ML20236R866

Submits Response to NRC RAI Re License Amend Request to Reduce Dose Equivalent Iodine for Braidwood Station

ML20236R866
Person / Time
Site:	Braidwood
Issue date:	07/17/1998
From:	Tulon T COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9807220327
Download: ML20236R866 (10)

Text

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Onumonwealth Ednon Company liraidwood 6cncrating Station a

Route el. Ilox Hi lira ( es inc, i1. 60 60'-96 I9 Tel H15-458 2801 July 17,1998 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Subject:

Braidwood Nuclear Power Station, Units 1 and 2 Facility Operating Licenses NPF-72 and NPF-77 NRC Docket Numbers: 50-456 and 50-457 AdditionalInformation Related to License Amendment Request to Reduce Dose Equivalent lodine for Braidwood Station, Unit 1

References:

1) T. J. Tulon to U.S. NRC letter dated January 14,1998. Request for Amendment for Lowering RCS Dose Equivalent lodine. Braidwood Nuclear Power Station, Units 1 and 2.

2)IL Gene Stanley to USNRC letter dated January 14,1998 providing the Braidwood Unit 13.0 Volt !PC Full Cycle Operation Technical Basis Supplement to Braidwood Unit 1 Cycle 7 IPC Report.

3) S. N. Bailey (U.S. NRC) to O. D. Kingsley letter dated June 4,1998 requesting additional information related to amendment request to reduce dose equivalent iodine.

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Reference 1 transmitted the Commonwealth Edison Company's (Comed) request to 2

amend Appendix A, Technical Specifications, for Facility Operating Licenses NPF-72

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and NPF-77, for the Braidwood Nuclear Power Station, Units 1 and 2, respectively. The changes would reduce the allowable Unit 1 Reactor System (RCS) Dose Equivalent lodine-131 (DE I-131), activity from 0.35 microCuries/ gram to 0.05 microcuri The Technical Specification change is necessary to provide additional margin to the maximum site allowable leakage limit for the predicted main steam line break leakage.

The main steam line break leakage due to implementation of 3.0 Volt IPC was conservatively calculated as discussed in Reference 2.

l Braidwood 1 is currently operating under an operability assessment which limits RCS del-131 to 0.05 microCuries/ gram for the remainder of Cycle 7 until steam generator replacement.

9007220327 980717 PDR ADOCK 05000456' P

PDR A t'nicom company

i Documen,t Control Deskr L

. July 17,'1998 Page 2

- Reference 3 transmitted the NRC request for additional information regarding leakage during locked rotor and rod ejection events from implementation of 3.0 Volt IPC.

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Response to the request for additional information is provided in the two attachments to -

l this letter. A consideration of the locked rotor and control rod ejection accidents is contained in Attachment 1. An offsite dose assumption comparison between the locked Rotor event and the steamline break accident is contained in Attachment 2.

' This response affects Braidwood Unit 1 only, but is being submitted for Braidwood Unit I and Braidwood Unit 2 because the Reference I request was docketed for both units.

- Please address any comments or questions regarding this matter to T. Simpkin at (815) 458-2801 extension 2980.

Sincerely, -

ot y J. 'ulon ite Vice President

- Braidwood Nuclear Generating Station Attachments: 1) Consideration of the Locked Rotor and Control Rod Ejection Accidents

2) Locked Rotor and Steamline Break Offsite Dose Assumption Comparison 1

cc:

. C.= J. Paperiello, Acting Regional Administrator-Rill C. J. Phillips, Senior Resident inspector-Braidwood nrc.. 98042tjt. doc h

Consideration of the Locked Rotor and Control Rod Ejection Accidents l

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U.S. NRC request for additional information (dated June 4,1998):

" Provide an assessment to confirm that the main steamline break (MSLH) is the bounding DBA for offsite and control room doses for the level of steam generator tube degradation that is assumed as tise basis for the proposed license amendment.

