L-2005-004, Supplemental Information Alternate Source Term License Amendment

Supplemental Information Alternate Source Term License Amendment

ML050120140
Person / Time
Site:	Saint Lucie
Issue date:	01/07/2005
From:	Jefferson W Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2005-004
Download: ML050120140 (18)

Text

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 FPL January 7, 2005 L-2005-004 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 RE:

St. Lucie Unit 2 Docket No. 50-389 Supplemental Information Alternate Source Term (AST) License Amendment On September 18, 2003, Florida Power and Light Company (FPL) submitted the St. Lucie Unit 2 Alternate Source Term (AST) license amendment request via FPL letter L-2003-220. As a result of the FPL submittals and a teleconference with the NRC staff on June 28, 2004, the NRC requested additional information to assist in their review of the proposed amendment via NRC letter dated July 9, 2004. An additional NRC question was provided to FPL via an NRC (Arroyo) to FPL (Madden) e-mail dated August 2, 2004. FPL letter L-2004-203 dated September 21, 2004, provided the additional information requested. FPL letter L-2004-206 dated September 24, 2004 supplemented the Unit 2 AST submittal to provide additional technical specification requirements for the Unit 2 emergency core cooling system (ECCS) area ventilation system charcoal filters.

The original intent was to have AST fully approved for St. Lucie Unit 2 prior to restart from the SL2-15 refueling outage; however, as the NRC review process progressed it became apparent that complete approval of the submittal would not be possible prior to the planned unit restart. During a December 17, 2004 teleconference between St. Lucie Plant management and NRC management, it was decided to focus the NRC AST review on those design basis events necessary to support restart, with NRC review of the other design basis events to follow at a later time. To this end, the NRC provided FPL with a list of questions via an NRC (Moroney) to FPL (Madden) e-mail dated January 5, 2005. This supplemental response formally documents the FPL response to that NRC request.

As part of the selective implementation of the AST, the reactor coolant system (RCS) operational leakage Technical Specification (TS) change (TS 3.4.6.2) would need to be issued to support the 30% steam generator tube plugging amendment.

New regulatory commitments included in this supplement include:

In Attachment I response to RAI Number 2, FPL commits to revising plant procedures to ensure that, in the event of a plant accident involving a secondary release, the steam generators are isolated once shutdown cooling is placed in service.

The revised procedure(s) will be in place at the time of implementation of this amendment request.

In Attachment 1 response to RAI Number 5, FPL commits to revise the Chemistry procedures discussed in the response prior to startup from SL2-15 currently scheduled for late January 2005.

A1 an FPL Group company

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Page 2 The original no significant hazards analysis submitted by FPL letter L-2003-220 remains bounding. In accordance with 10 CFR 50.91(b)(1), a copy is being forwarded to the State Designee for the State of Florida.

Please contact George Madden at 772-467-7155 if there are any questions about this submittal.

vic Pd ntt St. Lucie Plant WJ/GRM Attachments cc:

Mr. W. A. Passetti, Florida Department of Health a

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Page 3 STATE OF FLORIDA

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)ss.

COUNTY OF ST. LUCIE

)

William Jefferson, Jr. being first duly sworn, deposes and says:

That he is Vice President, St. Lucie Plant, of Florida Power and Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, that h thorized to execute the document on behalf of said Licensee.

Willian rrJfferj!

F STATE OF FLORIDA COUNTY OF ST. LUCIE Subscribed and sworn to before me this 7 day of 3@8!2005.

Name Not R

Pudbc so Ed~re Jun 17 200a ame ta Pblic(Typ or rint)

William Jefferson, Jr. is personally known to me.

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 1 Page 1 St. Lucie Unit 2 Alternate Source Term Supplemental Response to the NRC Request for Additional Information Related to the SGTR Event This attachment documents the FPL response to the NRC third request for additional information (RAI) transmitted via an NRC January 5, 2005 e-mail. The question numbers identified below refer to the items as numbered in the NRC e-mail.

NRC Question 1 - Provide a basis for the determination that the Steam Generator Tube Rupture (SGTR) event is the only AST analysis affected by the 30% tube plugging amendment request. Include a technical justification for the fact that the dose consequences for non-SGTR events remain applicable with 30% SGTP. Also, identify the applicable technical specification (TS) changes necessary to support approval of AST for SGTR.

FPL Response - Attachment 2 of this letter provides the requested technical justification that demonstrates, for events other than SGTR, the dose consequences associated with 30% SGTP are bounded by the existing analysis of record (AOR).

