ML20112C109

Safety Evaluation Supporting Amend 83 to License NPF-76

ML20112C109
Person / Time
Site:	South Texas
Issue date:	05/22/1996
From:	NRC (Affiliation Not Assigned)
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ML20112C106	List: ML20112C103 ML20112C109
References
NUDOCS 9605240182
Download: ML20112C109 (13)

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'a UNITED STATES g

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NUCLEAR RE2ULATCRY COMMISSION 2

WASHINGTON, D.C. 3000H001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIQN RELATED TO AMENDMENT NO. 83 FACILITY OPERATING LICENSE NO. NPF-76 HOUSTON LIGHTING & POWER COMPANY CITY PUBLIC SERVICE BOARD OF SAN ANTONIO CENTRAL POWER AND LIGHT COMPANY CITY OF AUSTIN. TEXAS DOCKET NO. 50-498 SOUTH TEXAS PROJECT. UNIT 1

1.0 INTRODUCTION

By application dated January 22, 1996, as supplemented April 4 and May 2, 1996, Houston Lighting & Power Company, et.al., (the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-76) for the South Texas Project, Unit 1 (STP). The letter dated May 2,1996, provided clarifying information that did not change the initial proposed no significant hazards consideration determination nor did it expand the amendment request beyond its original scope.

The proposed amendment would revise Technical Specifications (TSs) 3/4.4.5 and 3/4.4.6.2 including associated Bases 3/4.4.5 and 3/4.4.6.2 to allow the implementation of steam generator (SG) voltage-based repair criteria for the tube support plate / tube intersections for STP Unit 1.

The voltage-based steam generator tube repair criteria allows axially oriented outside diameter stress corrosion cracking (ODSCC) confined within the thickness of the tube support plates (TSPs) to remain in service based on the magnitude of the bobbin coil voltage response. The allowed primary-to-secondary operational leakage from any one steam generator will be reduced from 500 gallons per day (gpd) to 150 gpd. The licensee has stated that the proposed amendment request is consistent with the guidance provided in Generic Letter (GL) 95-05,

" Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking [0DSCC]."

2.0 VOLTAGE-BASED STEAM GENERATOR TUBE REPAIR CRITERIA 2.1 Discussiqa The NRC staff documented its generic position on voltage-based limits for ODSCC affecting the SG tubes at the TSP elevations in GL 95-05 and its 9605240182 960522 DR ADOCK 0500 8

. supporting documentation.

This approach takes no credit for the TSPs in preventing and/or reducing the likelihood of a tube from bursting and/or leaking during postulated accident conditions.

In essence it assumes that the degradation affecting the SG tubes at the TSP elevation is in the tube free span.

The licensee's proposed amendment requests a permanent change to the TSs to incorporate the voltage-based repair criteria described in GL 95-05. The guidance contained in GL 95-05 specifies, in part, that: (1) the repair criteria is only applicable to predominantly axially oriented 00 SCC located within the bounds of the TSPs; (2) licensees should perform an evaluation to confirm that the SG tubes will retain adequate structural and leakage integrity until the next scheduled inspection; (3) licensees should adhere to specific inspection criteria to ensure consistency in methods between inspections; (4) tubes must be periodically removed from the SGs to verify the morphology of the degradation and provide additional data for structural and leakage integrity evaluations; (5) the operational leakage limit should be reduced; and (6) specific reporting requirements shall be followed, some of which will be incorporated into the plant TSs.

The licensee's current proposal requests a permanent amendment to the TSs and, in general, incorporates the guidance of GL 95-05.

Exceptions and alternatives to the methodology specified in GL 95-05 are described below.

2.2 Evaluation The licensee has proposed to follow the requested actions of GL 95-05 for implementing the voltage-based plugging criteria. GL 95-05, however, permits licensees to implement various alternatives to the methodology specifically stated in the GL.

For example, licensees can, subject to NRC approval, (1) choose to implement the voltage-based tube repair criteria at tube-to-flow distribution baffle plate intersections; (2) choose to implement an alternative to the probability of detection value of 0.6; (3) choose to include only a fraction (rather than all) of the bobbin indications which were not confirmed with a rotating pancake coil (RPC) probe in the determination of the beginning-of-cycle voltage distribution; (4) choose to implement an alternative to the main steam line break leak rate of 2496 liters per hour assigned to the V.C. Summer tube R28C41; (5) choose to implement an alternative to the probe wear criteria which requires all tubes since the last successful probe wear check to be reinspected with a new calibrated probe when a probe is found to be out of specification; (6) choose to use probe sizes different than the nominal probe size; and (7) choose to implement an industry alternative to the tube pull program specified in GL 95-05.

