ML20062C696
| ML20062C696 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/22/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20062C674 | List: |
| References | |
| NUDOCS 7811130170 | |
| Download: ML20062C696 (18) | |
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UNITED STATES NUCLEAR REGULATORY COMMISslON y waswanaTom, o.c. 2 ossa e
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i
SUPPORT!HG AENDENT NOS. 38. AND 31 TO LICENSE NOS. OPR-31 AND OPR-41 l
FLORIDA POWER AND LIGHT COMPANY t
TURKEY POINT NUCLEAR GENERATING UNIT N05. 3 AND 4 DOCKET NOS. 50-250 AND 50-251 Introduction O ., a,,iication dated aune i , i,78 and su,,iemented on ouiy i0 .
and 20, August 9 and 16 and September 13, 1978 (1, 2, 3, 4, 5, :
16)*, the Florida Power and Light Company (the licensee) requested '
amendments to Operating License Nos. OPR-31 and OPR-41 for the l Turkey Point Plant Unit Nos. 3 and 4. The application, which ,
contains accident analyses and proposed Technical Specification '
chances is in support of a recuest to modify the Technical Specifi-eatinns in ennnection with the refueling of Unit No. 4 for Cycle '
5 aperation and the operation of Unit Nos. 3 and 4 with up to -
an avarace of 25% of the tubes in the three steam generators in each unit in a plunged condition. The application also responds l to the Order for Modification of the Licenses dated June 7,1978 '
l (17). That order recuired that FPL submit a reevaluation of the ECCS cooling perfomance corrected for certain errors in the zirconium water reaction. :
In addition, the steam generator inspection report for Turkey Point s Unit No. 4 required by the Orders for Modification of License dated -
August 3 and 11, 1977 (18) and March 8, 1978 (19) has been submitted for NRC review and approval.
l Turkey Point Unit No. 4 has been reloaded for Cycle 5 operation and l is expected to be ready for restart on or about September 22, 1978. '
There are no changes in fuel or in the Technical Specifications brought about directly by this reload. However, NRC Orders (18,19) [
require a steam generator tube inspection which must be approved by the NRC before the reactor may be returned to operation. An early i estimate of the number of steam generator tubes that might require !
niugging indicated that it might be necessary to plug more than the 19% allowed by current Technical Specifications. Consequently, FPL l
- References are indicated by numbers in parenthesis and may be found at the end of this safety evaluation.
7S11210 00-7 8II I 3G ng6
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i requested pemission to plug up to 25% of the steam generator tubes !
in each unit.
The NRC requirements for approval to operate with plugged steam l
generator tubes include an ECCS reevaluation. The NRC Order of ;
June 7, 1978 (17) required that "as soon as possible, the licensee i shall submit a reevaluation of ECCS cooling perfonnance calculated in !
accordance with the Westinghouse Evaluation Model, approved by '
the staff and corrected for the errors described within.". Consequently, .
since the model had been corrected for the errors and we had I approved that correction (11) the FPL application (4) for pennission to plug up to 25% of the steam generator tubes also requested that
{' the provisions of the June 7,1978 Order (12) be deleted. [
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Our Orders for Modification of the License dated August 3 and 11,1977 !
(18) and March 8, 1978 (19) placed limitations on the operation of j Turkey Point Unit No. 4 in relation to steam generator tubes. l These limitations are being retained in the ifcense by this amend- !
ment and the Orders are thus removed. The basis for this change !
is that east experience and the review of the latest inspection ;
af the steam generators with plugged tubes has shown that the j reouf red margin of safety is being retained by the licensee.
1 Following is our evaluation for the action discussed above which ~
nrovides the basis for concluding that the Turkey Point Unit No. 4 can be safely returned to operation upon completion of the current steam generator plugging and refueling operation. .
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I I. RELOAD UNIT 4 CYCLE 5 AND l
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- 25% STEAM GENERATOR TUSE !
