ML20082G248
| ML20082G248 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/07/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20082G244 | List: |
| References | |
| NUDOCS 9504130172 | |
| Download: ML20082G248 (11) | |
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t UNITED STATES 4
S NUCLEAR REGULATORY COMMISSION.
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. WASHINGTON, D.C. 20066-4001 -
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR' REGULATION-
- RELATED TO AMENDMENT NO.106.TO FACILITY OPERATING LICENSE NO.- NPF SOUTHERN NUCLEAR OPERATING COMPANY. INC.
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< JOSEPH M. FARLEY NUCLEAR PLANT.' UNIT 2-DOCKET NO. 50-364 i
1.0 INTRODUCTION
i By letter dated December 7, 1994,.as supplemented February 14"and March 20,
)
1995, Southern Nuclear Operating Company (the licensee) submitted a request for changing the Joseph M. Farley Nuclear Plant, Unit 2, (Farley Unit 2).
, Technical Specifications (TS). The requested amendment revises, in part, TS 4.4.6.2, 4.4.6.4, 4.4.6.5,. and 3.4.7.2 for the Farley Unit 2,- Cycle' 11 operation-to' permit the use of a voltage-based steam generator tube repair-criteria for defects confined within the thickness of the tube support plate.
l The February 14 and March 20,1995,' letters provided clarifying information
.that did not change the December 7, 1994, application and the proposed no, 1
.significant hazards consideration determination or expand'the scope of the i
c original Enderal Reaister notice.
1
2.0 BACKGROUND
The staff previously approved similar_. requests from the licensee to apply the voltage-based tube repair criteria at Farley Unit 2.
Implementation of the
)
voltage-based tube repair criteria for the ninth operating cycle was approved =
as documented in Amendment No. 87 to Facility Operating ~ License No. NPF-8
-issued April 1,1992; as corrected by letter dated April ~ 22,1992.
Similarly, implementation of the voltage-based tube repair criteria for the tenth operating cycle was approved by Amendment No. 94 dated October 20,.1993.
The staff concluded that the tube repair limits and leakage limits would ensure adequate structural and leakage integrity for indications' accepted for continued service' under the voltage-based repair criteria at Farley Unit 2 consistent with applicable regulatory requirements,- for the ninth and tenth' operating cycles.
This evaluation addresses comparable tube repair criteria for operating Cycle 11;-however, in this amendment, the licensee has proposed to increase the voltage limits from 1.0/3.6 volts to 2.0/5.6 volts. Voltage limits of 2.0/3.6 volts were approved for Farley Unit 1 in Amendment No.106 dated April 5, 1994.
The staff is currently developing a generic interim position on voltage-based limits for outside diameter stress corrosion cracking (00 SCC) confined to the thickness within the tube support plates..The NRC staff has published several conclusions regarding voltage-based repair criteria in draft NUREG-1477, 9504130172 950407 DR ADOCK O 4
~~
.:c.
. " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" and -in a draft generic letter titled " Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes." The latter document was published for public comment in the Federal Reaister on August 12, 1994 (59 FR 41520). However, the staff is continuing to evaluate an acceptable generic position that will take into consideration public comments on the draft generic letter cited above, domestic operating experience under the voltage-based repair criteria, and additional data which have been made available from European nuclear power plants. The NRC staff currently plans to document its final position on this matter in a generic letter.
Pending completion and issuance of the staff's final generic position on the voltage-based tube repair criteria, the staff is continuing to evaluate voltage-based repair criteria proposals on a case-specific basis.
Each of the case-specific evaluations of the voltage-based repair criteria are limited to one cycle of operation.
The licensee's current proposal is applicable to Cycle 11 operation and is similar to the licensee's previous proposals that were approved. Furthermore, the licensee's submittal is, for the most part, consistent with the draft generic letter issued for public comment on August 12, 1994, except as noted below.
3.0 PROPOSED INTERIM TUBE REPAIR CRITERIA The Joseph M. Farley Nuclear Plant, Unit 2, TS 4.4.6.2,. 4.4.6.4, 4.4.6.5, and 3.4.7.2 and Bases 3/4.4.6 and 3/4.4.7 are revised by this amendment request to specify the voltage-based tube repair criteria for ODSCC confined to within the thickness of the tube support plate. Modifications have been made to the previously approved (Cycle 9 and 10) TS pertaining to the implementation of j
the voltage-based tube repair criteria to make the currently proposed TS similar to those proposed in the draft generic letter.
