ML20059A017
| ML20059A017 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/14/1993 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19311B289 | List: |
| References | |
| NUDOCS 9312290152 | |
| Download: ML20059A017 (24) | |
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ATTACIIAIENT 1 ADOCK 05000413 $I 9312290152 931214 PDR e
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P LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l
SECTION PAGE 3/4.4.2 SAFETY VALVES t
Shutdown..
3/4 4-7 Operating...
3/4 4-3 L
3/4.4.3 PRE 55URIZER....
3/4 4 9 3/4.4.4 RELIEF VALVES......................................,3/4 4-10 3/4.4.5 STEAM GENERATOR 5..........
3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION..
3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION...............
3/4 4-18
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3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................
3/4 4-19 Operational Leakage.....................................
3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRES 5URE ISOLATION VALVES.....
3/4 4-22 3/4.4.7 CHEMISTRY.........
3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMIT 5...............
3/4 4-25 TABLE 4.4-3. REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS....................
3/4 4-26 3/4 ".0 Sr:CITIC ACTIVITY............
3/4 4 27 4RE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC instgv A recm ACTIVITY LIMIT VER5US PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCi/ gram
^ " * " " ' " "^
DOSE EQUIVALENT I-131............-.........................
3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................................
3/4 4-30 t
3/4.4.9 PRES 5URE/ TEMPERATURE LIMITS Reactor Coolant 5ystem...................................
3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO 16 EFPY.................................
3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -
APPLICABLE UP TO 16 EFPY.................................
3/4.4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -
WITH0RAWAL SCHEDULE......................................
3/4 4-35 Pressurizer..............................................
'3/4 4-36 Overpressure Protection 5ystems..........................
3/4 4-37 3/4.4.10 STRUCTURAL INTEGRITY...............................
3/4 4-39 3/4.4.11 REACTOR COOLANT SYSTEM VENT 5.............................
3/4 4-40 t
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l CATAWBA - UNITS 1 & 2 VII i
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INSERT A 3/4.4.8 SPECIFIC ACTIVITY (FOR UNIT 1) 3 / 4 A4 2 7 3/4.4.8 SPECIFIC ACTIVITY (FOR UNIT 2) 3 /4 B 4 - 27 9% c 2 c4 G 3
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SURVEILLANCE REQUIREMENTS (Continued) 3.
is verified to result in total primary to secondary leakage less than -1g0 gpm (includes operational and accident leak-age).
The basis for determining expected leak rates from the projected crack distribution is provided in --SECL-92-26h 2.
A tube can remain in service with a bobbin coil signal amplitude greater than 1.0 volt but less than 2.0 volts providedarotatingpancakecoil(RPC)inspectio'$sto,esnot detect degradation.
2-[ N y--m vw oN L)ctif- {3 Lev. h.
Indications of degradation with a flaw type' bobbin coil signal amplitude of equal to or greater than-R.;5-volts will be plugged or repaired.
CertaintubesasidentifiedinSECh-92-262,willbeexcluded from application of the Interim Plugging Criteria Limit as it has been determined that these tubes may collapse or deform following a postulated LOCA + SSE Event.
b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the repair limit and all tubes containing through-wall cracks) required by Table 4.4-2.
For Unit 1, tubes with defects below F* fall under the alternate tube plugging criteria and do not have to be plugged.
- 4. 4. 5. 5 Reports a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2; b.
The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:
1)
Number and extent of tubes inspected, CATAWBA - UNITS 1 & 2 3/4 4-16a
--Amendmente-Nor102-(-Unit --Amendment-No. 96 (Ug;t 2) e 3. c[ C
e REACTOR COOLANT SYSTEM i) 3/4.4.8 SPECIFIC ACTIVITY ( Fo Ft uurT LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity f4hg reactor coola'nt shall be limited to:
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a.
Less than or equal t niicrocurie per gram DOSE EQUIVALENT I-131',
and b.