I PLUGGING - UNIT 3 AND 4 f
Discussion ,
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By the ap lic fan daten June 19, 1978 U) supplemented July 10,1 l July 20,$978 , August 9, and 16,1978 (16Y and September 13, 1978. )8,(2) Florida Power ;
and Light Company (the licensee) proposed to change the Technical Specifications >
for the Turkey point Units Nos. 3 and 4 in connection with the refueling of 0 Unit 4 mr Cycle 5 operation. The first reference concerns reloading Unit 4 only. It states that subsequent submittals will contain license amend-(
ment requests to allow full power operation with 25% steam generator tubes !
plugged and to incorporate ECCS model changes. The current Turkey point j 3 and 4 safety analyses are valid for steam generstar tube plugging levels ;
of up to 19%. The proposed license amendment to allow operation with- ;
25% steam generator plugging is contained in references 2 , 3 and 16. The t ECCS model changes are discussed in references 4 and 5.
The refueling consists of the replacement of 61 burned fuel assemblies by -
12 f_resh assablies and 49 previously burned assemblies. The previously l burned assemblies are: 24 assemblies last irradiated in Cycle 2 with an ;
approximate average burnup of 25,000 MWO/MTU and 25 assembifes last l
irradiated in Cycle 3 with an approximate average burnup of 27,700 MWD /MTU.
Use of a 1faited number of fresh assembif es and a large number of assemblies l with high burnup will make Cycle 5 a short cycle of approximately six i Q months duration. The licenses has elected to pursue this course of action to provide for contingencies in possiblo steam generator replacement.
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i In order to flatten the radial power distribution the licensee will place !
8 fresh borosilicate burnable poison rods in each of four centrally located i once burned fuel assembifes, and 12 depleted borosilicate burnable poison !
rods in each of 28 fuel assemblies, spaced throughout the core.
j Analyses perfonned for the Cycle 5 reload core design were based on the !
following assumptions: ,
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- 1. Cycle 4 operation is terminated after 9400-100 .."WD/MTU
- 2. Cycle 5 burnup is limited to the end of full power capability, and -
- 3. There is adherence to plant operating limitations given in the Technical Speci fications.
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i The licensee has proposed the following ' changes to the Technical Specifications for both Units 3 and 4 as a result of its analyses of operation with 25% steam generator tube plugging and the LOCA:
- 1. Add a figure defining the safety limits for 3 loop operation with between 19 and 25 percent of the steam generator tubes plugged. t
- 2. Change the overpower AT trip function constants consistent with the above safety limits.
- 3. Reduce the reactor coolant flow rate to 255,075 gpm (95% of former value).
- 4. Change the total core peaking factor, Fq to 2.03.
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- 5. s Revise the newthe axiai (large Fo )haping break factor figure LOCA analysis toexisting with the reflect afunction.
reno maii n tion cf Evaluation Fuel Mechanical Design
' The mechanical design of the fresh fuel assemblies (Region 7) is identical to the Region 6 fuel loaded in the last core reload.
Clad flattening will not occur during Cycle 5. Clad flattening time is predicted to be greater than 34,000 EFpH for all fuel regions being jgadi-ated during Cy:le 5 using the approved Westinghouse Evaluation Modalt01 Since the maximum cumulative irradiation time through Cycle 5 for the limiting region (Region 3) is expected to be approximately 25,600 EFPH, clad flattening will not occur.
Control Rod Insertion Limits There are no changes proposed to the control rod insertion limits for Cycle 5.
There are a number of criteria which the control rod insertion limits are checked against each cycle. The most important of these are shutdown margin, ',
ejected control rod worth, and F3g. The existing insertion limits remain adequate to meet the control requirements for Cycle 5.
Shutdown Margin The hot full power shutdown margin is predicted by the licensee to be 3.23% ao at BOC and 2.69% ao at EOC, compared to a shutdown margin requirement of 1.36% ao at BOL* and 1.77% ao at ECL as assumed in the steam line break
- The normal BOL requirenent is 1% shutdown margin. Because of the short Cycle 5 life, the initial baron concentration will be low enough to require a 1.36% shutdown margin.