The changes in the TS i
for Cycle 11 implementation of the voltage-based tube repair criteria include, in part:
a.
Specifying that tube support plate indications left in service as a result of application of the tube support plate plugging criteria shall be inspected by bobbin coil probe during the following refueling outages.
b.
Specifying that the implementation of the steam generator tube support plate plugging criteria requires a 100% bobbin coil inspection for hot-leg tube support plate intersections and cold-leg intersections down to the lowest cold-leg tube support plate with known outside diameter i
stress corrosion cracking (0DSCC) indications. The determination of the cold-leg tube support plate intersections having ODSCC indications shall be based on the performance of at least 20 percent random sampling of tubes inspected over their full length.
j c.
Changing the Cycle 10 repair limits for tube support plate intersections with indications of 00 SCC from 1.0 and 3.6 volts to the following for Cycle 11:
i
'l
. 1.
Degradation attributed to ODSCC within the bounds of the tube support plate with bobbin voltage less than or equal to 2.0 volts will be allowed to remain in service.
2.
Degradation attributed to ODSCC within the bounds of the tube support plate with bobbin voltage greater than 2.0 volts-will be
-repaired or plugged except as noted in c.3 below.
3.
Indications of potential degradation attributed to 00 SCC within the bounds of the tube support plate with a bobbin voltage greater-than 2.0 volts but less than or equal to 5.6 volts may remain in service if a rotating pancake coil inspection does not detect degradation.
Indications of ODSCC degradation with a bobbin voltage greater than 5.6 volts will be plugged or repaired.
j d.
Adding the following reporting requirements:
For implementation of the voltage-based repair criteria to tube support plate intersections, notify the NRC staff prior to returning the steam generators to service (Mode 4) should any of the following conditions 1
arise:
1.
If the estimated leakage based on the actual measured end-of-cycle voltage distribution would have exceeded the leak limit (for the postulated main steam line break utilizing licensing basis assumptions) during the previous operating cycle.
2.
If circumferential crack-like indications are detected at the tube support plate intersections.
J 3.
If the indications are identified that extend beyond the confines of the tube support plate.
4.
If the calculated conditional burst probability exceeds 1 x 10,
notify the NRC and provide an assessment of the safety significance of the occurrence.
e.
Permanently reducing the limits on primary-to-secondary leakage through all steam generators to 450 gallons per day and 150 gallons per day through any one steam generator.
In addition to the above TS changes, the licensee has also made the following commitments for implementing the voltage-based tube repair criteria:
1.
The requested actions of the draft generic letter will be followed with a few exceptions.
Exceptions to the draft generic letter include the following items:
(1) calibration of the bobbin coil, (2) use of the probe wear standard, (3) limiting new probe variability, (4) removing specimens for destructive examination and reporting of the results, and (5) the application of data exclusion criteria.
These exceptions are discussed in Sections 4.1, 4.2, and 4.3 of this evaluation.
In addition, the licensee has proposed not to include the mid-cycle
1
. equation for determining the voltage limits in the event of a forced outage not attributable to ODSCC at _the tube support plates pending issuance of the final generic letter.
2.
Calculation of the conditional probability of burst and total leak rate during a main steam line break (MSLB) will follow the methodology described in WCAP-14277, "SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections," dated January 1995.
As discussed in WCAP-14277, these methods are intended to be-in accord with the draft generic letter on voltage-based tube repair criteria.
3.
The NRC will be notified prior to restart if any indications of primary water stress corrosion cracking (PWSCC) are detected at the tube support plate elevations.
Furthermore, the data analysts will be briefed on the possibility that PWSCC can occur at tube support plate elevations.
4.
A tube pull aimed at obtaining three (3) tube support plate intersections will be performed during this outage. The tube pull will be successful if at least two intersections are successfully removed.
5.
No distribution cutoff will be applied to the voltage measurement variability distribution.
6.
All intersections where copper signals interfere with the detection of flaws will be inspected with a motorized rotating pancake coil probe.
7.