Less than or equal to 1006 microcuries pe.' gram of gross specific activity.
MODES 1,2,3,4,and5.(UmTI)
APPLICABILITY:
ACTION:
MODES 1, 2 and 3*:
na With the specific activity of the reactor coolant greater than o.5b -l-microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less avg than 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; b.
With_the gross specific activity of the reactor coolant greater than 100/E microcuries per gram of g'ross radioactivity, be in at least HOT STANDBY with T,yg less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and c.
The provisions of Specification 3.0.4 are not applicable.
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"With T,yg greater than or equal to 500*F.
A4-27 CATAWBA - UNIT / 1,fr 3/ M
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REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY (Fog. oust 2)
LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
a.
Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/E microcuries per gram of gross specific activity.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
( umT '2.)
ACTION:
MODES 1, 2 and 3*:
a.
With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less avg than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; b.
With_the gross specific activity of the reactor coolant greater than 100/E microcuries per gram of gross radioactivity,.be in at least HOT STANDBY with T,yg less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and c.
The provisions of Specification 3.0.4 are not applicable.
4 "with T,yg greater than or equal to 500 F.
CATAWBA-UNIT [4-&2 3/44-2 -Amendment No. 25 (Unit 1)
-Amendment M.15 (Unit 2) 5 ck 0 Y
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REACTOR COOLANT SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry, ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.
Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.
The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.- The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking correc-tive actions to restore the contaminant concentrations te within the Steady-State Limits.
The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appro-priately small fraction of Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state primary-to-secondary steam generator leakage rate of 0.4 gpm.
The values for l
the limits on specific activity represent limits based upon a parametric eval-uation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the Catawba site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited
-m time periods with the reactor coolant's specific activity greater than
'h-h0-microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possib iodine spiking phenomenon which may occur following changes in THERMAL POWER.
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_T - 131 br CO b CATAWBA - UNITS 1 & 2 8 3/4 4-5 Amendiacnt t.'"? (Unit 1)
Am:ndm;nt t. "" (Unit 2)
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TECIINICAL JUSTIFICATION 1
Proposed Chances To Previous Reunest For Renewal Of The Interim Plugeing Criteria l
(IPC) For Catawba UniLI Cycle 8 Catawba is requesting modification of the Technical Specification changes submitted on October 5,1993 which requested renewal of the voltage based steam generator tube interim plugging criteria for Unit 1 Cycle 8. These changes are necessary due to a request by the NRC staff to
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include a data point in the leakage correlation that had been previously excluded from the data i
base. The inclusion of this data point results in a calculated leakage value for Cycle 8 (calculated in acconlance with the guidelines given in Draft NUREG-1477) which exceeds the projected leakage limit previously submitted (Reference I). As a result, the following changes
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and additions are needed:
1)
Increase the total primary-to-secondary leakage limit given in Surveillance Requirement 4.4.5.4.a.13)l. (page 3/4 4-16a) from 17.5 gpm to 30.0 ppm. Update the reference f
which is the basis for determining the expected leak rate fmm the projected emek distribution.
2)
For Unit I only, decrease the limit for specific activity of the reactor coolant given in Technical Specification 3.4.8.a. (page 3/4 4-27) from i microcurie per gram DOSE EQUIVALENT I-131 to 0.58 microcurie per gram DOSE EQUIVALENT I-131.
3)
In BASES 3/4.4.8 (page B 3/4 4-5), change the reactor coolant specific activity limit for Unit I given in the second paragmph from I microcurie / gram DOSE EQUIVALENT I-131 to 0.58 microcurie / gram DOSE EQUIVALENT I-131.
l Technical Justification Of Proposed Changes On October 5,1993, Catawba Nuclear Station submitted proposed changes to Technical Specifications to allow for renewal of the voltage based steam generator tube interim plugging criteria for Unit 1 Cycle 8.