With respect to the items listed above, the licensee has elected not to implement the voltage-based tube repair criteria at the flow distribution baffle plate intersections; however, the licensee stated that if, in the future, they elect to apply the repair criteria to flow distribution baffle plate intersections, the technical bases will be submitted for NRC review and approval. Furthermore, the NRC has not approved the use of (1) alternatives

- - - - - - ~ ~ - - - - -. -

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to the probability of detection value of 0.6; (2) a generic alternative to i

include only a fraction of the bobbin indications which were not confirmed l

J with an RPC probe in the determination of the beginning-of-cycle voltage l

distribution (note that an alternative was approved for Beaver Valley Unit 1; however, generic implementation was not approved); (3) an alternative to the i

2496 liters per hour leak rate for V.C. Summer tube R28C41; (4) probe sizes different than the nominal probe size; and (5) an industry alternative to the i

j tube pull program specified in GL 95-05. As a result, the licensee has proposed to use the methodology specified in the GL. The staff finds this acceptable and notes that if alternatives are approved by the staff on a generic basis they may be used by the licensee. The NRC has generically i

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approved an alternative to the probe wear criteria specified in GL 95-05 by

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letter dated March 18, 1996, from Brian W. Sheron (NRC) to Mr. Alex Marion 1

(Nuclear Energy Institute). As a result, the licensee can implement either j

this approved alternative or the methodology specified in GL 95-05.

In addition, the staff's letter dated March 18, 1996, addressed a methodology for d

controlling new probe variability which the staff found acceptable for implementation.

i In GL 95-05, the NRC indicated that licensees should (1) submit the i

i methodology for calculating the conditional burst probability for NRC review l

and approval; (2) relate burst pressure to bobbin voltage using an empirical i

model (the currently approved model consists of determining a linear first-order equation between the burst pressure and the logarithm (base 10) of the bobbin voltage; this model may need to be changed as additional information is acquired; the alternative model is subject to NRC approval);

(3) relate probability of leakage to the bobbin voltage using an empirical model (the currently approved model consists of using a log-logistic function 4

to fit the data; this model may need to be changed as additional information j

is acquired; the alternative model is subject to NRC approval); (4) relate conditional leak rate to bobbin voltage using an empirical model (the 4

l currently approved model consists of determining a linear first-order equation i

between the logarithm (base 10) of the conditional leak rate and the logarithm i

(base 10) of the bobbin voltage; this model may need to be changed as additional information is acquired; the alternative model is subject to NRC l

approval); (5) not exclude data based on data exclusion criteria 3a, 3b, and 3c unless approved by the NRC; and (6) justify lowering the I-131 limits in the TSs to a value below 0.35 microcuries per gram (pCi/g), if applicable.

1 With respect to the items listed above, the licensee has submitted a methodology for calculating the conditional burst probability and the staff's i

review of this methodology is documented below.

In addition, the licensee has submitted their methodology for calculating the end-of-cycle voltage distribution and the total leak rate during postulated accident conditions (e.g., main steam line break). A review of these methodologies is also documented below. With respect to the burst pressure, probability of leakage, and conditional leak rate correlations, the licensee has proposed to use, based on currently available data, a linear first-order equation to relate the burst pressure and the logarithm (base 10) of the bobbin voltage, a log-logistic function to fit the probability of leakage data; and a linear 1

first-order equation to relate the logarithm (base 10) of the conditional leak i

q rate and the logarithm (base 10) of the bobbin voltage, respectively. The staff finds this acceptable for the current databases; however, the adequacy of these correlations should be assessed as additional data is acquired and, if the model(s) require changing as a result of this additional information, J

the revised model(s) should be submitted for NRC review and-approval per i

GL 95-05. With respect to the data exclusion criteria, the licensee did not exclude any data from the databases based on data exclusion criteria 3a, 3b, and 3c; and, as a result, the staff finds this acceptable. With respect to l

the I-131 limits in the TSs, the licensee did not request to lower the I-131 limits; therefore, no additional justification was necessary.

j 1

With respect to the methodologies for calculating the end-of-cycle voltage i

distribution, the probability of burst, and the total leak rate during postulated accident conditions, the staff has the following comments. The licensee has indicated that it will use the probabilistic methodology specified in WCAP-14277 for calculating the conditional probability of burst and the total leak rate during postulated accident conditions. The staff has reviewed these probabilistic methodologies which involves Monte Carlo simulations and has concluded that they are consistent with the methodology i

outlined in GL 95-05 and, therefore, are acceptable.