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l analysis. This is acceptable because of extra margin between predicted and required shutdown margin throughout cycle life, In addition.' the predicted -
shutdown margin is conservative because a 10% s,alculational uncertainty is :
subtracted from the all rods inserted except for the highest worth stuck l rod calculation in determining the predicted shutdown margin. Furthermore, :
confirmation of the validity of the prediction is made during the startup '
physics test program by measuring the regulating banks, which contain about l
half of the total control rod wortb. These measured worths are compared !
with predictions for the measurement conditions made with the same model
- l used for calculating the shutdown margin. !
Reload Transient and Accident Analysis O The licensee has presented the results of Westinghouse predir vs of the core kinetics parameters for Cycle 5. These are calculated w:ca methods used and accepted for all recent reloads of Westinghouse designed reactors. The ,
Cycle 5 kinetics parameters remain within the bounds of the limits found l acceptable for previous cycles. .
l The licensee's evaluation of peaking factors for the rod out of position t l
and dropped red cluster control assembly (RCCA) incidents show that departure j from nucleate boiling ratio (DNBR) is maintained above 1.30. For the dropped ;
bank incident, the turtine runback is suff.icient to present a ONBR less than 1.30. Since the DNBR remains above 1.30, the consequences of these incidents for Cycle 5 are acceptable.
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1 The licensee evaluated the hypothetical steam line break cases with and j without a loss of offsite power. The results of this evaluation indicated i gd that a reanalysis of the hypothetical breaks inside containment without off- !
site power was required. The analysis used the same design methods and j assumptions approved fbr previously submitted accident analyses except for i the method of calculating the Doppler power coefficient. The Cycle 5 l coefficient properly accounts for the effects of reduced reactor coolant !
flow which exists for the cases with loss of offsite power. This includes ;
the effects of local density variations as a function of flow rate and i power level . The transient results show that for all hypothetical steam :
line break cases the DNS acceptance criteria are met. The conclusions of i the FSAR relative to meeting safety criteria remain valid and the results of this reanalysis are therefore acceptable.
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Transient and accident analysis of both Unit 3 and Unit 4 with steam generator !
tube plugging up to 25% are considered in the following sections:
Ranctor Coolant System Mow Rate As the level of steam generator tube plugging increases the reactor c:olant system (RCS) flow rata decreases. To quantitatively assess the effect of steam generator tube plugging or RCS loop flow, the licensee has taken measure- .
ments to obtain the loop flow rate at several levels of steam generator tube plugging. (
The data points were compared with the flow rate predictions obtained with -
the Westinghouse analytical model . The maximum deviation between the !
measured and predicted curves was used as a constant bias to reduce the predicted curve of flow rate versus percent steam generator tubes plugged. !
This curve was then further reduced by 2% to account for measurement uncertainty, i which the licensee has shown to be greater than the 2a confidence limit on l the measured flow rate. -
l' The resulting curve indica *es that at a plugging level of 25%, the flow rate will not be more than 5% below the thermal design flow rate of 89,500 gpm per loop. This flow rate, 85,025 gpm per loop, was then used in the j
evaluation of postulated transients and accidents for 25% steam generator tube plugging. The staff finds this acceptacle. f
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l Transients and Accidents i As a result of increasing the level of steam generator tube plugging to 25%
three principal factors affect the assumptions used in the analyses of postulated transients and accidents. These factors are: l
- 1. The RCS flow rate is lower than the thermal design value,
- 2. The RCS volume is less than that assumed in the reference analyses, and
- 3. The pump coastdown characteristics are more severe than those assumed in the reference analyses. l The licensee submitted an assessment of the impact.of sigam generator tube plugging up to a level of 25% on the non-LOCA incidentsu 3) for both Units 3 and 4. For each event the important parameters which were affected by ,
Cv. l the higher level of steam generator tube plugging were identified. Each L event was then evaluated to determine how the impacted parameters affected the analysis. The evaluations were based on the following assumptions:
Parameter This Analysis Re'erence Analyses Thermal design flow, 85,025 89,500 gpm/ loop S. G. tube plugging, 25 19 (O FSAR)
- Power level, Mwt (100%) 2200 2200 l
- Tavg at 100% power, 'F 574.2 574.2 l
q AT at 100% power, *F 58.9 55.9 i Steady state ONBR 1.72 1.8 (1.62 FSAR) . !
