAs.
Removal of the thermal shield could result in a somewhat higher calculated peak clad temperature for a large break LOCA.
This is attributed, in part, to additional time required to replenish coolant to the downcomer prior to beginnino of reflood.
The licensee estimates a 2 second delay in the time to bottom of core recovery (BOCREC). Since the calculated peak clad tempera-ture (PCT) prior to removal of the thermal shield is only 20550F, there exists ample margin to the 10 CFR 50.46 limit of 22000F.
We are confident, therefore, that reanalysis of the limiting large break with the thermal shield removed would not result in a calculated PCT in excess of 22000F.
We therefore find LOCA consecuences of removing the thermal shield are acceptable on con-dition that the licensee confirm, prior to next refueling outage, either that the calculated peak clad temperature does not increase by more than 20vF or submit an ECCS analysis for the limiting large break (per Section II.1.b to 10 CFR Part 50, Appendix K).
5.4 Steam Line Breaks The steam line break event analyzed in support of the Cycle-6 reload was calculated by Westinghouse.
Based on the licensee's response to NRC question 440.1, it appears that Westinghouse did not model thermal-l hydraulic and neutronic asymmetry. A large steam line break will lead to conplex asymmetric thermal-hydraulics and neutronics within the reactor vessel.
This will result in greater moderator feedback at the core quadrant nearest the affected loop (loop with the broken steam line).
In response to question 440.2, the licensee stated:
...when mixing is good, the upper head temperature tends to fall at a slower rate, due to the fact that more flow from the cold loop is allowed to mix with the hot loop flow in the inlet and outlet of the vessel.
Part of the inlet flow is routed to cool the upper head.
Since this water is warmer than in the poor mixing case, the upper head temperature would be higher.
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- s. The licensee assumed homogeneous reactor coolant mixing in,the upper head of the reactor vessel when analyzing the steam line break event.
After the pressurizer is emptied of liquid inventory, the primary system pressure is governed by the saturation temperature in the upper head.
Neglecting a separate model for the upper head and its associated metal wall heat capacity, the calculated system depressurization could be non-conservatively low (by 300-400 psi), thereby resulting in excessive ECCS injected boron.
The lack of upper head mixing was observed in the St.
Lucie-1 natural circulation cooldown event and resulted in the NSSS vendor (C-E) changing its analytical model.
The Westinghouse methodology did no'
'unt for this in the steam line break analysis.
We require justi n that the analytical model used adequately addresses this phenon-The licensee assumed complete mixing of the fluid from tne intact and affected coolant loops as it enters the reactor vessel.
Both Combustion Engineering and Westinghouse have developed proprietary data to credit some mixing, which is advantageous for DNBR consideration.
Assuming ideal or complete mixing requires additional justification.
The licensee analyzed the limiting steam line break to occur for zero power conditions with offsite power available. Qualitative discussions were presented in response to NRC question 440.1 to address steam line breaks with loss of offsite power.
The licensee stated:
Since the reactor coolant pumps are coasting down with the loss of offsite power, the ability of the emptying steam generator to extract heat from the reactor coolant system is reduced.
The closest approach to criticality would occur later in the transient and the core power increase would be slower than in the similar case with offsite power available.
The staff agrees that during natural circulation the primary system will depressurize less. However, it is not obvious that the event with loss of offsite power would not be more severe,'when accounting for asymmetric thennal-hydraulics.
We base our concern on previous vendor and staff calculations which showed 50*F to 100 F lower coolant tempera-tures in the affected loop during pump coastdown versus the case with offsite power available.
Consequently, the moderator reactivity feed-
-back could be significantly different, with the pump coastdown event being more severe. Since Westinghouse specifically designed a mitigating system for its plants which initiates a safety injection actuation signal during a steam line break event, the above concerns may not be applicable to Westinghouse plants. However, Millstone 2 was designed by Combustion Engineering, does not have the Westinghouse protective system, and consequently has its ECCS initiMed much later into the transient (relative to a Westinghouse plant). Consequently, the case with loss of offsite power for Millstone 2 has a potential for being limiting.
