ML20195G327

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Safety Evaluation Supporting Amends 237 & 227 to Licenses DPR-77 & DPR-79,respectively
ML20195G327
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 11/17/1998
From:
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20195G316 List:
References
GL-95-05, GL-95-5, NUDOCS 9811200259
Download: ML20195G327 (8)


Text

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066H001 l

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

RELATED TO AMENDMENT NO.m TO FACILITY OPERATINC LICENSE NO. DPR-77 AND AMENDMENT NO. ??7 TO FACILITY OPERATING LICENSE NO. DPR-79 TENNESSEE VALLEY AUTHORITY j

SEQUOYAH NUCLEAR PLANT. UNITS 1 AND 2 I

DOCKET NOS. 50-327 AND 50-326

1.0 INTRODUCTION

in a letter to the U.S. Nuclear Regulatory Commission (NRC) dated June 26,1998, the Tennessee Valley Authority (TVA) requested amendments to Operating Licenses DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SON), Units 1and 2, respectively. The amendments',

I would modify Technical Specification (TS) Figure 3.4-1, reducing the maximum allowable I

instantaneous value of dose equivalent iodine-131 (l)in primary coolant as a function of reactor power level and modify TS 3.4.8.a, reducing the 48-hour value for dose equivalent ' 'l in primary coolant. In conjunction with these proposed changes, TVA also planned to increase the maximum allowable primar/-to-secondary leakage for the faulted steam generator (SG)

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assumed in their analysis of the consequences of a main steamline break (MSLB) accident.

The proposed increase was from 3.7 gpm to 8.21 gpm at 70 F.

With the proposed change to TS Figure 3.4-1, the unacceptable range of operation, the maximum allowable specific activity level at which the plant would be required to initiate shutdown actions at rated thermal power levels of 80% or above, would now be at primary 3

coolant acti.'ty levels greater than 20 rnicrocuries per gram (pCi/g) of dose equivalent I.

i Previously, this value had been 60 pCilg. A similar reduction in primary coolant activity levels of j

dose equivalent '8'l in the power range 20-80% was incorporated into the figure.

j TVA proposed to lower the 48-hour value of dose equivalent "'l in primary coolant in TS 3.4.8.a to 0.35 pCilg from the previous value of 1.0 pCilg. TVA considered such a reduction in primary coolant activity level appropriate with the concomitant increase in primary to secondary leakage and consistent with the guidance provided in NRC Generic Letter (GL) 95-05, ' Voltage-Based i

Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking."

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2.0 BACKGROUND

j TVA previously requested changes to the SON Units 1 and 2 TSs (via Change Requests 95-15 and 95-23) to add an alternate SG tube plugging criteria for tubes with outside diameter stress corrosion cracking (ODSCC) indications at nondented tube support plate intersections in accordance with GL 5-05. The NRC approved these changes in TS Amendments 214 (Unit 1) 7 9811200259 901117 1

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and 211 (Unit 2), dated October 11,1995, and April 3,1996, respectively. At the time that these amendment requests were made, SON did not request a decrease in the specific activity of the primary coolant dose equivalent '* I in order to expedite the review process. The NRC staff concluded that such a reduction in primary coolant activity level was not required since the NRC calculated doses for the control room, the Exclusion Area Boundary (EAB), and the Low Population Zone (LPZ), still met the staff's acceptance criteria for an MSLB accident even with the increase in primary to secondary leakage resulting from the implementation of the attemate -

repair criteria.

l The altemate SG tube plugging criteria require each indication left in service be evaluated and assigned a leakage quantity that would be postulated to occur in the event of an MSLB l

accident. The summation of the leakage quantity must be less than a specified value so that the resultant offsite dose would be a fraction of the 10 CFR Part 100 offsite allowable dose limit.

Since the offsite done is directly dependent on the reactor coolant specific activity of dose equivalent l, decreasing the maximum allowable reactor coolant specific activity will allow a larger quantity of tubes with axial ODSCC to remain in service by allowing a proportional increase in primary to secondary leakage during a postulated MSLB accident.