This submittal contained a leakage calculation performed in acconlance with Dmft NUREG-1477 using a primary-to-secondary pressure differential of 2560 psid (Reference 1). This calculation estimated the projected primary-to-secondary leakage under steam line break conditions at EOC 8 would be 14.7 gpm based on projections for EOC 7 and the assumption that Cycle 8 would be similar to Cycle 7. In this earlier submittal, Catawba included a dose analysis calculation which showed that primary-to-secondary leakage of up to 17.5 ppm in the faulted steam generator resulted in dose consequences less than the peninent 10 CFR 100 limits (Reference 2).
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1 TECIINICAL JUSTIFICATION I
Recent discussions with the NRC staff on December 6,8 and 9,1993, called into question the exclusion of certain data points from Table.4-1 and the burst correlation of Figure 4-1 of WCAP-13854 (Reference 1).
On November 15, 1993, Catawba submitted the bases for excluding the data points in question. After review of this infonnation, the NRC staff requested j
that data point R28C41 he placed into the data base at 'a calculated leak nite together with model j
boiler specimens 598-1 and.598-3 at their measured leak rates and be used in the leakage correlation. Inclusion of data point R28C41 results in a substantial increase in the mean and variance of the leak rate which was given in the table on page 6-2 of WCAP-13854 (Reference 1). This increase in the mean and variance of the leak rate forces an increase in the total primary-to-secondary leakage when calculated in accordance with the methodology given in Draft NUREG-1477 at the 98% confidence level. When using the revised mean and variance of the leak rate, the new leakage value for Catawba Unit 1 increases to 29.35 gpm. This calculation utilizes the results of the recently completed Unit 1 EOC 7 steam genemtor tube inspection to p oject the EOC 8 voltage distribution.
If the mean and standard deviation of the table on page 6-2 of WCAP-13854 was applied to the projected EOC 8 distribution, the estimated leak rate would be 19.9 gpm at 98 % con 0dence and 17.6 gpm at 95% con 0dence. This increase above the 14.7 gpm projected for EOC 7 results from an increase in the number and voltages of indications in the NRC model for Cycle 8 compared to Cycle 7. If the leak rate versus voltage correlation of WCAP-13854 were applied rather than the NRC model, the projected EOC 8 leak mte would be estimated at 0.245 ppm j
(1.65 gpm with the additional data in the correlation).
Since this new leakage value exceeds the limit of 17.5 gpm which was previously submitted, Catawba proposes to increase the primary-to-secondary leakage limit given in Surveillance.
Requirement 4.4.5.4.a.13)l, to 30.0 gpm and decrease the specific activity of the reactor coolant
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given in Technical Specification 3.4.8.a. to 0.58 microcurie per gram DOSE EQUIVALENT 1-131 for Unit I only. These new values ensure the calculated steam line break leak rate is bounded and that the resulting dose consequences are less than the peninent 10 CFR 100 limits.
, is a letter from Westinghouse which provides the results of the leak rate calculations and corresponding analysis which utilizes data point R28C41 and the Unit 1 EOC 7 steam genemtor tube inspection data.
, is a revision of the previously submitted offsite dose calculation utilizing the increased calculated leakage. This analysis indicates that a steam line break accident with a total leakage (preexistent and growth) of 30.0 ppm in the faulted generator results in EAB and LPZ doses remaining within 10% of the 10 CFR 100 values of 25 Rem whole body and 300 Rem thyroid for the accident-initiated iodine spike, and 10 CFR 100 values for the pre-accident iodine spike.
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TECIINICAL JUSTIFICATION r
Referencrs i
1.
WCAP-13854, " Technical Suppon for Cycle 8 Steam Generator Tube Interim Plugging i
Criteria For Catawba Unit 1," September 1993.
2.
Duke Power Calculation No. CNC-1227.00-004)051, Rev. 2, "Offsite Dose From A Postulated Alain Steam Line Break," 9/29/93.