Similarly, the methodology for calculating the end-of-cycle voltage distribution is consistent with the methodology outlined in GL 95-05 and, therefore, is acceptable. To provide additional assurance of the adequacy of these calculational methodologies, the staff notes that it may periodically verify the results of these calculations and assess the effectiveness of the i

methodologies as indicated in GL 95-05.

GL 95-05 also recommends that licensees use updated databases (e.g., burst pressure, probability of leakage, and conditional leak rate databases) in their tube integrity evaluations (e.g., calculation of tube repair limits, conditional burst probability, and total leakage under postulated accident conditions). The industry is currently working on a generic process for updating the applicable databases. Once developed, the staff will review the adequacy of this process.

Comments have been supplied to the industry on this issue by a letter dated August 4,1995, from Brian W. Sheron to Mr. Alex Marion.

Pending completion of the development of the industry process for updating the applicable databases, the staff has reviewed the data supplied by the licensee and has found it to be acceptable. The staff notes that if the generic industry process for updating the databases is approved by the staff, this process would provide the mechanism for assuring NRC approval with the databases used by the' licensee for application of this repair criteria in future outages.

2.3 Conclusion The staff has previously evaluated the acceptability of the voltage-based tube repair criteria that the licensee is' proposing as documented in GL 95-05. As a result, based on this and the above evaluation, the staff finds the licensee's proposal acceptable.

Further technical details on the voltage-based tube repair criteria methodology are contained within GL 95-05 and its supporting technical documentation.

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. 3.0 TUBE LOCATIONS BEING EXCLUDED FROM THE ALTERNATE REPAIR CRITERIA 3.1 Discussion In accordance with GL 95-05, the alternate repair criteria (ARC) cannot be applied at TSP locations where tubes may collapse or deform following a postulated loss-of-coolant accident (LOCA) plus a safe shutdown earthquake (SSE) event.

Specifically, the tube locations adjacent to wedge supports at the upper TSPs are of primary concern due to the potential yielding of the plate and subsequent deformation of the tubes during a main steam line break (MStB). Consequently, an evaluation of the support plates in these regions has been performed. The staff has reviewed the licensee's submittal including Framatome Technologies, Inc. Topical Report BAW-10204P, " South Texas Project Tube Repair Criteria for ODSCC at Tube Support Plates." The licensee's submittal contained the analysis methodology and identification of tube locations excluded from the ARC at STP Unit 1.

3.2 Evaluation The loads on the SG tubes at TSP intersections during a LOCA are primarily due to the cumulative effect of pressure waves initiated at the pipe break locations and the compressive differential pressure across the tubes. The resultant effect of the pressure waves is to cause in-plane bending loads on the SG tubes at the upper TSPs.

In addition, the pipe break hydraulic forces cause a shaking of the reactor coolant system (RCS) as a whole, which further transmits inertial loads to the TSPs. A dynamic load factor is applied to the LOCA loads which are then probabilistically combined with the seismic loads via the square root-of-the-sum-of-the-squares (SRSS) method.

As specified in General Design Criteria (GDC) 4, dynamic effects of pipe ruptures in nuclear power plant units may be excluded from the design basis provided it is demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design of piping. Dynamic effects covered by GDC 4 include missile generation, pipe whipping, pipe break I

reaction forces, jet impingement forces, decompression waves within the ruptured pipe and dynamic pressurization in cavities and subcompartments. The NRC has concluded in References 2 and 3 that STP is in compliance with GDC 4.

As such, the staff believes that the probability of a rupture of the primary reactor coolant piping and the surge line is extremely low. Hence, the dynamic effects of postulated pipe ruptures of the large primary piping and the surge line are eliminated from the design basis at STP. The design loadings for the analysis of the upper tube support plate at STP, was based on a 12-inch diameter, schedule 140, attachment line break.

LOCA loads for STP were based on loads from a similar replacement recirculating steam generator (RSG) analytical model for another plant.

Some of the conservatisms in the analysis, according to the licensee, include the following:

The loads were based on a larger attachment line break (14-inch diameter schedule 140 versus the actual STP 12-inch diameter schedule 140).

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i A stress-strain curve based on the American Society of Mechanical Engineers (ASME) Code minimum yield and tensile strength properties for the support plate was used.