1 Currently, TVA projects the end-of-cycle MSLB leak rate from tubes left in service to be less than 4.0 gallons per minute (gpm) total or 3.7 gpm for the SG in the faulted loop based on thk reactor coolant specific activity limited to 1.0 pCi/g dose equivalent '8'I. This projection includes a probability of detection adjustment, allowances for nondestructive examination uncertainties, and ODSCC growth rates. If the maximum allowable 48-hour value dose equivalent l in primary coolant activity is reduced to 0.35 pCi/g, the primary-to-secondary leakage rate associated with an MSLB accident could be increased to 11.9 gpm total or 11.6 gpm for the SG in the faulted loop at operating conditions (2250 psia and 590'F). This leak rate equates to 8.51 gpm total and 8.21 gpm for the faulted loop at atmospheric conditions and 70*F.

Presently, indications less than the alternate repair limit (i.e., indications allowed to remain in service by the alternate plugging criteria) are required to be plugged or repaired in order "

prevent the allowable leakage limit from being exceeded.

' 3.0 EVALUATION 3.1 Assessment of Radioloaical Conseauences 3.1.1 Background i

TVA performed an evaluation to determine the maximum permissible SG primary-to-secondary leak rate during an MSLB. The evaluation considered both pre-existing and accident-initiated iodine spike cases as required by GL 95-05.

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TVA selected 30 rem as the thyroid dose acceptance criteria for an MSLB, with an assum l

accident-initiated iodine spike, based on the guidance of the Standard Review Plan (NURE 0800) Section 15.1.5, Appendix A. They selected 50 percent of the 10 CFR Part 100 th dose guideline, or 150 rem, as the thyroid dose acceptance criteria for the case of a pre-l existing iodine spike. TVA also calculated the whole-body doses to both offsite and control l

room personnel, as well as the skin and thyroid doses to control room personnel. The EAB l

doses were calculated for the initial 2-hour period following the MSLB. It is assumed the operator takes action to cool down and depressurize the plant, and place the residual heat removal (RHR) system into service for further reactor coolant system (RCS) heat removal t

within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident. Once the RHR system is placed into service and the RCS has been depressurized, there are no more steam releases from the faulted and intact SGs. Thus i

LPZ and control room doses are based on activity releases for the initial 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following th MSLB. TVA extended the control room dose calculation beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (to 30 days) because i

activity will remain in the control room atmosphere beyond the period of time in which it is brought initially into the control room. This activity remains in the control room until it is removed either via filtration and purging due to outleakage from the control room. TVA determined that the quantity of activity was no longer significant after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After that there was no increase in the thyroid dose.

TVA indicated that the methodology and assumptiohs performed in support of this amendrnent request are the same as the evaluation documented in Section 4.0 of Westinghouse Commercial Atomic Power Topical Report 13990 (WCAP-13990), entitled "Sequoyah Units 1 and 2 SG Tube Plugging Criteria for Indications at Tube Support Plates," May 1994, with the following differences (WCAP-13990 was transmitted to NRC with SQN TS Changes 95-15 and 95-23):

1.

Initial primary coolant iodine activity - 0.35 pCi/g dose equivalent "'l.

2.

Initial secondary coolant iodine activity - 0.1 pCi/g dose equivalent "'I (at NRC request during a previous submittal).

3.

The iodine partition coefficient for the primary-to-secondary leakage in the intact SGs was assumed to be 0.01 to reflect that the leakage is below the mixture level.

i The results of the TVA evaluation led them to conclude that the thyroid dose at the EAB for the accident-initiated spike case yields the limiting leak rate. TVA drew this conclusion based on a 30-rem thyroid dose for the accident-initiated spike case and initial primary and secondary coolant iodine activity levels of 0.35 pCi/g and 0.1 pCi/g dose equivalent "'l, respectively.