The proposed change to TS 3.4.6.2, Operational Leakage, is the only TS change required to support the SGTR event. This proposed change reduces the allowed amount of primary-to-secondary leakage and was included on Page 3 of Attachment 3 to the FPL amendment request provided via letter L-2003-220 dated September 18, 2003.

NRC Question 2 - Section 5.3 of RG 1.183, states:

'The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature of the leakage is less than 1000C (212'F). The release of radioactivity from unaffected steam generators [SGs]

should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated."

In response to RAI question 6 (L-2004-203), FPL stated that uAlthough capable of steaming until a RCS temperature of 212'F, under such condition steaming of the unaffected SG (while on shutdown cooling) would be restricted by coordination between the Shift Manager and the Emergency Coordinator using Emergency Operating Procedures and Emergency Plan Implementing Procedure guidance, both of which focus on minimizing offsite releases."

The NRC staff does not find that this is consistent with the intent of the guidance provided in RG 1.183, and does not provide a conservative assessment of the radiological consequences of accidents that use the assumption that steaming stops when shutdown cooling is established.

The NRC staff requests that additional justification be provided for this apparent non-conservative model, or that the analyses include the impact of modeling the release until the steaming stops with the reactor coolant system (RCS) at temperatures less than 212'F.

FPL Response - This question is an extension of NRC Question 6 from the original RAI. The original intent of the question was to verify that procedures exist to ensure that the secondary release is terminated upon initiation of shutdown cooling at 300OF (AST analysis assumes release is terminated upon initiation of shutdown cooling). FPL's initial response to the question

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 1 Page 2 was unacceptable, and FPL had proposed pursuing an analytical approach related to the amount of steam released, as noted in the above question. Rather than pursue that option, FPL commits to revising plant procedures to ensure that, in the event of a plant accident involving a secondary release, the steam generators are isolated once shutdown cooling is placed in service.

The revised procedure(s) will be in place at the time of implementation of this amendment request.

NRC Question 3-RG 1.183, Section 5.1.2, states:

'Assumptions regarding the occurrence and timing of a loss of offsite power [LOOP] should be selected with the objective of maximizing the postulated radiological consequences."

For accidents that have taken credit for the condenser before the reactor trip (such as the SG Tube Rupture Accident) it is not clear that crediting the condenser until the time of the reactor trip maximizes the postulated radiological consequences. Please justify the condenser being available given a LOOP is assumed to occur at the time of the accident.

The NRC staff notes that consideration for this concurrent LOOP appears to be part of the St.

Lucie Unit 2 design basis (Updated Final Safety Analysis Report (UFSAR), Section 15.6.1.7.1, Steam Generator Tube Rupture (SGTR) with a Concurrent Loss of Offsite Power (Reload Cycles)).

Also, UFSAR Section 15.0.1.9.1, Question 440.9 indicates that a stuck open atmospheric dump valve (ADV) is currently part of the St. Lucie Unit 2 licensing basis. It appears that this was considered for other design basis accidents. In the case of the SGTR, the single failure of the stuck open ADV was considered to be bounded by the letdown line break outside of containment. Likewise the ADV being stuck open appears to be explicitly considered for the Single Reactor Coolant Pump Shaft Seizure/Sheared Shaft analysis presented in UFSAR Section 15.3.5.1.7.1.

The response to RAI question 9 stated that "the ADVs are not allowed in automatic mode," so it was not considered for this accident. The NRC staff believes that the accident considerations of the design basis accidents analyzed are not limited to only full power operation. Routinely accidents are analyzed for both full power and zero power modes. RG 1.183, Section 5.1.2,

'Credit for Engineering Safeguard Features," states: 'The single active component failure that results in the most limiting radiological consequences should be assumed."

Please provide an analysis of the impact of the ADV being stuck open or other limiting single failure with respect to radiological consequences for the accidents impacted by the 30 percent SGTP amendment.

a. Provide a timeline for the SGTR event comparing the case in which the accident and LOOP are concurrent with the case in which the LOOP occurs at the time of the reactor trip.

b. Provide verification that the analysis for the SGTR includes the worst-case single failure.

FPL Response - The following is a brief timeline describing the steam releases, atmospheric dispersion factors and control room configuration for the currently analyzed SGTR event with a LOOP and reactor trip at 379.2 seconds and for an SGTR event with LOOP and reactor trip at 0 seconds. Important points to note are:

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 1 Page 3

1. Delaying the LOOP and associated reactor trip allows for an additional contaminated steam release of over 1,250,000 Ibm up to 1800 seconds (30 minutes).

2. The analysis assumes no control room isolation until 30 seconds after the LOOP, so delaying the LOOP delays the isolation of the control room.