We request further justification to demonstrate the acceptability of the steam line break analysis performed by Westinghouse.
In addition, we request confirmation that General Design Criterion 17 (GDC-17) is met for
the steam line break event. This requirer an assessment of the steam line break event with loss of offsite power.
5.5 CEA Ejection and Seized Rotor Events The licensee's evaluation of the CEA Ejection and Seized Rotor events 'is not entirely acceptable. The events were assessed on a peak clad temperature (PCT) criterion. Unless well founded and clearly acceptable fuel failure criteria based on a PCT'and pellet-clad-interaction (PCI) are presented, all rods experiencing a DNBR below the 95/95 limit should be assumed to fail. This criterion has been a long standing NRC Dosition.
We therefore request a confimatory evaluation to demonstrate that the offsite radiological consequences are acceptable assuming failure of all fuel rods that have a calculated DNBR below the 95/95 DNBR limit.
5.6 Steam Generator Tube Rupture
~The steam generator tube rupture event was analyzed by the licensee using the RETRAN comouter In support of the RETRAN nodaTintion for 1111Tstone proaram, licensee submi'tted a calculation of a 2, the
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turbine trip event and compared it with plant data.
The calculated parameters of interest (i.e., pressuro, temperature and pressurizer level) were in good agreement with the data.
However, the transient was mild and did not challenge the model such that its applicability to rore severe events is established. We therefore require the licensee to demonstrate the applicability of the model to a steam generator tube rupture event.
Specifically, the nodalization of the upper head may be inappropriate i
for events leading to voiding within that region.
Thermal-hydraulic behavior in the upper head could significantly alter the consequences of such an event.
Ideal nixing of fluid in the upper head with the upper plenum coolant may be inappropriate. Similarly, lumping the steam generator inlet plenum with half of the steam generator tubes could result in improper primary systen thermal conditions.
Staff evaluations with similar codes have show that finer nodalization is reovired for modeling such events.
The initial conditions for the steam generator tube rupture analysis-consisted of nominal operating conditions and instrumentatien uncertain-ties. This initialization may be appropriate for a nost probable consequences assessment but may not be appropriate for licensing evalua-tions. Licensing calculations should bound the operating conditions of i
the plant. Typically, these are detemined by the tecnnical specifica-tions for plant operation. The licensee should perfo m the analysis at the technical specification limits or otherwise show tnat the conditions at which the plant was analyzed provide bounding results for all allowed plant operating conditions.
The licensee has not incorporated the limiting single failure in its steam generator tube rupture analysis. The consequences of the limiting active failure with and without offsite power available must be assessed.
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6.0 Open items Requiring Confirmatory Resoonse We have reviewed the licensee's submittal of the Cycle-6 reload for Millstone 2 and, to a limited extent, the qualifications of NNECO to perform licensing analyses.
We require the licensee to demon-strate that the analyses submitted in support of the Cycle-6-reload were conservative and comply with regulatory recuirements and criteria.
In addition, the licensee must demonstrate qualification for.perforning licensing analyses and demonstrate understanding of plant responses under transient and accident conditions, as requested in NRC Generic Letter 83-11.
ccpy of the Millstone-2 RETRAN deck for the stean generator tube A
rupture analysis should be submitted to the NRC.
The staff will examine the input and modeling techniques as part of the qualification review of the licensee.
Tne licensee has reanalyzed the steam line break, the CEA ejection, the seized rotor and the steam generator tube ruptureevents in support fo" the Cycle-6 reload.
However, the licensee did not evaluate the conse-quences resulting from a postulated loss of offsite power for the above events, as required by GDC-17 nor has the licensee postulateo the limiting active single failure in accordance with 'present and past regulatory practices.