The assessment should consider (1) increased primary-to-secondary leakage during j

a Locked Rotor or Control Rod Ejection accident,(2) any assumed fuel failure for l

these accidents, and (3) potential for steam generator (SG) tube uncovery. Comed i

may include, as part ofits assessment, an evaluation of the constraining effect against leakage provided by the support plates and corrosion product in the tube-to-l support plate crevices under the assumed accident pressure / temperature conditions, l

including the potential for steam flashing to cut through the corrosion product.

Supporting data, analyses, and assumptions should be provided."

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1 Locked Rotor:

i The strategy used to show that the consequences from a main steam line break bound l

l those of a locked rotor event is two step. First the leak rates for the two events are compared by evaluating the conditions during the two events. The second step compares j

the inputs and assumptions used in the two events dose assessments.

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1. Increased Primary to Secondary Leakage The Braidwood Unit I licensing and design basis for a locked rotor accident is that no failed fuel results from the accident and the primary to secondary leakage is I gpm at normal operating pressure and temperature (NOTP). The locked rotor event results in a l

maximum primary pressure of 2720 psia 3.6 seconds into the event while the secondary side remains pressurized at a no load saturation pressure of approximately 1100 psia for a maximum pressure differential of 1620 psid compared to the main steam line break l

pressure differential of 2560 psid. Therefore, the primary to secondary leak rate during a l

locked rotor event is bounded by the main steam line break leak rate due to the reduced pressure differential across the tube wall.

2. Assumed Fuel Failure l

The design basis locked rotor and main steam line break events result in 0% failed fuel.

l A comparison of key input assumptions for the consequences evaluation between a main steam line break and a locked rotor event was performed and it was determined that the I

locked rotor event inputs and assumptions are bound by the main steam line break. A summary of this comparison is included Table 2 of Attachment 2.

l A similar conclusion was reached in Supplement 1 of NUREG - 1002," Safety Evaluation Report related to operation of Braidwood Station. Units 1 and 2," Section 15.3.6," Reactor Coolant Pump Rotor Seizure and Shaft Break". In this SER it was concluded that in the absence of DNB during the accident fuel failure is not predicted to l

occur. The SER goes on to snte for a locked rotor accident "In the event that a secondary relief valve stuck open, the utisite does consequences would be bounded by those of a postulated main steamline break outside containment. These results have already been found to be acceptable by the staffin SER Section 15.4.2."

The current UFSAR Section 15.3.3 analysis for the locked rotor transient response and current reload analysis for which the Technical Specification change is requested predicts no fuel failure during the locked rotor event. UFSAR Section 15.3.3.5 analyses for the locked rotor radiological consequence determines that consequences are bound by a Main Steam Line Break event which uses 0% failed fuel during the event. UFSAR Section 15.3.3.3 discusses an additional radiological consequences analysis assuming 5% fuel failure during the event to cover the potential fuel failure prediction for future reload changes. This condition is not applicable to the current cycle since Cycle 7 reload analysis shows no fuel failures. Therefore,0% fuel failure during the event is used in this i

assessment with an initial condition of 1% fuel failure existing at steady state conditions.

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Conclusion:==

Based on evaluation ofleak rates and dose assessment inputs and assumptions the locked rotor event is bound by the main steam line break event.

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l Control Rod Election:

The strategy used to show that the consequences from a main steam line break bound l

those of a rod ejection event is discussed here. First, a predicted leak rate for the rod ejection event is identified and compared to that used in the rod ejection design basis analysis documented in the UFSAR. Then the predicted leak rate is used to adjust the control rod ejection accident dose by using the ratio of the predicted leak rate to that used in the design basis analysis. The dose is then compared to the main steam line break dose.