3. Once operators take control of the steam release via the secondary side (e.g., manual use of the ADVs) at 1800 seconds (30 minutes) for both cases, the steam releases are identical.

Current 30% SGTP SGTR Analysis Timeline

• SGTR occurs at 0 seconds LOOP with reactor trip occurs at 379.2 seconds

• Control room isolation occurs at 409.2 seconds (assumed coincident with LOOP with 30 second delay for isolation)

From 0 seconds to time of reactor trip, there is full power steam flow to the condenser Full power steam flow to the condenser ceases due to reactor trip From 379.2 to 1800 seconds, steam is released via the MSSVs. This steam is produced by the sensible heat stored on the primary side and the decay heat produced by the reactor core

• At 1800 seconds, operators take control of steam releases from the secondary side Steam Releases 0 - 379.2 seconds 661,842 Ibm from the condenser (ruptured SG)

X/Q = 2.47E-3 656,568 Ibm from the condenser (unaffected SG)

X/Q = 2.47E-3 0 Ibm from the MSSVs 379.2 - 1800 seconds 0 Ibm from the condenser 85,089 Ibm from MSSVs (ruptured SG)

X/Q = 6.69E-3 from 379.2 - 409.2 seconds X/Q = 3.11 E-3 from 409.2 - 1800 seconds 83,989 Ibm from MSSVs (unaffected SG)

X/Q = 6.69E-3 from 379.2 - 409.2 seconds X/Q = 3.11E-3 from 409.2-1800 seconds NOTE: The current analysis applies a combination of atmospheric dispersion factors (XIQs) that utilize the higher MSSV-to-control room X/Q (6.69E-3 versus the 2.47E-3 condenser release to the control room) beginning at 329.4 seconds during the high condenser steam release period and assume control room isolation at 360 seconds. These X/Q applicability times are based on the original SGTR timing (versus the 30% SGTP timing) and produce an insignificant change in the dose results.

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 1 Page 4 Alternate SGTR Analysis Timeline

• SGTR occurs at 0 seconds LOOP at 0 seconds with reactor trip within 2 to 3 seconds Control room isolation occurs at 30 seconds (assumed coincident with LOOP with 30 second delay for isolation)

Since reactor trip occurs at 2 to 3 seconds, there is no significant steam flow to the condenser From 0 to 1800 seconds, steam is released via MSSVs. This steam is produced by the decay heat and sensible heat stored on the primary side At 1800 seconds, operators take control of steam releases from the secondary side Steam Releases 0 - 1800 seconds Less than 6,000 Ibm from the condenser (ruptured SG)

X/Q = 2.47E-3 Less than 6,000 Ibm from the condenser (unaffected SG)

X/Q = 2.47E-3 Less than 108,000 Ibm from MSSVs (ruptured SG)

X/Q = 6.69E-3 up to 30 seconds X/Q = 3.11 E-3 from 30 - 1800 seconds Less than 107,000 Ibm from MSSVs (unaffected SG)

X/Q = 6.69E-3 up to 30 seconds X/Q = 3.11E-3 from 30- 1800 seconds The St. Lucie AST analyses (all events, both units) were developed using conservative and bounding assumptions for key analytical parameters. In addition, the assumed plant response to these analyzed events was in accordance with the current plants licensing basis, which includes consideration of limiting single failures and a loss of offsite power. This was done in order to maximize the postulated radiological consequences.

For the radiological consequences of a SGTR event, the limiting single active failure assumed in the AST analysis is a failure that results in the loss of one train of the control room emergency air cleanup system. This assumed failure serves to maximize dose to control room operators.

NRC Question 4 - The proposed AST amendment (FPL letter L-2003-220 dated September 18, 2003) includes a change in the TS definition of Dose Equivalent Iodine (DEI).

Provide justification as to why this TS change is not required for the SGTR analysis.

FPL Response - The current St. Lucie Unit 2 Technical Specification definition for dose equivalent lodine-131 refers to ICRP-30 for thyroid dose conversion factors. The proposed Technical Specification definition for the St. Lucie Unit 2 Alternative Source Term submittal refers to Federal Guidance Report No. 11 (FGR-11) for the thyroid dose conversion factors.

FPL's proposal for selective implementation of AST for the SGTR event does not include the above Technical Specification change to the definition of dose equivalent 1-131. For the SGTR event, the radionuclides of interest are the radioactive iodines. For these isotopes, the thyroid dose conversion factors provided by ICRP-30 and FGR-1 1 are equivalent. Thus, this Technical

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 1 Page 5 Specification change is not necessary for the proposed selective implementation.