The staff is unable to conclude on the acceptability of the submitted analyses unless the licensee either:
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(a) [Reanaiyzes the abo've events. assuming.the limiting single failure, with and without a loss of offsite power (as required by GDC-17) while using an acceptable model, or (b) Provides justification and/or requests an exemption for i
devicting from current regulatory requirements.
Conditional upon the licensee's commitment to acceptably respond to our j
concerns listed above, we believe that the confirmatory analyses would not substantially alter the conclusions for these events and the continued operation of the plant does not endanger the health and safety of the public.
Our acceptance of the confirmatory submittal will be conditional upon an acceptable inspection of the licensee's quality assurance (QA) program as applied to computer code development and use practices. This i
inspection will be performed by NRC Region IV during the first week of Februa ry,1984 In addition, NRR will further assess the qualifications of NoFtheast Utilities to perforn licensing submittals with the RETRAN computer program. Northeast Utilities should provide a detailed presentation, at the time of the QA inspection, of their analytical cualifications and understanding of plant responses to postulated transient and accident events. We also request that during that meeting, the licensee submit a copy of their RETRAN deck for the steam cenerator tube rupture calculation.
4
. 7.0 Technical Specification C;.ar.gc?
Technical Specification :hanges proposed by the licensee in Reference 1 are acceptable as follows. No additional Technical Specification changes were required as a result of the reload reanalysis.
A.
Reduced Reactar Coolant Flow Rate - This proposed change affects pp. 2-2, 2-4, and 3/4 2-14 of the Technical Specifications.
It involves lowering the required primary coolant flow rate from 362,500 gpm to 350,000 gpm. This new lower flow is established to correspond to a plugging level of 2500 steam generator tubes, and was used in the Cycle 6 analysis, lle find it acceptable since it was offset by the reduction in F *r B.
CEA Drop Time - This proposed change to p. 3/41-26 of the Technical Specifications involves a revision of the CEA drop time. At the beginning of Cycle 3, four small flow hole test assemblies were put into the core under CEA locations in an effort to nitigate the guide tube wear problem. At that time, the CEA drop time was changed from 2.75 seconds to 3.1 seconds due to a larger dashpot effect realized with the reduced flow holes. This design is no longer being used as the "guide tube wear" fix at Millstone Unit 2 and the four test assemblie., will be removed from the core during this 1983 refueling.
The licensee, therefore, proposed changing the CEA drop time back to the original value.
C.
New Axial Shape Index Tent - The change to p. 3/42-4 involves a new axial shape index (ASI) monitoring tent for figure 3.2-2 of the Technical Specifications. This tent is used to verify the kw/ft limit of 15.6 which is input to the LOCA analyses.
Operation within the tent ensures that the maximum local power is less then 15.6 kw/f t. and thus satisfies the Technical Specification surveillance requirement.
Under nomal conditions the kw/ft surveillance limit is verified with the incore monitoring system and the only time the ASI tent is used is if the incore system is inoperable.
D.
Revised total planar peaking factor, F curve - This change affects pp. 3/4 2-6 and 3/4 2-8 of the9e,chnical Specifications and involves restoring the planar radial peaking factor, F l
tenitoring limits back to the original Beginning of Cycle 80C) 5 values. The Cycle 6 licensing analyses support this proposed revision.
E.
Revised total radial peaking factor (F ) curve - This proposed r
change affects pp. 3/4 2-8 and 3/4 2-9 of the Technical Specifications.
In comparing the BOC 5 values to BOC 6 values, the required primary flow is being reduccu by 5.4". (370,000 gpm to 350,000 gpm). Although the current licensed primary coolant flow rate is 362,600 gpm, B0C 5 values are being used since these values correspond with those of the last transient analysis.
The Cycle 4 Reload Safety Analyses have shown that the DNB
o 11 _
analysis penalty which results from a reduction of 2% in primary flow can be offset with an approximate 1% reduction in F.
Therefore, the 4% reduction in allowable F more than offsets the penalty associated with a 5.4% reduEtion in primary flow.