Fgg 1.55 1.55 (1.75 FSAR)
Fg maximum (non-LOCA) 2.05 2.55 1
- The analyses conservatively used 102 power (2244) and Tavg + 4' (578.2) l
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The results of the evaluation indicated that these events were Ifmiting or most sensitive to the higher steam generator tube plugging level. {
- 1. Uncontrolled Control Rod Asserrbly Withdrawal at power i
An uncontrolled control rod assembly withdrawal at power produces a i mismatch in reactor power and steam flow. The result is an increase in l
reactor coolant temperature. The increased steam generator tube plugging affects the analysis due to the reduction in RCS flew, the elevation in l outlet temperature and the increase in loop transient time. As a result, f the minimum departure from nucleate boiling ratio (DNBR) reported in the
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FSAR for fast reactivity insertion rates would be reduced by approximately !
5%. However, FSAR analysis assumed an F N4 g of 1.75 versus the current Q limit 1.55. This would result in approximately 20% additional DNBR margin. Thus there would be a net increase in the minimum DNBR reported l
fn the FSAR for fast reactivity insertions, i The overtemperature equation constants were recalculated consistent with the new Core Thermal and Hydraulic Safety Limits (T. S. Figure 2.1-lb) and compared to the FSAR valugs. The FSAR values were shown to be mere limiting due to the higher F H which was used for the original Core Thermal Limits. To offset t e effects of the RCS flow reduction, the FSAR overtemperature 4T tri i
. Specifications (page 2.3-2)p constants will besame By using these maintained in the setpoints, theTechnical reduction !
in DNBR during the transient would be approximately the same. Thus the minimum DNBR for the 25% steam generator tube plugging case is expected to be greater than the FSAR value since the initial steady state DNBR has increased, i
- 2. Loss of Reactor Coolant Flow i b" !
The most severe loss of flow transient is the simultaneous loss of f electrical power to all three reactor coolant pumps. The increase in steam generator tube plugging affects the analysis due to increased loop resistance which results in a more rapid pump coastdown. This event was reanalyzed for 25% tube plugging and the resultant minimum (
ONBR is 1.48. Thus, adequate margin exists for the loss of flow event with the higher level of steam generator tube plugging.
- 3. Chemical and Volume Control System Malfunction The analyses of boron dilution events are affected by increased steam generator tube plugging due to the reduction in RCS volume. The analysis l of boron dilution during refueling will not be affected since for this case the volume of reactor coolant in the steam generators is not considered.
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For dilution during startup and at power the reactor coolant volume in the steam generator tubes is assumed to be reduced by 25% (510 ft3) due to the increased tube plugging. Thus the total RCS volume used in the analysis is reduced from 7800 ft3 to 72903ft . This results in approximately a 7% reduction in dilution time from startup conditions. ,
The resultant 223 minutes for operator action is significantly greater than the acceptance criteria of 15 minutes.
For the dilution during power operation case the reactivity insertion rate versus bomn concentration curve has been recalculated consistent with the reduced RCS volume. The results show that the reactivity b ~
insertion rate assumed in the FSA.4 is still valid. Thus the FSAR analysis of boron dilution during power operation is acceptable and the 15 minute acceptance criteria will be met for the higher level of steam generator tube plugging.