This equivalency has been previously accepted by the NRC via approval of Amendment 107 to the Shearon Harris Nuclear Power Plant Technical Specifications (documented in a letter dated October 12, 2001 from N. Kalyanam to James Scarola, more specifically in Enclosure 1 and on page 22 of Enclosure 2 to the letter).

NRC Question 5 - Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," provides assumptions acceptable to the NRC staff for evaluating the radiological consequences of a design basis accident at pressurized-water reactors and states the following:

"The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,

Ibm/hr) should be consistent with the basis of surveillance tests used to show compliance with leak rate Technical Specifications [TS]. These tests are typically based on cool liquid. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids.

In most cases, the density should be assumed to be 1.0 gm/cc (62.4 Ibm/cubic feet)."

The value for the reactor coolant system (RCS) density provided by FPL in Attachment 5, page 3 of 4 of letter L-2004-203, is 45.2 lbs/cubic feet. It is stated that "the RCS volume is also converted to mass using the same density."

Provide the density values used for the event analyses that model primary-to-secondary leakage. Confirm that the plant surveillance tests used to show compliance with leak rate TS are consistent with the density assumed. For example, if these tests are based upon cool liquids, the density used should be 1.0 gm/cc.

Describe any other procedures in place to limit primary-to-secondary leakage to less than the TS limit.

FPL Response - As a result of FPL and industry operating experience on this subject, St. Lucie is in the process of revising its primary to secondary leak rate monitoring to include compensation for RCS density differences between the cold monitoring condition and the hot conditions assumed in the analyses. As such, the plant's accident analyses and primary-to-secondary leakage monitoring will be made consistent.

Although identified during the development of AST, this density discrepancy issue represents a current plant condition adverse to quality; it is not the result of AST. This condition and the required corrective actions are documented in the plant's corrective action program via condition report (CR).

In addition, plant procedures provide conservative restrictions on plant operations with identified tube leakage.

Upon indication of a tube leak, off normal operating procedure (ONOP) 2-0830030, Steam Generator Tube Leak, is entered. In accordance with this procedure, a plant shutdown is required for the following conditions:

• if total tube leakage is greater than or equal to 150 gpd (about 0.1 gpm),

• if total tube leakage is greater than or equal to 75 gpd (about 0.05 gpm) and increasing at a rate of greater than or equal to 30 gpd/hr, or

• if total tube leakage is greater than or equal to 75 gpd for greater than or equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 1 Page 6 These procedural limits are much more restrictive than both the current and proposed technical specification limits. The current technical specification limits require a plant shutdown when tube leakage exceeds 1 gpm total leakage, or 720 gpd (0.5 gpm) per any one steam generator.

The proposed technical specification limits are 0.3 gpm total or 216 gpd (0.15 gpm) per any one steam generator.

The CR disposition requires a revision to Chemistry procedures to provide a density correction factor of 1.4 to their current leak rate calculation results; however, this action is not immediately required since there is sufficient margin between the conservative limits of procedure ONOP 2-0830030 and the TS limits to ensure the TS limit is not violated.

In order to exceed the current TS allowed limit of 1.0 gpm total and 0.5 gpm per any one steam generator, the uncorrected measured leakage would have to exceed 0.7 gpm and 0.35 gpm, respectively. As described above, plant procedures require a plant shutdown if leakage reaches 0.1 gpm.

Thus, plant procedures are sufficiently conservative to ensure that the assumed primary-to-secondary leak rate will not be exceeded.

Upon NRC approval of alternate source term (AST), the TS steam generator tube leakage limits will be reduced to 0.3 gpm total and 0.15 gpm per any one steam generator. Even with these reduced limits, plant procedures remain sufficiently conservative to ensure the TS limit is not exceeded since the uncorrected measured leakage would have to exceed 0.2 gpm and 0.1 gpm, respectively.

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 1 Evaluation of Radiological Dose Evaluation for Safety Analyses Based on the WCAP-9272 Method with 30% Steam Generator Tubes Plugged

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 2 Evaluation of Radiological Dose Evaluation for Safety Analyses Based on the WCAP-9272 Method with 30% Steam Generator Tubes Plugged Introduction Recently, Florida Power and Light Company (FPL) has completed a set of non-LOCA safety analyses which support operation with up to 30% of the steam generator tubes plugged. These analyses also transitioned from the CESEC based Combustion Engineering analytical processes to the RETRAN based WCAP-9272 analytical process. This report presents the results of an evaluation of impact of these two changes on the radiological consequences calculation, other than the steam generator tube rupture event.