All intersections with large mixed residuals will be inspected with a rotating pancake coil probe.
8.
All bobbin flaw indications with voltages greater than 1.5 volts will be inspected with a rotating pancake coil probe.
4.0 EVALUATION 4.1 Insoection Issues The licensee intends to incorporate the inspection guidance of the draft generic letter into their inspection program with the exception of the bobbin coil calibration procedure, the implementation of limits on new probe variability, and the probe wear re-inspection requirements.
For the calibration of the bobbin coil, the licensee intends to calibrate the bobbin coil on the 4-20 percent holes rather than the 4-100 percent holes recommended in the draft generic letter.
For the limits.on new probe variability, the licensee proposes to implement such limits when probes are available and certified to meet the limits in the draft generic letter.
For the re-inspection of probes that do not meet the probe wear re-inspection requirements, the licensee proposes to use the same practices used during the last Farley Unit 1 steam generator inspections as discussed in a letter dated February 23, 1994.
The licensee has calibrated the bobbin coil on the 4-20 percent through-wall holes, since initial implementation of the voltage-based tube repair criteria iin 1992. The staff has concluded that calibrating'on the 4-20 percent through-wall holes rather than the 4-100 percent through-wall holes is acceptable based on (1) a review of the material provided in EPRI report NP-7480-L Volume 1 pertaining to assessing the.use of 20 percent and 100 percent through-wall holes and 100 percent through-wall EDM slots, and (2)- the results obtained by an independent contractor pertaining to the repeatability of voltage measurements between standards containing 20 percent through-wall holes,100 percent through-wall holes, and 100 percent electromagnetic discharge method (EDM) notches ~. These two studies showed that the 20 percent through-wall holes were more reproducible and the voltage readings obtained on these holes were more repeatable. Although deeper defects are typically the ones of most concern and the 100 percent through-wall holes are more representative of these defects, the staff has concluded that the better reproducibility of the.20 percent holes and the better measurement repeatability provided with these holes, in conjunction with the limits on new probe variability on the other holes in the standard (i.e., the 40, 60, 80, and 100 percent through-wall holes), justifies calibrating the bobbin coil on the 4-20 percent through-wall holes. Although the new probe variability requirements may not be implemented during this outage at Farley Unit 2, the staff finds the licensee's proposal to calibrate the bobbin coil on the 4-20 percent through-wall holes to be acceptable for this one cycle. This is consistent with previous practice at Farley Unit 2.
With respect to implementing the limits on new probe variability discussed in the draft generic letter, the staff has concluded that pending finalization.of.
the draft generic letter that the licensee's proposal on new probe variability is acceptable.
With respect to the use of alternate procedures (i.e., those which differ from the draft generic letter) for re-inspecting tubes that fail to meet the probe wear criterion, the staff has concluded that alternate probe wear methods may be used on a continuing basis provided an assessment is performed demonstrating that (1) they provide equivalent detection and sizing capability on a statistically significant basis when compared to the guidance in the draft generic letter and (2) they are consistent with current methods for determining the end-of-cycle (EOC) voltage distributions which are used in the tube integrity analyses.
These assessments, along with the statistical criteria for demonstrating that the techniques are equivalent, should be provided to the NRC for review and approval. With respect to this cycle specific application, however, the NRC staff has concluded that the methods which have been previously employed for reinspecting tubes when a probe fails to meet the probe wear criterion are acceptable.
As a result of the potential for the possible development of primary water stress corrosion cracking (PWSCC) flaws at dented tube support plate intersections, the licensee will brief their eddy current analysts of the potential for PWSCC to occur at these locations.
Furthermore, the licensee has agreed to notify the NRC prior to plant restart if any PWSCC indications are detected at the tube support plate elevations. The staff notes that PWSCC may be detected at tube support plate elevations.
If this occurs, an
~.
. evaluation may need to be performed to ensure that the voltage-based repair criteria is only applied to the ODSCC indications.
In summary, the. staff concludes that the inspection guidelines submitted by the licensee are acceptable since the proposed repair criteria is limited to one cycle, and the calibration, recording, and analysis requirements are consistent with the 1
methodology used in the development of the databases and supporting
'l evaluations.