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NO SIGNIFICANT IIAZARDS CONSIDERATION In accordance with the three factor test of 10 CFR 50.92(c), implementation of the proposed changes to the license amendment is analyzed using the following standards and found not to:
- 1) involve a signincant increase in the probability or consequences of an accident pmviously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in the margin of safety.
Confonnance of the proposed amendment to the standards for a detennination of no significant l
l hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:
i 1)
Operation of Catawha Unit 1 in accordance with the proposed license amendment changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The steam line break leakage rate calculation methodology prescribed in Section 3.3 of Draft NUREG-1477 was used to calculate an EOC 8 SLB leakage value of 29.35 gpm. The NRC leakage rate calculation methodology applies a 98% confidence limit on leakage that is independent of voltage. This method for calculating SLB leakage is conservative as it assumes no correlation exists between SLB leakage and bobbin probe voltage as is shown to be the case in Reference 1. This increased leakage value necessitates a reduction in the limit on specific activity of the reactor coolant for Unit-I to 0.58 microcurie per gram DOSE EQUIVALENT I-131 to ensure that the dose consequences of a steam line break are within acceptable limits.
The dose analysis given in Attachment 5 shows that all peninent 10 CFR 100 limits are met.
Therefore, since the increase in projected steam line break primary-to-secondary leakage combined with the reduction in the limit for specific activity of the reactor coolant results in acceptable dose consequences, the proposed amendment does not result in any significant increase in the probability or consequences of an accident previously evaluated.
2)
The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The increase in projected steam line break primary-to-secondary leakage combined with the reduction in the limit for specific activity of the reactor coolant for Unit I does not introduce any significant changes to the plant design basis and will not initiate any new or different kind of accident. The reduction in the limit for specific activity coupled with the new primary-to-secondary leakage value ensures that the dose consequences will be within the peninent 10 CFR-100 limits.
Therefore, as the existing accident analyses assumptions and results will continue to be met, the proposed license amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
Page 1 of 3
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NO SIGNIFICANT IIAZARDS CONSIDERATION i
3)
TI e proposed license amendment does not involve a significant reduction in the margin of safety.
To maintain the same margin of safety as given in the previous analysis (References 2 and 3),
a dose analysis calculation was performed using a maximum primary-to-secondary leak rate that bounded the leak rate detennined in the Westinghouse calculation. This leakage was combined with a reduction in the limit for specific activity of the reactor coolant to ensure offsite doses were still within the peninent 10 CFR 100 limits.
This dose analysis of the steam line break accident with a total leakage (preexistent and growth).
of 30.0 gpm in the faulted generator and a maximum reactor coolant specific activity of 0.58 microcurie per gram DOSE EQUIVALENT I-131 results in the EAB and LPZ doses remaining within 10% of the 10 CFR 100 values of 25 Rem whole body and 300 Rem thyroid for the accident initiated iodine spike, and 10 CFR 100 values for the pre-accident iodine spike.
Based on the above, it is concluded that the proposed changes to the license amendment request do not result in a significtmt reduction in a margin with respect to plant safety as defined in the Fmal Safety Analysis Repon or any Bases of the plant Technical Specifications.
CONCLUSION Based on the preceding analysis, it is concluded that the above described changes to the license amendment request for Catawba Unit I are acceptable and the proposed changes do not involve a significant hazards consideration as defined in 10 CFR 50.92.
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NO SIGNIFICANT IIAZARDS CONSIDERATION References f
1.
WCAP-13715, "NRC Requested Catawba-1 Steam Generator Leakage Evaluation," April 1993.
t 2.
Duke Power Calculation No. CNC-1227.00-00-0051, Rev. 2, "Offsite Dose From A Postulated Main Steam Line Break," 9/29/93.
l 3.
WCAP-13854, " Technical Suppon for Cycle 8 Steam Generator Tube Interim Pluggmg Criteria For Catawba Unit 1," September 1993.