It is considered likely that the actual material property values are greater than Code minimum allowables.

The incr' ase in yield strength due to the rapidly applied load was e

neglected.

It was assumed that the tube deformation is equal to the TSP hole deformation in the finite element analysis even though a gap may exist at the intersection. Also, the TSP stiffness neglected any contribution provided by the tubing.

It was assumed that the interface between the support plate and wedges is frictionless even though the wedges were snugly installed and are securely welded to wrapper support blocks.

It was assumed that the entire LOCA pressure wave loading is acting at the top support plate only.

With regard to the item relating to the ASME Code minimum yield and tensile strength properties stated above, the staff does not consider its usage as a quantifiable conservatism and expects the licensee to normally' use the -

remaining properties in the course of performing calculations. The licensee's contention of increase in yield strength due to rapidly-applied loads is also not considered to be a quantifiable conservatism, unless supported by analytic or test data for the specific application. The staff finds the remaining conservatisms in the analysis stated above reasonable and* acceptable and the analytic model applicable to the STP replacement steam generators.

The ANSYS finite element computer code was utilized to g'eneeate an inelastic model of'the steam generator and evaluate the loadings on the tubes in the vicinity of the wedge supports. The seismic loading evaluation was previously performed and the results are contained in Reference 7.

The in-plane seismic loadings on the TSPs were determined on the basis of a time history analysis.

In determining the combined effects of seismic and LOCA loads on the TSPs, each of the lower support plates was conservatively assumed to exhibit this worst case loading as well. However, since the wedge groups are vertically aligned, the number of tubes affected is small. The tubes located in the vicinity of the wedge supports undergo the maximum deformation. Tubes that are projected to deform greater than.a certain critical magnitude under combined LOCA plus SSE loads were considered to be unacceptable..This acceptance criterion based on a critical magnitude of deformation relies on analysis and previous test data.

It is similar to the acceptance criterion developed for similar SGs for other plants and it has been reviewed and found acceptable by the staff. The calculations performed in Reference 6 have identified all tubes exceeding this acceptance criterion. A summary of the excluded tubes on this basis is also provided in Reference 1.

TSP locations and designations for identifying the elevation of the wedge group locations, as well as the circumferential locations of the wedge groups for TSPs have been identified. A submittal from another plant shows the limiting wedge

vg group exclusion region to be approximately 30 tubes for a similar steam generator with slightly higher loading conditions from a large primary pipe i

break (Reference 8). The exclusion region for STP is nearly 1.5 times as large for the limiting wedge group. The staff finds the number of excluded tubes at STP to be reasonable and acceptable.

3.3 Conclusion The staff has reviewed the licensee's analysis relative to the effects of accident loads at specific tube support plate locations where ARC cannot be applied. Based on this review, the staff finds that the licensee's analytical methodology and identification of tube locations excluded from the ARC are acceptable.

3.4 References 1.

Report BAW-10204P Rev. 02, " South Texas Report Tube Repair Criteria at Tube Support Plates for ODSCC Framatome Technology, Inc," dated January 1996.

2.

NUREG-0781, " Safety Evaluation Report Related to the Operation of South Texas Project, Units 1 & 2," Supplement No. 2, January 1987.

3.

NUREG-0781, " Safety Evaluation Report Related to the Operation of South Texas Project, Units 1 & 2," Supplement No. 4, July 1987.

4.

HL&P to USNRC Letter ST-HL-AE-3016, " Pressurizer Surge Line Thermal Stratification," March 14, 1989.

5.

USNRC to HL&P Letter, NRC Bulletin 88-11, " Pressurizer Surge Line Thermal Stratification - South Texas Project, Units 1 and 2 (TAC No. 72168),"

September 17, 1990.

6.

BWNT Document 32-1236240, " Calculation for Wedge Deformation in W-E RSG's."

7.

HL&P Document No. 120 (1) 00019-CWN, "Model E2 Steam Generator Stress Report," and addendum.

8.

NRC Letter from George F. Dick, Jr. to D. L. Farrar " Issuance of Amendments (TAC Nos. M90052 and M90053)," dated October 24, 1994; Amendment No. 66, Docket No. STN 50-454 p. 15.