J 4.2 Structural Intearity 4.2.1 Deterministic-Structural Intearity Assessment
'l The licensee's tube repair limits are based on a correlation between the burst pressure and the bobbin voltage of pulled tube and model boiler data. This correlation is similar to that used in approving the voltage limits in the licensee's previous submittals and those used in the draft generic letter.
2 The staff finds the licensee's proposed voltage limits acceptable given the j
current burst-pressure / bobbin voltage database, the licensee's growth rates, and the non-destructive examination uncertainty estimates.
To confirm the nature of the degradation occurring at the tube-support plate elevations, tubes are periodically removed from the steam generators for destructive analysis.
Tube pulls confirm that the nature of the degradation being observed at the tube support plate elevations is predominantly axially oriented 00 SCC and also provide data for assessing the reliability of the inspection methods and for supplementing existing databases (e.g., burst pressure, probability of leakage, and leak rate). The draft generic letter contains guidance that states utilities should remove.6 intersections for destructive examination every other outage.
To follow the draft generic letter guidance on tube pulls, the licensee would need to pull 6 intersections from their steam generators during this outage since their last tube pulls were in 1990.
Pending finalization of the final generic letter position on
.l tube pulls, the staff has concluded that the licensee should remove tubes for destructive examination at Farley Unit 2 during this outage. The staff has concluded that the licensee's commitment for obtaining additional pulled tube specimens with an objective of retrieving three intersections and obtaining a minimum of two intersections is acceptable.
Furthermore, the staff has concluded that the licensee's commitment to provide the metallurgical results from these pulled tube specimens within 120 days is acceptable for this cycle specific application.
4.2.2 Probabilistic Structural Intearity Assessment A probabilistic analysis for the potential for steam generator tube ruptures, given a MSLB, has been performed for the previous applications of this tube repair criteria'. The draft generic letter contains additional guidance on this analysis. The licensee intends to perform this calculation per the guidance in the draft generic letter that will most likely result in a higher conditional probability of burst than would have been obtained using the previous methodology because it includes parametric uncertainty. The result of the probabilistic analysis will be compared to a threshold value of 1x10',s per the guidance in the draft generic letter.
This threshold value will
. provide assurance that the probability of-burst is acceptable considering the assumptions'of the calculation and the results of the staff's generic risk assessment for steam generators contained in NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity." Failure to meet the threshold value indicates that 00 SCC confined to within the thickness of the tube support plate could contribute a significant fraction to the overall conditional probability of tube rupture from all forms of degradation that was assumed and evaluated as acceptable in NUREG-0844.
The licensee intends to calculate the conditional probability of burst per the guidance of the draft generic letter. The licensee referenced WCAP-14277, "SLB Leak Rate and Tube Burst Probability Analysis Methods for 00 SCC at TSP Intersections," dated January 1995, as a document containing the details of the methodology for calculating the conditional probability of burst given a MSLB. The staff finds the licensee's proposal to perform the calculation per the guidance in the draft generic letter to be acceptable for this outage-specific application. As noted above, the NRC staff expects this calculation to result in a higher probability of burst than would have been calculated previously because it includes parametric uncertainty. The staff notes that all applicable data should be included in the burst pressure database when performing this calculation except as discussed below.
4.2.3 Data Exclusion from the Burst Pressure Database During the performance of the pulled tube examinations, malfunctions in the test equipment or improper specimen preparation can occasionally occur which could result in erroneous readings. Data like this should not be included in a database because it could result in invalid results and/or conclusions. The-staff, therefore, concluded in draft NUREG-1477 that eliminating data from the burst pressure database was appropriate provided that the data could be shown to be erroneous or the result of an invalid test.
The staff provided additional guidance regarding the exclusion of data from the burst pressure database in a meeting with the industry on February 8, 1994. As a result of this guidance, the industry provided criteria (i.e., data exclusion criteria) for determining whether data may be removed from the burst pressure database in an April 22, 1994, letter from Electric Power Research Institute to NRC.
This letter was referenced and discussed in the draft generic letter and in the licensee's submittal dated February 14, 1995.
The staff concluded that the exclusion of the burst pressure data points cited in the April 22, 1994, letter, from the burst pressure database is appropriate. However, the staff is continuing to assess the appropriateness of excluding data points from the burst pressure database on a case-by-case j
basis pending further review of the generic data exclusion criteria presented i
in the April 22, 1994, letter.