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1 ENVIRONMENTAL ISIPACT STATEMENT i
The proposed amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. As described above, the proposed amendment changes do not i
involve any significant hazards consideration, nor a signific:mt increase or change in the types l
or amounts of effluents that may be released offsite, nor a significant increase in the individual j
or cumulative occupational radiation exposures. Therefore, the proposed amendment changes meet the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.
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ATTACIIMENT 4.
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O December 13,1993 SG-93-12-006 NSD-RFK-93-013 DPC-93-306 R. Kevin Seasely Duke Power Company Catawba Nuclear Station 4800 Conord Road York, SC 29745
Subject:
Catawba 1, SG "C", Estimated SLB Irak Rate at EOC 8
References:
(1) WCAP-13854 (Proprietary), " Technical Support for Cycle 8 Steam Generator Tube Interim Plugging Criteria for Catawba Unit 1,"
Westinghouse Electric Corporation, September,1993.
(2) NUREG-1477 (Draft), " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report," United States Nuclear Regulatory Commission (NRC), June 1,1993 The purpose of this communique is to summarize recent analyses performed by.
Westinghouse to estimate potential end of cycle (EOC) 8 leak rates at Catawba 1 during a postulated steam line break (SLB) event. Based on inspection of the EOC 7 data for all SG's it was judged that the leak rate results from either SG "C" or SG "D" would bound SG's "A" and "B". The estimation of the leak rate at EOC 8 was restricted to an evaluation of steam generator (SG) "C" based on further screening analyses performed for SG's "C" and "D" utilizing the EOC 7 ODSCC distribution of indications. It was concluded from those analyses that an anal sis of SG "C" would bound the expected results from performing a similar analysis for SG "D". Using the methodologies described in References 1 and 2, the estimated EOC 8 leak rato during a postulated SLB event for SG "C" was found to be 29.35 GPM per the prescribed method of NUREG-1477 (including Plant S tube R28C41 at a calculated leak rate), and 1.65 GPM by applying a correlation between leak rate and voltage to the prescribed method of NUREG-1477. The analyses performed are described in the following paragraphs.
SLBLEAKC. WIT, 1
Dm ml.cr 13,1993
4 EOC 8 TSP Indications Distribution L
End of cycle inspection data for tube support plate (TSP) indications for all of the SG's at the Catawba 1 nuclear station were provided by Duke Power to Westinghouse for analysis. These data were compared to similar data for EOC 6 (resized at EOC 7) to determine estimated growth rates for those indications. The average bobbin amplitude for the EOC 7 TSP indications (2765) in SG "C" was found to be 0.71V with a standard deviation of 0.28V. For the resized data from the EOC 6 inspection the average and standard deviation were found to be 0.81V and 0.29V (2816 indications). Comparing all corresponding indications (2765) between the two inspections, the average growth rate was found to be -0.1V with a standard deviation of 0.27V. The negative average growth rate results from uncertainties in evaluating voltages for the small bobbin coil indications.
Using the methodology prescribed in NUREG-1477, the EOC 7 distribution was divided by a probability of detection (POD) value of 0.6 to obtain an estimate of the EOC 7 distribution ofindications accounting for indications postulated to be not detected. The distribution ofindications plugged was subtracted from the postulated EOC 7 distribution to obtain a postulated BOC 8 distribution. Monte Carlo simulation of the growth rate (the distribution is truncated to allow for no negative growth) and BOC 8 distribution was performed to obtain an EOC 8 distribution of TSP indication bobbin amplitudes. Since cycle 8 is projected to be up to 390 EFPD versus the actual 347.2 EFPD for cycle 7, the growth rate was linearly extrapolated to 390 EFPD.