Based upon these conditions, TVA calculated a limiting leak rate of 11.9 gpm total or 11.6 gpm to the faulted SG at operating conditions of 2250 psia and 590*F. At atmospheric conditions i

and 70"F, this leak rate equates to 8.51 gpm total and 8.21 gpm for the faulted SG and i

150 gallons per day (approximately 0.1 gpm/SG) for the three intact SGs.

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3.1.2 Analysis The staff has reviewed the licensee's calculations and performed confirmatory dose calculations for an MSLB accident using the methodology associated with Standard Review Plan (SRP) 15.1.5, Appendix A. Two assessments were performed. One was based upon a pre-existing iodine spike activity level of 20 pCi/g of dose equivalent '8'l in primary coolant and the other was based upon an accident-initiated iodine spike activity level of 0.35 pCi/g of dose equivalent '8'l in primary coolant. Fnr the accident-initiated spike, the staff assumed that the accident-initiated spike case resulted in an increase in the release rate of iodine from the fuel by a factor of 500 over the release rate to mamtain an activity level of 0.35 pci/g of dose equivalent

l. For the accident-initiated spike and the pre-existing spike cases, the staff calculated doses J

for individuals located at the EAB, LPZ and the control room. The parameters that were used in the staff's assessment are shown in Table 1. The resulting doses calculated by the staff are shown in Table 2.

In WCAP-13990, the atmospheric dispersion factor, x/Q, for the 0-8 hour period for the LPZ and control room was divided into two increments,0-2 hours and 2-8 hours. During conference calls with TVA on September 17,21, and 25,1998, these x/O values were discussed. TVA explained that the SON interim plugging criteria amendment x/Q values are the same values which were presented in Table 15A-2 of the SON Final Safety Analysis Report (FSAR). Thehe values were calculated from a year's worth of site meteorological data from the period April 1971 through March 1972 and remain unchanged since the plant was originally licensed. TVA did not make separate calculations of the control room x/Q values specifically for the postulated MSLB accident for SON. Instead, TVA explained that the values used in this MSLB amendment request analysis (i.e., the factor of two increase above the SQN loss-of-coolant accident (LOCA) x/O values) reflect engineering judgment based upon the SON building geometry and a comparison with the ratio of the MSLB and LOCA control room x/Q values

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calculated for the Watts Bar plant. The licensee noted the strong similarity of the Watts Bar and SON plant building geometries. In addition, the staff noted that the wind speed estimates used in the licensee's assessment were somewhat different than expected, based on data presented in Table 2.3.2-1 of the SON FSAR.

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Since the licensee's control room x/Q values for an MSLB accident release point were based 2

on the application of engineering judgments to the LOCA control room x/Q values, for the confirmatory calculations, the staff decided to utilize the staffs operating license safety evaluation x/Q values for the LPZ and to calculate a x/Q value for the control room using the Murphy /Campe methodology of SRP 6.4. The staff calculated a value for the control room and applied it for the 0-2 hour period. To estimate a 2-8 hour v&e, the staff then took the 0-2 hour value and divided it by a factor of 4 in accordance with the guidance of SRP 6.4 for control room designs with dual inlets and the capability to manually select the inlet with the lower inlet activity level. The control room design qualified for this factor of 4 by meeting the Seismic Category 1, tornado missile, redundant radiation monitor requirements, and the single failure criteria presented in SRP 6.4.

With the incorporation of the above noted x/O values, the staff calculated the LPZ and control room operator doses to be within the guidelines of SRP 15.1.5, Appendix A, and SRP 6.4.

Therefore, the staff concluded that the licensee's analyses are acceptable and that a leak rate

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of 8.21 gpm is an acceptable limit for the maximum primary-to-secondary leakage initiated in the faulted SG for the MSLB ' ccident when coupled with the proposed change in primary a

coolant activity levels of dose equivalent '8'l.