The staff has concluded, based on the results of the evaluations and analyses perfomed by the licensee, that the effects of the postulated transients and accidents are acceptable at steam generator tube plugging levels up to 25%. '
l tCCS Analysis !
The licensee using the recentlyhas provided(4,5,1{}
modified reanalysis Westinghouse evaluation me ofi ECCS for botgn yntti}and4 o 9s'01 model was recently reviewed and approved by the staff.g 7 It includes This
{o 3 correction for the Ir-water error.
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. i presently, Turkey Point Station is operating with the interim values of total :
peaking factor of 2.02 and 1.97 for Units 3 and 4, respectiv9 1 These i values were imposed by the Order for Modification of LicensellgI after an j
error in the heat generated by Zr-water reaction had been discovered in the Westinghouse ECCS evaluation model. The order requested the licensee to i i
submit, as soon as possible, a reevaluation of the ECCS perfomance calculated in accordance with the corrected and approved evaluation model. The present l submittal ful fills this requirement. l The licensee has evaluated the ECCS perfomance for a large break LOCA using '
the modified Westinghouse evaluation model and assuming 25 percent of steam I generator tubes plugged. The analysis was performed for a double ended '
guillotine cold leg break (CECLG) with a discharge coefficient of C =0.4 '
l The licensee has shown in the previous submittaill3) that this brea0 size i corresponds to the highest values of peak cladding temperature and Zr-water !
reaction. The licensee has also demonstrated that the break size remaios '
unaffected by the amount of the steam generator tubes plugged.(14) ;
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l The input parameters assumed in the analysis are listed below:
l Core power: 102 percent of 2200 Mwt (rated power) i Peak Linear power: 102 percent of 11.53 kw/ft Peaking Factor: 2.03
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Accumulator Water Volume: 875 ft3 per accumulator.
The results of the ECCS analysis indicate a peak cladding temperature of 2173*F a maximum local Zr-water reaction of 7.68 percent and a total Ir-water reaction of less than 0.3 percent. All these values are balow the limits specified in 10 CFR 50.46, !
i The licensee did not include small break analysis since neither steam 7 generator tube plugging nor correction of the Ir-water error affect signi- l ficantly the results of this analysis.
The "18 case FAC analysis" was provided by the licensee (5) because the limiting peaking factor used in the analysis was below the value for which the excore detectors could give reliable measurements. The analysis showed that the, maximum predicted Fo that could occur for the remainder of Unit 3 Cycle 5 and for the upcoming Unit 4 Cycle 5 would never exceed the maximum
- allowable value derived from the corrected ECCS evaluation. The plant could therefore operate during Cycle 5 of both Units 3 and 4 without the ,
augmented surveillance procedures which were discussed in reference 15.
Based on the review of the submitted documents, the staff concludes that the results of the ECCS reanalysis, performed with the revised February 1978 version of the Westinghouse ECCS evaluation model corrected for Ir-water
' reaction error and including the assumption of 25 percent steam generator
{s) tubes plugged, yield the values of LOCA parameters which are conservative '
relative to the 10 CFR 50.46 criteria. The staff considers the submitted ECCS reanalysis, for operation of the plants with up to a maximum of 25 percent steam generator tubes plugged, acceptable. {
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_ Technical Soecifications The licensee has proposed changes to the Technical Specifications to pennit f operation with up to 25% steam generator tube plugging for both Units 3 and 4 Figure 2.1-lb, " Reactor Core Thermal and Hydraulic Safety Limits, 3 Loop Operati on" has been added to the Technical Specifications for operation !
t with between 19 and 25 percent of the steam generator tubes plugged. These i limits were generated by Westinghouse and are consistant with the limits for !
lower levels of plugging.
The Overtemperature aT equation is unchanged for plugging up to 25%. The conservative FSAR constants will not be change as discussed above. I
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The Overpower 47 equation constants were recalculated for 25t steam '
generator tube plugging. The values were more limiting for the reduced flow conditions and are therefore incorporated into the Technical Speci-l fications.