Method of Evaluation The method of this evaluation is to compare key inputs to the radiological consequences calculation from the recent set of analyses to those used in the existing radiological analysis of record. Changes to the inputs are examined and either 1) changes which tend to be in the favorable direction are sufficiently greater than those in the adverse direction to demonstrate the continued applicability of the radiological consequences analysis of record or 2) the margin between the existing analysis of record results and the regulatory limit are judged to be greater than the impact of the adverse consequences. For non-LOCA fuel failure events, the dose consequences are dominated by the amount of fuel failures and the primary-to-secondary leakage.

1.1. Parameters Unaffected By Tube Plugging The majority of the parameters which are part of a radiological consequences calculation are governed by regulation, technical specifications or agreed upon analytical assumptions in licensing interactions.

Table 1 lists these key parameters.

These parameters were not impacted by either the increase in steam generator tube plugging nor the transition to WCAP-9272 based reload methodology.

Table I Parameters Unaffected by Tube Plugging Parameter Atmospheric Dispersion Factor, 0-2 Hour Exclusion Area Boundary Atmospheric Dispersion Factor, 0-8 Hour Low Population Zone Iodine Spiking Factor Iodine Partitioning at Steam-Liquid Interface Breathing Rate Dose Conversion Factors Technical Specification Primary Iodine Activity Technical Specification Secondary Iodine Activity Technical Specification Noble Gas Activity

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 3 1.2. Parameters Affected By Tube Plugginq The changes listed in this section are changes, which might impact a radiological consequences calculation based upon 30% of the steam generator tubes having been plugged.

The radiological consequences calculation is performed separately from the transient analyses.

Thus, the discussion of steam generator inventory in this section deals with the impact of having a different steam generator mass used in radiological consequences calculation. Any impact of having a different steam generator mass that is discharged through a break is dealt with in the specific event section later in this report.

The most important impact of the increase in steam generator tube plugging is the reduction in heat transfer area in the steam generator. The reduction results in a decrease in secondary system pressure for any given primary system temperature and power.

The heat of vaporization of water varies with pressure. Table 2 demonstrates that for any steam generator pressure, a lower pressure resulting from the plugged tubes will result in a greater heat of vaporization. Therefore, slightly lowered steaming would be necessary to remove core decay heat and stored energy in the RCS metal and liquid mass during the cooldown.

Another consideration due to the pressure difference is the mass of secondary system liquid.

The iodine activity leaking from the primary-to-secondary mixes with the mass in the steam generator. Thus, a greater steam generator mass would lead to a more dilute concentration for any given amount of secondary system leakage.

Table 3 demonstrates that the lower saturation pressure resulting from the plugging of 30% of the steam generator tubes results in an increase in the density of the liquid in the steam generator.

From Tables 2 and 3 it is seen that the combined effect, although small, is beneficial for 30%

steam generator tube plugging (SGTP).

Table 2 Reduction in Steam Flow Due to 30% Tube Plugging Current Co nfiguration 30% Plugged SG Tubes SG Pressure, SG Pressure, Reduction in PSIA Hf, BTU/lbm PSIA Hf, BTU/lbm Steam Flow %

900 669.375 850 679.301 1.4829 800 689.4 750 699.698 1.4938 700 710.218 650 721.001 1.5183 600 732.099 550 743.55 1.5641 500 755.431 450 767.827 1.6409 400 780.836 350 794.626 1.7661 300 809.391 250 825.462 1.9856 200 843.32 150 863.88 2.438

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 4 Table 3 Increase in Steam Generator Liquid Mass with 30% Tube Plugging Current C nfiguration 30% Plugged SG Tubes

Density, Density, Increase in SG SG Pressure Ibm/Ft3 SG Pressure Ibm/Ft3 Liquid Mass %

900 47.0813 850 47.4842 0.8558 800 47.8956 750 48.3166 0.879 700 48.7487 650 49.1937 0.9128 600 49.6534 550 50.1303 0.9605 500 50.6275 450 51.1489 1.0299 400 51.6996 350 52.2865 1.1352 300 52.919 250 53.6116 1.3088 200 54.3867 150 55.2836 1.6491 1.3. Parameters Affected By WCAP-9272 Process Transition to the WCAP-9272 reload methodology has the potential to produce changes in the analytical results, which impact a radiological consequences calculation. As the WCAP-9272 based analyses also include modeling with 30% of the steam generator tubes plugged, some of the changes listed are the result of the combined effect. In addition to the changes associated with the change in reload philosophy, a technical specification change is also being made in support of licensing dose analyses based on alternate source term methodology. This technical specification change will result in a change in the primary-to-secondary steam generator tube leakage limit from the current value of 1.0 gpm to 0.3 gpm total for the two steam generators.