The Cycle 6 licensing analyses support this proposed revision.
F.
Auxiliary Feedwater Pumps - These proposed changes make Millstone Unit 2 Technical Specifications, specifically p. 3/4 7-4, consistent with NUREG-212, Revision 2 Standard Technical Spect-fications for Combustion Engineering Pressurized Water Reactors.
In addition, the proposed revision modifies the Technical Specifications to reflect the actual plant conditions applicable to Mode 4 under which there is insufficient steam to allow the steam turbine driven auxiliev feedwater pump to meet the required discharge pressur4 These changes are all acceptable because they are consistent with the Cycle 6 licensing analysis, or, in the case of the latter item, make the Millstone Unit 2 Technical Specifications consistent with the accepted specifications of NUREG-212.
8.0 Conclusions We nave reviewed Millstone Unit 2 Cycle 6 reload and the proposed chenges to the Technical Specifications and find they are acceptable.
The reload uses approved fuel types and will not cause any change in the types or in-crease in the amount of effluents or any change in the authorized power level of the facility.
The transients and accidents, and provisions for reactivity control meet applicable criteria.'
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E.1 Environmental Consideration We 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 environmental impact and, pursuant to 10 CFa 51.5(d)(4), that an eavironmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendnent.
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- 4 3.2 Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the puolic will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: December 30, 1983 Principal Contributors:
M. Bunenfeld, CPB A. Gill, CPB J. Guttman, RSB R. Barrett, RSS K. Heitner, DL l
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-1 REFERENCES 1.
W. G. Counsil (NNECO), letter to R. A. Clark (NRC), with Millstone Unit 2 Roload Safety Analysis, April 13, 1983.
2.
"Basic Safety Report," Westinghouse proprietary report for Millstone Unit 2, Docket Number 50-336, submitted via letter, W. G. Counsil (NU) to R. Reid (NRC), March 6, 1980.
3.
W. G. Counsil (NNECO), letter to R. A. Clark, November 17, 1981.
4.
L. S. Rubenstein (NRC), memorandum for T. M. Novak, "SER Input on Millstone Unit 2 BSR," February 16, 1982.
5.
W. G. Counsil (NNEC9), letter to R. A. Clark (NRC), June 22, 1983.
6.
W.' G. Counsil (NNECO), letter to J. R. Miller (NRC), November 4,1983.
7.
E. J. Mroczke (NNEC0), letter to T. E. Murley (NRC), July 1,1983.
8.
W. G. Counsil (NNECO), letter to J. R. Miller (NRC), September 15, 1983.
9.
W. G. Counsil (NNECO), letter to J. R. Miller (NRC), November 17, 1983.
- 10. E. J. Mroczka (NNECO), letter to T. E. Murley (NRC), August 12, 1983.
- 11. W. G. Counsil (NNECO), letter to J. R. Miller (NRC), December 1,1983.
- 12. R. A. Clark (NRC), letter to W. G. Counsil (NNECO), October 6,1980.
13.
W. G. Counsil (NNECO), letter to R. A. Clark, June 3,1980,
- 14. Letter f rom E. L. Conner (NRC) to W. G. Counsil (NNECO), concerning amendment 74 to the Millstone Unit 2 Facility Operatin9 License, March 5, 1982.
- 15. i.etter from W. G. Counsil (NNECO) to R. A. Clark (NRC), "Millstone Nuclear Power Station, Unit No. 2; Proposed Revisions to Technical Specifications; Large and Small Break Loss-of-Coolant Accident Evaluation." October 22, 1982.
16.
C. Thompson, V. Esposito, "Perturbation Techniques for Calculating ECCS Cooling Performance," WRAP-8986 (Februray) 1977).
- 17. Letter B10927 from W. G. Counsil (NNECO) to J.R. Miller (CRB-3),
"Millstone Nuclear Power Station, Unit No. 2 Loss of Coolant Reassessments," dated November 2, 1963.
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