1. Increased Primary to Secondary Leakage:

The Braidwood Unit I design basis accident for a control rod ejection does not result in a significant pressure differential increase compared to normal full power operation. After an initial small increase in pressure the rod ejection accident results in a de-pressurization of the primary coolant due to the loss of coolant inventory. The primary side pressure increase is less than 100 psi during the accident while the secondary side remains pressurized. With this minimal change in pressure differential the leakage during the accident can be considered the same as during normal full power operation which is limited by Technical Specifications to 600 room temperature (RT) gallons per day (gpd) in all four SG's.

Analysis of consequences for a Postulated Rod Ejection Accident assumes a primary to secondary leakage of I gpm (1440 RTgpd). Therefore the margin in the consequences analysis compared to the actual accident leakage (same as normal full power operation) is 1440/600 or 2.4. The dose analysis results are adjusted to the Technical Specification allowable leak rate in Table i below.

2. Assumed Fuel Failure:

Comparison of the rod ejection (10% failed fuel) and main steam line break (0% failed fuel) consequences analyses is shown below for the applicable predicted accident leak rate.

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Table 1: Control Rod Ejection Assessment Accident Leak Rate Predicted Leak Accident Predicted Acceptance Assumed in Rate (RTgpm)

Analysis Dose Accident Criteria Accident Analysis

(@lpci/g)

(Thyroid)

Dose (Thyroid)

(RTgpm)

(REM)

(Thyroid)

(REM)

(fitlpci/g)

(REM)

Control Rod 1

0.4167 59.9 25 75 Ejection MSLB (Accident 6.64' 6.64 30 30 30 Initiated Spike at Exclusion Area Boundary)

• The present Technical Specification Amendment request for a DEI limit of 0.05 pei!g would translate into a site allowable leak limit at MSLB conditions of 132.8 gpm (6.64 / 0.05) at room ten, wture conditions for use in dose assessment considerations.

The dose due to a rod ejection accident (25 REM) is bound by the main steam line break accident (30 REM).

it is also noted that since significant tube support plate (TSP) movement is not expected during a rod ejection accident, packed TSP crevices would preclude any significant leak through outside diameter stress corrosion cracking (ODSCC) indications. This is discussed in WCAP-14707, Revision 1,"Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube to Tube Support Plate Crevices," January 1997. The WCAP discusses leak rate tests performed on packed crevices. Leak rates measured for a throughwall hole drilled in the tube at the center of the TSP, which would leak at about 90 liters /hr under free span conditions, were < 0.1 liter /hr (0.00044gpm) with no TSP displacement for a reduction of about 1000 on the leak rate for packed crevice conditions. Similar results would be expected at Braidwood Unit I where similar tube and TSP materials have been used and no chemical cleaning has been performed. This WCAP result is supported by the in situ leak test of a domestic plant tube pull that had a throughwall crack of 0.42 at a TSP but did not leak at normal operating conditions.

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Conclusion:==

Comparison of the dose for the rod ejection and main steam line break events shows that the rod ejection event is bound by the main steamline event. In Reference 2 Comed demonstrated that the control room dose is bounded by the LOCA event.

Potential for Steam Generator Tube Uncovery:

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The assumption of no long-term tube uncovery is supported by the results of a Westinghouse Owners Group Program, Report on the Methodology for the Resolution of the Steam Generator Tube Uncovery Issue, WCAP-13247. A conclusion of the program indicates that steam generator tube uncovery does not increase the consequences of SGTR and non-SGTR events significantly. The current design basis accident analysis methodologies are adequate and remain valid.

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Locked Rotor and Steamline Break Offsite Dose Assumption Comparison i

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(ocked Rotor and Steamline Break Offsite Dose Assumption Comparison Failed fuel of 0% is assumed for the locked rotor and steamline break events; the two cases will be analyzed with the other assumptions remaining unchanged. A comparison of key input assumptions between steam line break and locked rotor for such a scenario is I

shown in Table 2. Information in Table 2 is obtained from Byron /Braidwood UFSAR I

Table 15.1.3 for steam line break and Table 15.3-3 for locked rotor.