1.4. Evaluation of Analysis of Record (AOR) Dose Analyses Against the 30% SGTP Program The following sections provide an event-by-event discussion of each event for which a dose analysis is performed to demonstrate the applicability of the current dose analysis of record to the work performed that supports 30% steam generator tube plugging.

1.4.1. Pre-Trip Steam Line Break (Inside and Outside Containment)

The current analysis of the pre-trip steam line break using the Combustion Engineering methodology limits the prediction of fuel rods experiencing DNB and predicted to fail for the purposes of a radiological consequences calculation. The WCAP-9272 analysis presented in the licensing submittal of 30% SGTP and the corresponding Cycle 15 core design has verified that the fuel failures for Cycle 15 are well below this limit. Nevertheless, this assessment will not credit the reduction in fuel failure in judging acceptability. Note that loss of offsite power (LOOP) assumptions (LOOP at reactor trip or 3 seconds later) will not make the current AOR dose consequences more adverse.

Table 4 lists the primary inputs to the radiological consequences calculation. In addition to the parameters listed in Table 4, the primary system mass should also be considered. Based on a primary system volume reduction associated with a 15% increase in steam generator tube plugging, the reduction in the total primary system mass is less than 10% from the current assumption of 435,000 Ibm. The radioisotopes carried from the primary system for release to

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 5 the atmosphere is the product of the tube leakage and the concentration in the primary coolant resulting from fuel failures. The decrease in the activity carried by the steam generator tube leakage, due to the lowered technical specification limit for the primary-to-secondary leakage, would have a beneficial effect which overwhelms the slight increase in the specific RCS activity due to the smaller primary system mass with plugged tubes.

The current analysis of record has a pre-trip steam line break limit for fuel failure of 33% (DNB) for an inside containment event, and 10.5% fuel failure limit (DNB) for an outside containment event. This led to 2 and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limiting thyroid doses of 144 and 288.75 REM, respectively.

Limiting whole body doses for 2 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> were found to be less than 1.0 and 1.5 REM respectively. These meet the 10 CFR 100 limits of 300 REM for thyroid dose and 25 REM for whole body dose.

Since the 30% SGTP analysis meets these fuel failure limits, the current analysis of record dose consequences would remain bounding for the 30% SGTP case.

Table 4 Pre-Trip Steam Line Break Inputs to the Radiological Consequences Calculation Parameter Analysis of Record 30% SGTP WCAP-9272 Result Fuel Failures Used for

< 33% DNB (inside Primary Activity Carried by containment)

Less than AOR limit SG Tube Leakage

< 10.5% DNB (outside containment)

Steam Generator Tube Leakage 1.0 GPM 0.3 GPM Initial Steam Generator 271,000 Ibm high SG setpoint Less than AOR Inventory analyzed*

• The analysis of record used a conservatively high initial steam generator inventory when calculating dose consequences for these events; most of this inventory is assumed to be released to the environment resulting in increased doses.

1.4.2. Post-Trip Steam Line Break (Inside and Outside Containment)

The current analysis of the post-trip steam line break using the Combustion Engineering methodology limits the prediction of fuel rods experiencing failure due to high linear heat rate and predicted to fail for the purposes of a radiological consequences calculation. The WCAP-9272 analysis and the Cycle 15 reload have verified that the fuel failures for Cycle 15 are well below this limit. Nevertheless, this assessment will not credit the reduction in fuel failure in judging acceptability. Note that loss of offsite power (LOOP) assumptions (LOOP at reactor trip or 3 second later) will not make the current AOR dose consequences more adverse.

Table 5 lists the primary inputs from the analytical result to the radiological consequences calculation. In addition to the parameters listed in Table 5, the primary system mass should also be considered.

Based on a primary system volume reduction associated with a 15%

increase in steam generator tube plugging, the reduction in the total primary system mass is less than 10% from the current assumption of 435,000 Ibm. The radioisotopes carried from the primary system for release to the atmosphere is the product of the tube leakage and the

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 6 concentration in the primary coolant.

The decrease in the activity carried by the steam generator tube leakage, due to the lowered technical specification limit for the primary-to-secondary leakage, would have a beneficial effect which overwhelms the slight increase in the specific RCS activity due to the smaller primary system mass with plugged tubes.