The minimum reactor coolant flow has been reduced to 255,075 gpm for steam generator tube plugging between 19 and 25 percent. This is consistant with the flow assumed in the transient and accident analysis. ,
The Technical Specification for the maximum allowable full power value of Fg is changed to ?.03. This is the value assumed as input for the LOCA analysis.
O The ao ==itzed axiai or shapins factor (risure 3.2-3) in the Technicai Specifications has been changed consistent with the assumptions used as input for the LOCA analysis.
The staff has reviewed the proposed changes to the Technical Specifications and finds than consistent with the analyses discussed in the preceeding sections and therefore acceptable.
Startup Tests The startup physics tests for Turkey Point Unit 4 Cycle 5 will verify nuclear design, power distribution and control rod worth predictions. This program includes low power critical boron concentration tests, temperature coefficient tests and rod worth and power distribution measurements. At higher powers, core power distribution and power coefficient tests will be performed. This program including the acceptance criteria for aach test was reviewed by the staff. The program is described in reference 5.
The results of this star',e physics test program will be submitted to the '
NRC in the form of a QVns i recort within 45 days of completion of the program. The staf M1ec.s that this program is acceptable.
Environmental Concl.jsions We have detemined that the amendments do 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 de-temination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of rivironmental impact and pursuant to 10 CFR 551.5(d)(4) that an anytronmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with issuance of these amendments.
Conclusion '
We have concluded, based on the considerations discussed above, that:
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(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 Comission's
! regulations and the issuance of these amdnements will not be inimical to the connon defense and security or to the health and safety of the i public.
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a II. STEAM GENERATOR TUBE l
INSPECTION - UNIT 4 ;
DISCUSSION By letter dated September 6, 1978 ?7), the licensee submitted the results of the steam generator tube inspection performed !
at Turkey Point Unit 4 during the August / September,1978 refueling outage, including the plugging criteria applied to the three [
steam generators. Based on these inspection results, the imple-mented plugging patterns, and previously submitted ECCS analysis,
(}') FPL concludes that the facility can be returned to full power
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operation for at least six equivalent months. !
Turkey Point Unit 4 has been operating under an August 3 and 11, '
1977 (18) and March 8, 1978 (19), NRC Orders for Modification of Facility Operating License No. OPR-41. One of the conditions !
of these Orders was that NRC approval shall be obtained prior j to res;aing power operation following the mandated inspection ,
of the steam generators. t EVALUATION Inspection Program The steam generator tube inspection performed during this shutdown included programs to assess the conditions associated with both
(]) the denting and " wastage" problems. For denting tube gauging I was done in all three steam generators in order to assess the extent and pattern of tube denting. On the hot leg side, all ,
tubes near the tube lane which were predicted to be bounded ,
by the 15% hoop strain contour were gauged. Based on previous i leaker history at Turkey Point Unit 4 and at similar units, as r well as previous gauging results, the gauging program also included. ;
wedge and patch plate regions. Additionally, when a restricted :
tube was found close to the inspection boundary, the inspection !
was expanded in that area. Gauging was also performed on cola l leg tubes in all three steam generators in conjunction with the '
U-bend inspection program conducted from the cola leg side. ;
The inspection for wastage was performed in accordance with .the [
provisions of Regulatory Guide 1.83. l l
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j 84anenle insoections of the visible tube support plates using nhotocraphs were perfomed in all three steam generators in order to assess the support plate conditions.
Results of Inspection and Corrective Action No leakina tubes were observed in any steam ?nerator during this inspection. Also, no tube leaks have occurred over the last six months of operation.
Gauging results indicated that any tube near the tube lane which restricted the 0.650" probe was within the 15% hoop strain contour.
O In additiaa. tubes restrictins the o.s4o Probe were within the 17.5% hoop strain contour boundary. In the tubelane region there were two tubes in the three steam generators that restricted I the 0.540" eddy current probe. Activity was noted in wedge areas including the cold leg wedge areas inspected. This was the first time wedge regions on the cold leg side which were inspected.