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i Leak Rate Distribution H
A series of discussions with NRC personnel resulted in the direction that three (3) experimental data points be introduced into the data base for leak rate. These had previously been excluded by Westinghouse for cause, based on either crack morphology (2) or insufficient test data (1). The data were for model boiler specimens 598-1 and 598-3 and for Plant S specimen R28C41. Of these, valid measured leak rates for R28C41 were not available. It was concluded by the NRC that the results from this specimen were potentially significant and the leak rate should be estimated using the Westinghouse calculation code CRACKFLO. For this calculation the through-wall length of the crack was to be estimated based on the assumption that an interior ligament would sever under SLB pressure. The resulting estimated through-wall crack length was 0.67" with an attendant expected leak rate of 2496 liters per hour (lph), or 11 gallons per minute (gpm) at a SLB differential pressure of 2560 psi. This result and the measured results for the other two specimens were introduced into the data base.
The average and standard deviation of the leak rate for 3/4" diameter tubes were calculated. These were found to be 149.21ph and 503.91ph respectively. Previous reported results, Reference 1, were 95.81ph and 352.5 lph respectively. The effect SLitLF.AKc.Wi%
2 Demher 13,1993
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ofinclusion of the two model boiler specimens results is small relative to the
' inclusion of the R28C41 estimated leak rate. It is seen that the average increases by ~50%, and the variance is approximately doubled by this single data point.
In addition, a correlation analysis between the common logarithm of the leak rate
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and the common logarithm of the bobbin amplitude was performed. The index of determination of the correlation was found to be 39% utilizing 42 data points.
Thus, the regression curve of log-leak rate on log-volts is significant at a level of 99.9991%. This means that the probability of obtaining the correlation on a random basis is less than 0.00001. It is noted that somewhat similar results were obtained from an analysis of 7/8" tube data. Therefore, the statistical evidence that a correlation exists between leak rate and bobbin amplitude is very strong.
The inclusion of model boiler specimen 598-3 data in the probability ofleak correlation (the other data was included for Reference 1 results) was investigated with no significant effect being found.
Leak Rate Calculations Using the analysis methods described in Section 6 of reference 1, the expected leak rates at EOC 8 were calculated at a 98% and 95% level of confidence using the NUREG-1477 formulae, and at a 95% confidence level using the NUREG-1477 formulae modified to account for the correlation ofleak rate to bobbin amplitude.
The results were 29.35 gpm,26.0 gpm, and 1.6 gpm respectively. Without l
modification of the data base, the corresponding leak rate results would be 19.9, 17.6, and 0.245 gpm. It is apparent that the inclusion of the R28C41 calculated f
data point significantly affects all of the results. In addition, the disregard of the correlation between leak rate and bobbin amplitude very strongly influences the results.
Sincerely,
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R. F. Keating, Advisory Engine r Westinghouse, Nuclear Services Division i
SLBLEAKC.WI%
3 December is,1993 j
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i XITACIDfENT 5 l
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F SWORN DECLARATION I, D. L. Rehn, being duly sworn, state as follows:
1.
I hold the position of Site Vice-President, Catawba Nuclear Station, for Duke Power Company. I am submitting this declaration in suppon of Duke Power Company's request that the Nuclear Regulatory Commission (NRC) withhold from public disclosure l (Calculation CNC-1227.00-004)051, Offsite Dose From a Postulated Main Steam Line Break, Rev. 3) of the supplement to Technical Specification amendment' being submitted to the NRC by Duke Power on this date, December 14, 1993, j
concerning the offsite dose analysis associated with a postulated main steam line break -
i at the Catawba Nuclear Station. I have bea: specifiu:!1y delegated the function of reviewing the infonnation sought to be withheld and am authorized to apply for its withholding on behalf of the Company.
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- 2. of Duke Power Company's December 14, 1993 Catawba supplement to Technical Specification amendment should be protected from public disclosure because it contains tmde secrets and conGdential commercial infonnation concerning accident assumptions regarding defective fuel behavior and the resulting reactor coolant system activity and iodine spiking during reactor transient conditions. Attachment 5 contains proprietary models, equations, calculations, and discussions which reGect the achievements of Duke engineers in developing an' alternative methodology for demonstmting that offsite dose analyses associated with a postulated main steam line break are within acceptable limits.