According to the current analysis of record, the fuel failure limit for dose consequences for a steam line break outside containment is 3.4% (centerline melt, CLM) and 13.5% (CLM) for a steam line break inside containment. This results in 2 and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> limiting thyroid doses of 93 and 300 REM, respectively. Limiting whole body doses are found to be less than 9.75 and 16.75 REM for 2 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, respectively. The fuel failures are however limited to less than 2% from DNB propagation viewpoint.

Under the new WCAP-9272 method, it was found that for the 30% SGTP, the fuel failures are well below the limits used in the current analysis of record for radiological consequences. The current analysis of record dose consequences would remain bounding for 30% SGTP.

Table 5 Post-Trip Steam Line Break Inputs to the Radiological Consequences Calculation Parameter Analysis of Record WCAP-9272 Result Fuel Failures Used for

< 13.5% CLM (inside Primary Activity Carried by containment)

Less than AOR limit SG Tube Leakage

< 3.4% CLM (outside containment)

Steam Generator Tube Leakage 1.0 GPM 0.3 GPM Initial Steam Generator 271,000 Ibm analyzed per 198,285 Ibm per SG (HZP)

Inventory SG*

• The analysis of record used a conservatively high initial steam generator inventory when calculating dose consequences for these events; most of this inventory is assumed to be released to the environment resulting in increased doses.

1.4.3. Feedwater Line Break Both the CESEC based analysis and the WCAP-9272 based analysis model the discharge of the inventory from one steam generator through the break. The two models have a different initial steam generator liquid inventory and thus, a different initial secondary system iodine activity for discharge from this source.

The radiological consequences analysis of record accounts for the discharge of the CESEC based steam generator activity along with the discharge of tube leakage at the technical specification concentration without iodine partitioning through the affected steam generator. Note that loss of offsite power (LOOP) assumptions (LOOP at reactor trip or 3 second later) will not make the current AOR dose consequences more adverse.

Discharge from the intact steam generator account for tube leakage into that steam generator and discharge during steaming with iodine partitioning.

Table 6 shows the only difference between the two analyses for the Feedwater line break.

The radiological consequences analysis of record calculated less than 2 REM thyroid for both the 2 and 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> dose values and less than 5 mREM whole body for both the 2 and 8-hour dose values. The feedwater line break reported less than 10% of the 10 CFR 100 limits [30 REM thyroid and 2.5 REM whole body].

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 7 The decrease in primary to secondary leakage from 0.5 gpm/SG to 0.15 gpm/SG is sufficient to compensate for any impact of the change in initial steam generator inventory.

Table 6 Feedwater Line Break Transient Analysis Inputs to the Radiological Consequences Calculation Parameter Analysis of Record WCAP-9272 Result Initial Steam Generator 271,000 Ibm analyzed per 123,840 Ibm per SG (HFP)

Inventory SG*

• The analysis of record used a conservatively high initial steam generator inventory when calculating dose consequences for this event; most of this inventory is assumed to be released to the environment resulting in increased doses.

1.4.4. CEA Ejection The control element assembly (CEA) ejection event is a result of a complete break of the control element drive mechanism (CEDM) housing or CEDM nozzle on the reactor vessel head. Fuel failure, which would lead to radiological consequences, occurs when specific enthalpy threshold values are violated. These include clad damage threshold which states that the total average enthalpy must remain below 200 cal/gm, the incipient centerline melting threshold which must not exceed 250 caVgm, and a fully molten centerline threshold which must remain below 310 cal/gm.

The above mentioned threshold values were met for the current analysis, therefore, no fuel failed. Radiological consequences, which are limited to a small fraction of the 10 CFR 100 guidelines, however, have assumed 0.05% of the rods are in centerline melting (CLM) and 9.5%

of the rods are in DNB.

The 30% steam generator tube plugging analysis, performed using the WCAP-9272 reload philosophy, found that the average fuel pellet enthalpy remained below 200 cal/g for all cases.

Thus this analysis is bounded by the current analysis of record.

During a CEA ejection, 0.5 weight %/day is released to the environment via the containment leakage, which is not impacted by the SGTP. A second source is steam generator tube leakage; the steaming from the SG's during cooldown.

Since this leakage rate has been reduced from 0.5 gpm to 0.15 gpm per steam generator for the 30% SGTP evaluation, the WCAP-9272 30% SGTP analysis becomes less limiting.

1.4.5. Inadvertent Opening of a Main Steam Safety Valve (MSSV)

Both the current analysis and the WCAP-9272 methodology have predicted no failure of fuel based on DNBR criteria for this event.