I Table 2 - Comparison of Key Assumptions for 0% Accident Induced Fuel Failure Assumption Steam line Break Locked Rotor Power 3565 M Wt 3565 MWt Fraction of Fuel with Defects 1%

1%

Before the Event Reactor Coolant Activity Prior to Accident-initiated spike case:

Accident-initiated spike case -

Accident 500 times the maximum would use the same assumption equilibrium primary system as steam line break for 0% fuel iodine concentration of 1 pCi/gm

failure, of D.E.1-131. The duration of the spike is assumed to be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Preaccident spike case : 60 Preaccident spike case: would use pCi/gm of D.E.1-131 the same assumption as steam line break for 0% fuel failure.

I Secondary Coolant Activity Prior 0.1 pCi/gm of D.E.1-131 0.1 pCi/gm of D.E.1-131 to Accident Duration of Plant Cooldown by 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 40 Hours Secondary System Afler Accidmt Offsite Power lost lost Total Steam generator Tube Le ik 9.4 gpm Igpm rate During Accident (NOTP)

Steam Generator lodine Partitio1 1.0 in defective generator 0.01 in all generators factor 0.1 in nondefective generator initial Steam Release from 96,000 lb (0 - 2 hr)

N/A Defective Generator Long Term Steam Release from 129,945 lb (0 - 40 hr)

N/A Defective Generator Steam Release from 3 406,716 lb ( 0 - 2 hr)

N/A Nondefective Generators 939,604 lb ( 2 - 8 hr) 665,215 lb ( 8 - 16 hr) 569,300 lb (16 - 24 hr) 510,514 lb (24 - 32 hr) 470.292 lb (32 - 40 hr)

Stear.. Release from 4 Steam N/A 561,540 lb (0 - 2 hr) g generators 3.791,260 lb (2 - 40 hr)

Table 2 shows the leak rate, iodine partition factor, and amount of steam releases to be different. Each of these assumptions is discussed below.

The iodine partition assumed for the steam line break event is larger than, and therefore bounds, the locked rotor event. But the steam releases assumed for the steam line break

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event is smaller than, and therefore does not bound, the locked rotor event. These two assumptio' ns need to be combined to determine which event is limiting.

The effect.ve steam release from the steam line break event from all four generators can be calculated as follows for the first 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

1.0

• 96,000 lb + 0.1

• 406,716 lb = 1.4E5 lb 3

l For the locked rotor event:

0.0l* 561540 lb = 5.6E3 lb Therefore, the effective steam release from steam line break bounds locked rotor for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

i The effective steam release from the steam line break event from the three unfaulted steam generators can be calculated as follows for the remaining 38 hour4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br />s:

0.1 * (939,604 lb + 665,215 lb + 569,300 lb + 510,514 lb + 470,292 lb) = 3.2E5 lb A discussion of the leakage from the faulted steam generator during the steam line break event is discussed below.

For the locked rotor event:

0.01

• 3,791,260 lb = 3.8E4 lb Therefore, the effective steam release from steam line break bounds locked rotor for the remaining 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> of cooldown.

The long term release from the defective steam generator for the steam line break event comes from primary to secondary leakage. The leakage is assumed to be released directly to the environment with no retention or partition in the secondary coolant. For the locked rotor radiological consequence discussed in UFSAR Section 15.3.3, partition in the secondary coolant is assumed. To keep the evaluation simple and conservative, as long as the leak rate for steam line break is greater than or equal to the leak rate for a locked rotor event, the steam line break event bounds the locked rotor event for steam release due to primary to secondary leakage. Due to the higher pressure differential during a main steam line break event compared to a locked rotor the leak rate for a locked rotor will be bound by the main steam line break.

Based on the above discussion, it can be concluded that, for 0% fuel failure, the radiological consequence from the steam line break event bounds the locked rotor event.

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