4.0 ASSESSMENT

OF RADIOLOGICAL CONSE0VENCES 4.1 Discussion The licensee performed an assessment of the radiological dose consequences of a main steam line break accident in support of its amendment request to apply a voltage-based repair limit for the STP Unit I steam generator TSP intersections experiencing outside diameter stress corrosion cracking. That

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i assessment was based upon a primary to secondary leakage of 5.0 gpm initiated by a main steam line break accident and 0.42 gpa (600 gpd) allowed by TS 3.4.6.2.

The licensee conservatively 7.ssumed that the 0.42 gpm leakage was i

divided into 0.147 gpa to the faulted steam generator and 0.273 to the intact steam generators. The licensee found the radiological dose consequences acceptable, assuming allowable activity levels in the primary coolant of 60 pCi/g dose equivalent I-131 for the pre-existing spike condition and 1.0 pC1/g dose equivalent I-131 for the accident-initiated spike condition.

l 4.2 Evaluation i

The staff has independently calculated the doses resulting from a MSLB accident using the methodology in Standard Review Plan (SRP) 15.1.5, Appendix A.

The staff performed two separate assessments. The first j

assessment was based upon a pre-existing iodine spike activity level of 60 #Ci/g of dose equivalent I-131 in the primary coolant. The second assessment was based upon an accident-initiated iodine spike. Both assessments utilized dose conversion factors listed in Regulatory Guide 1.109 (1977) for the calculation of dose equivalent I-131 in the primary and secondary coolants, as required by the Unit 1 TSs.

For the accident-initiated spike assessment, the staff assumed that the accident initiated an increase in the release rate of iodine from the fuel by a factor of 500 over the release rate to maintain an activity level of 1 pCi/g of dose equivalent I-131 in the primary coolant.

For each assessment, 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 radiological doses for each of the assessments are presented in Table 2.

Previous MSLB accident analyses for STP Unit I conservatively assumed additional coolant iodine activity as a result of potential fuel failures.

However, the applicant's previous analysis showed no departure from nucleate boiling (DNB) occurring as a result of an MSLB. The staff has reviewed this analysis and agrees with the licensee's assessment. Although use of alternate plugging criteria may change the amount of allowable steam generator tube leakage during plant operation, this change will not affect the transient DNB value following a design basis MSLB. Because there is no fuel failure for this event, the staff did not consider this scenario when reviewing this license amendment request.

4.3 Conclusion The staff's calculations, as shown in Table 2, show that the thyroid doses for the EAB and LPZ are within the acceptance criteria presented in SRP 15.1.5, Appendix A of NUREG-0800 for both the pre-existing spike and the accident-initiated spike cases. The control room operator thyroid doses are also within the acceptance criteria presented in SRP 6.4 of NUREG-0800.

Since the

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-g-calculated doses meet these acceptance criteria, the staff concludes that a leak rate of S.15 gpa is an acceptable limit for the maximum primary to secondary leakage initiated by the MSLB accident.

5.0 PROPOSED CHANGE

S TO TS 3/4.4.5 AND 3/4.4.6.2 AND ASSOCIATED BASES GL 95-05 provided model TS changes based on the NUREG-0452, Revision 4a, i

" Standard Technical Specifications (STS) for Westinghouse Pressurized Water Reactors." STP Unit 1 proposed the following TS changes:

New surveillance requirement (SR) 4.4.5.2.b.4 requires future bobbin coil inspection of all tubes left in service as a result of the application of l

voltage-based repair criteria; New SR 4.4.5.2.e requires a 100 percent bobbin coil inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with known 00 SCC indications. The determination of the tube support plate intersections having ODSCC indications shall be based on the perforniance of at least a 20 percent randem sampling of tubes inspected over their full length, j

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Modify SR 4.4.5.4.a.6 to include an exception to the current plugging l

l limits so that the definition does not apply to the region of the tube i

subject to the TSP intersections since the voltage-based repair criteria applies to this region; New SR 4.4.5.4.a.11 provides limitations applicable for the TSP voltage-

=

based repair criteria limit; New SR 4.4.5.5.d addresses additional reporting criteria for those tubes where the TSP voltage-based repair criteria has been applied; Modify TS 3.4.6.2.c by changing the I gpm limit and by changing the l

500 gpd limit for leakage through any one steam generator to 150 gpd for Unit 1 only; Modify TS Bases 3/4.4.5, Steam Generators, to reflect the reduction in l

Unit I daily steam generator leakage limits from 500 gpd to 150 gpd, delete "by radiation monitors of steam generator blowdown," and to add a reference to SRs for voltage-based repair criteria; Modify TS Bases 3/4.4.6.2, Operational Leakage, to address the new Unit 1 l

steam generator leakage limits.

i The above TS and Bases changes meet the guidance provided by the staff in GL 95-05. The proposed TS and Bases changes modify the STP Unit 1 TS to reflect the use of voltage-based repair criteria for steam generator tubes affected by ODSCC. The staff has reviewed the above TS and Bases changes and finds them acceptable.