According to the current analysis, the feedwater line break event is analyzed to the same dose criteria as that of the inadvertent opening of a MSSV. The inadvertent opening of a MSSV is bounded by the feedwater line break since it is similar to the feedwater line break, no fuel failure was predicted. Bounding this event by a feedwater line break is possible because once the steam generator has run dry, radiological doses will steam out of the break in the feedwater line in the same manner that they would steam out of the MSSV.

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 8 Since the WCAP-9272 30% SGTP analysis has concluded no fuel failure for a feedwater line break event, the initial RCS and steam generator concentrations remain at the technical specifications limits similar to the current analysis. Reducing the primary-to-secondary leakage to 0.15 gpm per steam generator gives additional conservatism for this event.

1.4.6. Locked Rotor/Sheared Shaft The current UFSAR analysis of the locked rotor/sheared shaft using the Combustion Engineering methodology limits the amount of fuel rods experiencing failure due to DNBR criteria to 13.6%, which is based on the assumption of a stuck open atmospheric dump valve and the radiological dose acceptance criteria of within 10 CFR 100 limits for offsite doses.

Without the assumption of a stuck open ADV, the amount of fuel failures are limited to 2.5% to abide by the acceptance criteria which states that the radiological doses must remain within 10% of the 10 CFR 100 limit for offsite doses.

The 30% SGTP analysis performed using the WCAP-9272 reload philosophy shows that the maximum percent of rods-in-DNB (the percent of rods that experience a DNBR less than the limit value) is less than 2.5%.

The radiological dose consequences are dominated by the amount of failed fuel and the primary-to-secondary leakage. The current dose consequences therefore remain bounding for 30% SGTP with reduced total primary-to-secondary leakage rate of 0.3 gpm.

1.4.7. Letdown (Primary) Line Break A letdown (primary) line break may result from a break in a letdown line, instrument line, or sample line. The current analysis selects the double ended break of the letdown line outside containment upstream of the outside containment isolation valve as it is the largest line and results in the largest release of reactor coolant to the environment. The current analysis of record and the 30% SGTP analyses both do not directly challenge any DNBR criterion.

Total activity released due to a primary line break is a resultant of the break itself in conjunction with any doses steamed from the steam generator. The total activity released via the break is not affected by the 30% SGTP, since RCS specific activity and the mass release through the break would be the same as those for the AOR. For the 30% SGTP case, the total activity would only be reduced as a result of the new steam generator leakage rate.

The dose consequences from the current analysis of record thus would remain bounding for the 30%

SGTP case.

1.4.8. Loss of Coolant Accident (LOCA)

The assumptions and parameters used to calculate the radiological source terms and dose consequences for a LOCA, as described in the UFSAR, are not changed as a result of the 30%

SGTP. As such, the doses as given in the UFSAR remain unchanged for 30% SGTP case.

1.4.9. Fuel Handling Accidents (FHA) and the Waste Gas Decay Tank Rupture (WGDT)

These events are not affected by the 30% SGTP analysis. FHAs are not impacted by the increase in tube plugging as these events occur outside the RCS. Since there is no change to

St. Lucie Unit 2 Docket No. 50-389 L-2005-004 Attachment 2 Page 9 the RCS Technical Specification activity, the current WGDT rupture analysis remains bounding for 30% SGTP.

Conclusions A set of non-LOCA safety analyses that support the operation with 30% SGTP recently completed by FPL have transitioned from the CESEC based Combustion Engineering process to the RETRAN based WCAP-9272 analytical process. In order to determine whether or not the radiological calculations under the current analysis are valid for the 30% SGTP WCAP-9272 method, key inputs are compared between the two analyses.

The majority of the parameters found in radiological consequences do not change as a result of the tube plugging or the WCAP-9272 method as they are governed by regulation, technical specifications or agreed upon analytical assumptions in licensing interactions (Table 1). Steam generator inventory, on the other hand, is impacted due to the 30% SGTP. A reduction in the heat transfer area in the steam generator due to the plugging results in a reduction of steam flow (Table 2). In addition, due to a lower saturation pressure resulting from the 30% SGTP, there is an increase in SG liquid mass (Table 3). Both changes are beneficial when calculating radiological consequences as they are in the conservative direction. A change in the steam generator tube leakage TS from the current 1 gpm to 0.3 gpm has a significant beneficial effect on the dose consequences.

After key inputs from the radiological consequences calculation from the recent set of (non-LOCA) analyses were compared to the 30% SGTP WCAP-9272 method based analyses, it was found that there were no dose violations in accordance to 10 CFR 100 limits. The radiological dose calculations from the current analysis of records for the above mentioned transients remain valid under the 30% SGTP.