It appears that the growth of magnetite and associated denting are following the similar patterns in the hot leg wedge regions.
The growth of magnetite and tube denting on the hot leg side are consistent with experience at other units. Indicated tube ,
restrictions on the cold leg side in the tubelane region fell I within aporooriate strain contour boundaries and were adjacent tn nrevious denting. The implementation of the plugging criteria
"<cosse<f below coribined with previous plugging for various causes,
- esulted in a total of approximately 18.7% of the tubes being nlucced.
O ~ ae uiator cuide i.83 <aso ction dete-iaed that a totai 68 20 tubes had to be plucced due to wall thinning.
Pluon'ino Criteria The plugging criteria triplemented by the licensee is essentially the same as that used at other units with similarly degraded steam generator conditions. As in the previously accepted plugging criteria; e.g., as those discussed in the SER attached to the Order dated March 8, 1978 (19), FPL has perfomed preventive plugging based on the projected growth of the critical tube hoop strain contours predicted by the finite element analysis. This same approach has been used to establish the extent of preventive plugging necessary for continued operation of T.:rkey Point Unit 3 and Surry Units 1 and 2.
l The nrnaression of strain contours over the intended operating ;
-ariod is utilized to preventively olug beyond a tube which does '
i ae+ allow passage of a 0.540" orobe. The prooression of the 17.5" orain contour has been used to define the extent of preventive ,
al"onino necessary. This is identical to the criteria applied to Surry Unit 2, following the March 1978 (20), inspection and to c urry Unit 1, followino the inspection performed during the ;
April /May, 1978 (21), refueling outage. . j l
SUMARY Turkey Point Unit 4 is one of the six lead PWR facilities that t were identified to have suffered moderate to extensive tube denting and that have been under close monitoring by the NRC staff following l the Septem Mr 15, 1976 (22) tube failure occurrence at Surry Unit 2. Our Safety Evaluation Report attached to Amendment No. 27 to OPR-31 of Turkey Point Unit 3 dated August 16, 1977 (23), t evaluated the background information concerning " denting" of
, steam generator tubes which has been experienced at Surry Units l 1 and 2 and Turkey Point Units 3 and 4. This background is incor-norated by reference and remains valid. The infomation discussed !
above represents an update on the condition of the steam generators ,
at Turkey Point Unit 4. '
The steam cenerator inspection was perfomed in accordance with ,
a nrocram that is consistent with previously implemented program :
4* Turkew point Unit 4 and other units. We consider this inspection l 1s adenuate in the establishment of the condition of steam generators !
at this unit. :
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The cannino procram nerfomed at Turkey Point Unit 4 was essentially
- e same as the programs perfomed at Turkey Point Unit 3 and i turry Units 1 and 2. As in the gauging program perfomed at l
Surry Unit 2 during March, 1978 (20), and Surry Unit 1 during i April /May, 1978 (21), the 15'. tube h op strain contour was used '
to define the gauging boundary. These gauging programs have been !
developed over the course of time in consultation with the NRC i staff and have been determined to be acceptable. The inspection i of the Turkey Point Unit 4 steam generators has demonstrated l that the tube degradation which has occurred to date follows !
the pattern experienced at Turkey Point Unit 3 and Surry Units ;
I and 2. Results of this inspection also indicated that not all :
tubes within the predicted 17.5". strain boundary restricted the !
0.540" probe, which demonstrated quantitatively the conservatism ;
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in the tube plugging criteria. Furthennore, the results of
- his inspection at Turkey Point Unit 4 indicates that no a e nacted deoradation is occurring and that no new phenomena has been uncovered.
The nreventive plucaing pattern bounds those tubes which may be anticipated to attain the level of strain which could lead to stress corrosion crackino during the next period of operation and arowide reasonable assurance that an acceptable margin of safety will be maintained in accordance with Regulatory Guide 1.121.