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6.0 STATE CONSULTATION

l In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendments. The State official had i

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no comments.

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7.0 ENVIRONMENTAL CONSIDERATION

1 The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 l

exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (61 FR 16651 and 61 FR 17735). 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 amendments.

8.0 CONCLUSION

The Commission 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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

J. Rajan, EMEB K. Karwoski, EMCB A. Huffert, PERB L. Brown, PERB Date: May 22, 1996

.. a TABLE 1 INPUT PARAMETERS FOR SOUTH TEXAS PROJECT UNIT 1 EVALUATION OF A MAIN STEAMLINE BREAK ACCIDENT 1.

Primary coolant concentration of 60 gi/g of dose equivalent '3'I.

Pre-existino Snike Value (nC1/c)

'3'I 45.14

=

isa!

52.67

'33*I 71.48

=

10.72 I

'33 39.50 I

=

2.

Volume of primary coolant and secondary coolant.

3 Primary Coolant Volume 13,103 ft Primary Coolant Temperature 592.0 *F l

Secondary Coolant Steam Volume 5,245 ft3 3

Secondary Coolant Liquid Volume 2,742 ft Secondary Coolant Steam Temperature 556.0 'F Secondary Coolant Feedwater Temperature 440.0 "F 3.

TS limits for DE '3'I in the primary and secondary coolant:

Primary Coolant DE '3'f, concentration (pC1/g) 1.0 Secondary Coolant DE I concentration (pC1/g) 0.1 4.

TS value for the primary to secondary leak rate:

Primary to secondary leak rate, any one SG 150 gpd Primary to secondary leak rate, total all SGs 600 gpd 5.

Maximum primary to secondary leak rate to the faulted and intact Sgs assumed for the MSLB analysis:

Faulted.SG (gpm) 5.15 Intact Sgs (gpm) 0.27 6.

Iodine Partition Factor Faulted SG 1.0 Intact SG 0.01 Primary to Seconda'ry Leakage 1.0

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TABLE 1 INPtfT PARAMETERS FOR SOUTH TEXAS PROJECT UNIT 1 EVALUATION OF A MAIN STEAMLINE BREAK ACCIDENT (continued)

I 7.

Steam Released to the Environment Faulted SG (0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 5.15 gpm Faulted SG (2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) 5.15 gpm Intact Sgs (0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 484,000 lbs plus primary to secondary leakage Intact Sgs (2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) 1,106,000 lbs plus primary to secondary leakage 8.

Release Rate for 1.0 pCi/g of Dose Equivalent 23'I Ci/hr 131 1

12.4 132 1 80.9 1

133 l

13'1 28.7 I

38.7 1

'35 1 28.0

=

9.

Letdown Flow Rate (gpm) - 100 3

l

10. Atmospheric Dispersion Factors (sec/m )

l EAB (0-2 hrs) 1.40 x 10

LPZ (0-8 hrs) 1.90 x 10 5 Control Room (0-8 hrs) 1.70 x 10'2*

11. Control Room Parameters Filter Efficiency (%)

makeup filter 95 (elemental I) recircu ation filter 95 (elemental I)

Volume (ft])

e 280,000 Makeup flow (cfm) 1800**

Recirculation Flow (cfm) 9000 1

Unfiltered Inleakage (cfm) 10 Occupancy Factor (0-8 hrs) 1 Calculat. ion based on Murphy-Campe methodology assuming a i

point source and point receptor. This estimate is conservative because it does not take into account enhanced atmospheric dispersion caused by the high release pressure and temperature of the effluent.

235 cfm of this flow does not pass through the control room recirculation filter units.

4 l

%E TABLE 2 CALCULATED THYROID DOSES FOR SOUTH TEXAS PROJECT UNIT 1 NAIN STEARLINE BREAK ACCIDENT

~ DOSE (res)'

LOCATION Pre-Existing Spike Accident-Initiated Spike EAB 4.9*

3.1**

LPZ 2.6*

6.4**

Control Room **

10.5 24.4 NUREG-0800 Acceptance Criterion =.300 rem thyroid NUREG-0800 Acceptance Cctierion = 30 rem thyroid

..