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D '4. 3 Leakaae Intearity
'4.3.1 Normal Doerational Leakaae Consistent with prior amendments approving the use of the voltage-based. repair criteria at Farley Unit 2, the licensee will continue to limit the amount of operating leakage through any one steam generator to 150 gallons per day (gpd) and will limit the amount of operating leakage through all steam generators to 450 gpd. This requirement will be made permanent with this amendment.
4.3.2 Accident Leakaae The licensee has proposed a model for calculating the steam generator tube leakage from the faulted steam generator during a postulated MSLB which consists of two major components:
(1) a model predicting the probability that a given indication will leak as a function of voltage (i.e., the probability of leakage model); and (2) a model predicting leak rate as a function of voltage, given that leakage occurs (i.e., the conditional leak rate model).
The calculational methodology being proposed by the licensee for Farley, Unit 2 for determining the amount of primary-to-secondary leakage'under postulated accident conditions has previously been reviewed and approved by the NRC staff in the Amendment No. 54 Safety Evaluation Related To Operating License NPF-72, Commonwealth Edison Company, Braidwood Station, Unit 1, Docket No. STN 50-456 dated August 18, 1994. The staff finds this methodology acceptable for Farley Unit 2.
The staff notes that all applicable data should be included in the probability of leakage and conditional leak rate databases i
when performing this calculation except as discussed below. The staff notes that some minor variations in the details of the modeling may be necessary for the case where the p-value test is invalid at the 5 percent level. The staff, however,_ finds the licensee's proposal to perform the calculation using a methodology intended to follow the guidance of the draft generic letter to be acceptable.
i The licensee has calculated the allowable steam generator leak rate in the faulted steam generator as discussed in Section 5.0.
This value is intended to be consistent with maintaining the radiological consequences of a release outside containment to within a small fraction of the guideline values in 10 CFR Part 100. As a result, if the primary-to-secondary leakage during a postulated MSLB is less than this allowable limit, the steam generator tubing will maintain adequate leakage integrity under these conditions.
4.3.3 Data Exclusion from the Leakaae Databases During the performance of the pulled tube examinations, malfunctions in the test equipment or improper specimen preparation can occasionally occur which could result in erroneous readings. Data such as this should not be included in a database since it could result in invalid results and/or conclusions.
The staff, therefore, concluded in draft NUREG-1477 that eliminating data from the steam generator leakage databases (i.e., the probability of leakage and the conditional leak rate databases) was appropriate provided that the data could be shown to be erroneous or the result of an invalid test.
The staff
M y
a l !
provided additional guidance regarding the exclusion of data from the steam generator leakage databases in a meeting with the industry on February 8, 1994. As a result of this guidance, the industry provided criteria (i.e.,
data exclusion criteria) for determining whether data may be removed from the leakage databases in an April 22, 1994, letter from EPRI to the NRC. This letter was referenced and discussed in the draft generic letter and in the licensee's submittal dated February 14, 1995.
l t
The staff has concluded that the exclusion of the probability of Irakage data points cited in the April 22, 1994 letter, from the probability of leakage
. database is appropriate.
Furthermore, the staff has concluded that exclusion i
of the conditional leak rate data points cited'in the April 22, 1994, letter from the 7/8-inch conditional leak rate database, with the exception of model
\\ boiler specimen 542-4 and pulled tube specimen J1-R8C74, is appropriate.
However, pending further review of the generic data exclusion criteria presented in the April 22, 1994, letter, the staff is continuing to assess the i
appropriateness of excluding data points from the leakage databases on a case-by-case basis.
]
5.0 ASSESSMENT
OF RADIOLOGICAL CONSEQUENCES In support of the amendment request to apply a voltage-based repair limit for.
the Farley Unit 2 steam generator tube support plate intersections experiencing outside diameter stress corrosion cracking, the licensee stated that their assessment of the radiological dose consequences of a main steam line break accident was based upon an 11.4 gpm primary to secondary leak initiated by the accident. The licensee's conclusion as to the acceptability of the radigl,ogical doses also assumed an allowable activity level of dose equivalent I of 0.5 pCi/g in the primary coolant and 0.1 #Ci/g in the secondary coolant.