({ The preventive plugging conducted by the licensee during the current outage justifies operation of the Turkey Point 4 steam generators for an additional six equivalent months.
We have concluded based on the considerations discussed above, that (1) Turkey Point Unit 4 may be operated for an additional six equivalent months under the restrictions delineated in the Ameneent to the license to which this SER applies; at the end of this period, Turkey Point Unit 4 is to be shut down, the steam generators are to be reprobed to determine the extent and pattern of additional tube denting and the results of this gauging program are to be submitted for our review and evaluation prior to the resumption of power operation, and (2) because the results of this inspection indicate that no unexpected degradation is occurring, no new phenomenon have been uncovered, the results were within
^he bounds of previously established criteria and that this -
change does not involve a significant increase in the probability
/~S or consecuences of accidents previously considered and does not involve a sionificant decrease in a safety margin; a significant
' =-ardt ennsideration is not involved.
enwimnmental Consideration up hava determined that the amendments do not authorize a change in effinent tvnes of total anounts nor an increase in power level and will not result in an.v sionificant environmental impact.
Havino nade this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact appraisal need not be prepared in connection with the issuance of these amenaments.
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Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do 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 these amendments will not be inimical to the common defense and security !
or to the health and safety of the public. l l
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References I i
- 2. Letter from R. E. Uhrig (FPL) to V. Stello (NRC) Serial No. L-78-230*
July 10,1978. ,
August 9,1978. ;
September 13, 1978.
- 6. R. A. Geroge, et. al ., " Revised Clad Flattening Model," WCAP-8377 '
(Proprietary) and WCAP-8381 (Non-Proprietary), July 1974. ,
- 8. Westinghouse letter NS-CE-1751 (C. Eicheldinger) to NRC (J. F. Stolz), .
(
"LOCA ECCS Analysis with Zirc/ Water Reaction Corrections," dated April 7, 1978.
- 9. Westir.ghouse letter NS-TMA-1830, " Supplementary Infonnation for WCAP-9220 " dated June 16, 1978. [
- 10. Westinghouse letter NS-TNA-1834, " Supplementary Information for WCAP-
, 9220," dated June 20, 1978.
- 11. NRC letter D. F. Ross, Jr. to D. B. Vassallo, " Safety Evaluation Report l, on Revised Westinghouse ECCS Evaluation Model," dated August 23, 1978. "
- 12. Letter from NRC (A. Schwencer) to Florida Power and Light Company (R. E. Uhrig), dated June 7,1978 transmitting the Order for Modification of Licenses dated June 7, 1978.
- 13. Florida Power and Light Company letter L-76-419 (R. E. Uhrig) to NRC (V. Stallo), dated December 9,1976, transmitting Major Reactor Coolant Systen Pipe Ruptures (Loss of Coolant Accident)."
- 14. Florida Power and Light Company letter L-77-217 (R. E. Uhrig) to NRC (G. Lear), dated July 11, 1977.
V
(V. Stello), dated April 10, 1978.
- f
- 16. Letter from R. E. Uhrig (FPL) to V. Stello, NRC, Serial No. 2-78-271 ,
dated August 16, 1978. i
[
l References
- 18. Letter from NRC (A. Schwencer) to FPL (R. E. Uhrig) dated August 3, '
1978, transmitting the Order for Modification of License No. OPR-41, dated August 2, 1977 (corrected August 11,1977). I
- 19. Letter from NRC (A. Schwencer) to FPL (R. E. Uhrig) dated March 8, 1978, transmitting the Order for Modification of License No. OPR-41 dated March 8,1978.
- 20. Order for Modification of License dated April 7,1978 (License
({} DPR-37, Docket No. 50-281).
- 21. Order for Modifications of License dated June 23, 1978 (License No. OPR-32, Docket No. 50-280) .
- 22. Letter from VEPCO (C. M. Stallings) to NRC (B. C. Rusche) dated October 19,1976 (Docket No. 50-281).
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