GL 95-05 states thht lowering the primary coolant dose equivalent l activity is an acceptable means for accepting higher projected primary-to-secondary SG leakage rates during a postulated MSLB accident. Therefore, based on the above stated dose values being a smhil fraction of 10 CFR Part 100 dose guideline values and consistent with SRP (NUREG-0800) acceptance criteria. Therefore, the proposed changes to TS Figure 3.4-1 to change the dose equivalent ' 'l coolant specific activity limit versus percent of rated power, and to reduce the primary coolant specific activity limit from 1.0 to 0.35 uCi/ in TS 3.4.8, are acceptable.

9 Associated changes were also made to the TS BASES consistent with the corresponding analysis of the MSLB.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changet a requirement with respect to installation or use of a facility i

component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 17235, dated April 8,1998). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmentalimpact statement or environmental assessment need be prnpared in connection with the issuance of the amendments.

6.0 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 rnanner, (2) st :h activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Hayes L. Brown Dated: Noveber 17, 1998

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.o TABLEI INPUT PARAMETERS.FOR SQN EVALUATION OF MAIN STEAMLINE BREAK ACCIDENT

1. Primary coolant concentration of 20 pCilg of dose equivalent "'I.

Pro.existina Snike Value fuCl/a) i~

l-131 15.4

=

4 l-132 ' =

5.6 l-133

=

24.7 l-134 =

3.5 j'

l-135

=

13.6

2. Volume of primary coolant and secondary coolant:

,i Primary Coolant Volume (ft')'

12,600 Primary Coolant Temperature (*F) 590 Mass of Primary Coolant (Ib) 554,000 Secondary Coolant Steam Volume (ft')

3,546 Secondary Coolant Liquid Volume (ft')

2,322 Secondary Coolant Steam Temperature (*F) 526.2 Secondary Coolant Feedwater Temperature ('F) 434.6

3. Technical Specifications limits for DE l-131 In the primary and secondary coolant:

Primary Coolant DE l-131 concentration (pCilg) 0.35 Secondary Coolant DE l-131 concentration (pCilg) 0.10

4. Technical Specifications value for the primary to secondary leak rate.

Primary to secondary leak rate, maximum any SG (gpd) 150 Primary to secondary leak rate, total all SGs (gpd) 600

5. Maximum primaty-to-secondary leak rate to the faulted and intact SGs:

Faulted SG(gpm) 8.21 Intact SGs (gpm/SG) 0.1 i

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6. lodine Partition Factor.

4 Faulted SG 1.0 Intact SG 0.01 Primary to Secondary Leakage 1.0 j

7. Steam Released to the environment Faulted SG (lbs,0-2 hours) 87,000 Intact SGs (Ibs,0-2 hours) 479,000 (lbs,2-8 hours) 1,030,000 a

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8. Letdown Flow Rate (gpm) 75
9. Release Rate for 0.35 pCilg DE l-131:

Cilhour l-131 1,750

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1-132 4,300

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1-133 4,415

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1-134

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6,316 l-135 4,515

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10. Atmospheric Dispersion Factors (s/m')

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EAB (0 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 1.64 x 104 i

LPZ (0-8 hours) 1.96 x 104 Control Room (0-2 hours) 5.2 x 104 Control Room (2-8 hours) 1.3 x 104 j

11. Control Room Parameters I

h Filter Efficiency (%)

95 Volume (ft')

260,000 Makeup flow (cfm) 1,000 Recirculation Flow (cfm) 2,600 Unfiltered inleakage (cfm) 51 Occuptncy Factors 0-1 day 1.0

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Table 2 THYROID DOSES FROM SQN MAIN STEAM LINE BREAK ACCIDENT (REM)

LOCATION DOSE Pre-existina Solke Accident-initiated Solke**

EAB 47.5*

26.9 LPZ 7.0*

15 Control Room **

14.6 21

  • Acceptance Criterion = 300 rem thyroid
    • Acceptance Criterion = 30 rem thyroid t

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