4 3.
I have knowledge of and can attest to the criteria used by Duke Power Company in designating infonnation as proprietary or conGdential.
- 4. to Duke Power Company's supplement to Technical Specification amendment contains the following proprietary infonnation:
(a) mathematical models and methodologies (and associated discussions) developed by Duke Power Company; (b) technical infonnation obtained from IIalden Boiling Water Reactor Project participants under specific written agreement of nondisclosure:
i (c) mathematical models and technical infonnation purchased from Combustion Engineering under a written agreement of nondisclosure.
t This infonnation constitutes trade secrets and confidential commercial infonnation of Duke Power Cornpany.
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SWORN DECLARATION 5.
Proprietary infonnation in this submittal, has been held in confidence by the Company, is of a type customarily held in confidence by the Company, and is being fonvarded to the NRC in confidence. To the best of Duke Power Company's knowledge and belief, this infonnation is not available from public sources.
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6.
Disclosure of the infonnation contained in Attachment 5 of this supplement to Technical d
Specification amendment could have a detrimental impact on the operations of Duke Power Company and would cause substantial hann to the competitive position of the Company, its engineering services afGliates, and the other owners of the infonnation.
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(a)
Disclosure of mathematical models, calculations, and methodologies developed by Duke Power Company would cause substantial hann to the competitive position of the Company and its engineering services af61iates. This infonnation was developed at considerable expense to the Company, in tenns of both money and technical resources, Public disclosure of this infonnation would preclude j
Duke Power, or its engineering services affiliate, Duke Engineering and Services, Inc., from recovering the costs of developing this infonnation by marketing it to other interested entities and would negatively impact the competitive position of the Company by not requiring others to expend resources to obtain the same infonnation.
(b) of the amendment change request also references and incorporates proprietary infonnation obtained from a Halden Reactor Project publication. The Halden Reactor Project is an experimental reactor and research operation sponsored through a consonium of international panicipants. This infonnation was provided to Duke Power Company in accordance with an explicit written notice that it should be neither disclosed nor reproduced, in whole or in pan, except with the written pennission of a Project member organization. While the i
company has obtained pennission from the Electric Power Research Institute, a Halden Project participant, to reference infonnation contained in the Project publication, that information remains proprietary and should not be disclosed to the public. Public disclosure would harm the competitive position of Halden l
Project panicipants and others who have a Hnancial interest in the proprietary infonnation in question. Disclosure would also damage the Company's reputation with the provider of this infonnation, and could thereby negatively affect the Company's operations should suppliers of such imponant technical infonnation be less willing to do business with the Company in the future.
l (c)
Third, Attachment 5 of the amendment supplement incomorates proprietary infonnation purchased by Duke Power Company from Combustion Engineering, j
again under an explicit agreement of nondisclosure. Making this infonnation available to the public at no cost would unfairly hann the competitive position of 5
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SWORN DECLARATION the Company which was required to purchase the infonnation, and would hann the competitive position of the original owner of the infonnation.' In addition, public disclosure would damage the Company's reputation with the provider of this infonnation, and could thereby negatively impact the Company's operations
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should suppliers ofimportant technical information become less willing to provide such infonnation to the Company.
7.
For the above reasons, Duke Power Company is requesting that Attachmeat 5.of the supplement to Technical Specification amendment dated December 14, 1993, for the I
Catawba Nuclear Station, be withheld from public disclosure in accordance with 10 CFR 2.790 (a)(4) and (b).
I declare under penalty of perjury that the foregoing is true and correct. Executed this fourteenth day of December,1993.
D. L. Rehn Site Vice-President Catawba Nuclear Station i
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