The staff has independently calculated the doses resulting from' a main steamline break accident using the methodology associated with SRP 15.1.5, Appendix A.
Two assessments were performed. OnewasbaseduponapgIand existing iodine spike activity level of 30 #Ci/g of dose equivalent the other was based upon an accident initiated iodine spike. 'The staff calculated doses for individuals located at the Exclusion Area Boundary (EAB) and at the Low-Population Zone (LPZ). The control room operator's thyroid dose was also calculated. The parameters which were utilized in the staff's assessment are presented in Table 1.
The staff's calculations showed that the thyroid doses for the EAB and LPZ would be less than the limits established by SRP 15.1.5, Appendix A.
The control room operator thyroid dose would be less than the limits of SRP 6.4 of NUREG-0800.
Therefore, the staff concluded that, based upon a limit of 300 rem thyroid at the EAB for the pre-existing spike case and a limit of 30 rem thyroid for the accident initiated spike case and for all control room operator dose assessments, a leak rate of 11.4 gpm is an acceptable limit for the maximum primary to secondary leakage initiated by the steam line break accident.
L --
> 6.0
SUMMARY
OF EVALUATION The licensee intends to follow the guidance of the draft generic letter on voltage-based tube repair criteria, except as noted above, for this cycle specific application. As a result, the staff concludes that adequate structural and leakage integrity can be ensured, consistent with applicable regulatory requirements, for indications to which the voltage-based repair criteria will be applied during Cycle 11 at Farley Nuclear Plant Unit 2.
The staff's approval of the proposed voltage-based repair criteria is based, in part, on the licensee being able to demonstrate that the conditional probability of burst and the primary-to-secondary leakage during a postulated MSLB will be acceptable.
7.0 STATE CONSULTATION
In accordance with the Comission's regulations, the State of Alabama official was notified of the proposed issuance of the amendment. The State official had no coments.
8.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes the surveillance requirement. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public coment on such finding (60 FR 8754). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
9.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Attachment:
Table 1 Principal Contributors: Kenneth Karwoski John Hayes Date: April 7, 1995
9 44-TABLE 1 l
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INPUT PARAMETERS FOR FARLEY EVALUATION 0F MAIN STEAMLINE BREAK ACCIDENT 1.
Primary coolant concentration of 30 pCi/g of dose equivalent 3'I.
Pre-existina" Soike Value (nCi/a)
- I
=
23.1-sal,=
8.3
'33 37.0 i
- I
=
I 5.6
=
a 133I=
20.4 2..
Volume o'f primary coolant and secondary coolant.
4 3
Primary Coolant Volume (ft )-
9146 Primary Coolant Temperature (*F) 578 3
Secondary Coolant Steam Volume (ft g) 3742 Secondary Coolant Liquid Volume (ft 2016~
i Secondary Coolant Steam Temperature (*F) *F) 518.3 Secondary Coolant Feedwater Temperature'(
437.3 I
TSlimitsforDE*Iig(theprimaryandsecondarycoolant.
3.
Primary Coolant DE concentration (sci /g) 0.5 Secondary Coolant DE I concentration (pci/g) 0.1 4.
TS value for the primary to secondary leak rate..
Primary to secondary leak rate, maximum any SG.(gpd) 150 Primary to secondary leak rate, total all SGs (gpd) 450 5.
Maximum primary / secondary leak rate to the faulted and intact SGs.
Faulted SG (gpm) 11.4 Intact SGs (gps /SG) 0.1 6.
Iodine Partition Factor Faulted SG 1.0 1
Intact SG 0.1
-Primary to Secondary Leakage 1.0 7.
Steam Released to the environment Faulted SG (lbs/2 hours).. 91,000 plus primary / secondary leakage l
Intact SGs (1bs/2 hours).. 479,000 plus primary / secondary leakage 8.
Letdown Flow Rate (gpm) 60 9.
Release Rate for 0.5 pCi/g of Dose Equivalent *I l
rdzhr
- I-4 i
I 9
'33 I 9.7
=
- I 14
=
135I 9.7
- 10. Atmospheric Dispersion Factors i
EAB (0-2 hours) 6.4 x 10
LPZ (0-8 hours) 1.0 x 10
+..
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