ML20236D194

Safety Evaluation Supporting Amends 60 & 52 to Licenses DPR-77 & DPR-79,respectively

ML20236D194
Person / Time
Site:	Sequoyah
Issue date:	10/19/1987
From:	NRC OFFICE OF SPECIAL PROJECTS
To:
Shared Package
ML20236D182	List: ML20236D178 ML20236D194
References
NUDOCS 8710280062
Download: ML20236D194 (5)

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%g UNITED STATES 3.-

NUCLEAR REGULATORY COMMISSION i

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WASHINGTON, D. C. 20555

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SAFETY ~ EVALUATION BY THE OFFICE 0F SPECIAL PROJECTS

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j LSUPPORTING AMENDMENT N0; 60 TO FACILITY OPERATING LICENSE NO. DPR-77;

AND AMENDMENT NO. 52 TO FACILITY OPERATING' LICENSE NO. OPR TENNESSEE VALLEY AUTHORITY

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'SEQUOYAH h0 CLEAR PLANT, UNITS 1 AND 2

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DOCKET NOS. 50-327 AND 50-328

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1.0 INTRODUCTION

.By letter dated October 22,.1985'the Tennessee Valley Authority (TVA) requested several-changes to the Technical-Specifications for the Sequoyah Nuclear Plant,

. Units 1 and-2.. This safety evaluation' addresses the proposal to increase the c

setpoints of the ' radiation monitors in the spent fuel pool areas and to clarify -

the enrichment-limits of the fuel stored in the spent fuel. pool and new fuel l

. storage areas.

2.0' EVALUATION j

.The' purpose lof the spent fuel pool radiation monitors (RM-90-102 and -103) is to automatically isolate the normal building ventilation system, which vents' through.the stack to atmosphere, and to initiate the auxiliary buildint; gas i

treatment system..which filters bui1 ding air through charcoal and HEPA filters, upon detection of high. radiation in the auxiliary building. The intent is to

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limit doses 'outside the plant site to acceptable levels in the event of a l

radiation release in the building.

.The present.setpoint on radiation monitors RM-90-102 and -103 is 15mR/hr. When 1

the setpoint is exceeded, an auxiliary building isolation (ABI) signal is generated and the building is isolated.

The licensee has stated that the present setpoint is so low that 13 spurious ABI signals were generated just in 1984.

This situation is undesirable so the licensee proposed to change the setpoint to 200 mR/hr.

The staff evaluated this proposed change by reviewing the fuel handling accident in accordance with Standard Review Plan, Section 15.7.4.

The results, summarized in Table.1, show that the thyroid dose is acceptably low (not more than 75 rem) if the auxiliary. building has been isolated. These results are consistent with the initial staff safety evaluation for the plant. NUREG-0011.

Therefore, the staff evaluated the proposed setpoint to ensure that the revised setpoint is sufficiently sensitive to provide building isolation for i

this event. The ganna radiation level that could be expected to be associated with a fuel handling accident was evaluated and found to be capable of actuating' the isolation signal at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, the time specified in the SRP i

8710280062 871019' gDR ADOCK 05000327 PDR

) (Table 2).

Furthermore, a comparison of the estimated gamma radiation levels in the fuel handling area, and the estimated off-site thyroid doses, indicated that as long as the radiciodines are capable of delivering doses corresponding

.to the 75 rem criterion, the noble gases will be capable of producing dose rates greater than 200 mR/hr (Table 3).

The proposed change in setpoints will limit doses within the SRP acceptance criteria and, therefore, the change is acceptable.

Specification 5.6.1.1 would be altered to insert the value of the maximum enrichment of the fuel to be stored in the spent fuel racks.

This value is 4.0 weight percent U-235 and has been previously approved in Amendments 13 and 4 to facility operating licenses DRP-77 and OPR-79, respectively, by letter dated May 4, 1982. This insertion clarifies the storage requirements and is acceptable.

The change to Specification 5.6.1.2 would add the caveat that new fuel enrich-ment is restricted to 4.0 weight percent of U-235. This restriction is inserted because although the new fuel storage area has been cualified for storage of fuel up to 4.5 weight percent U-235, other restrictions (TS 5.3.1 and 5.6.1.1) limit new reload fuel assemblies, which might be stored on these racks, to 4.0 weight percent U-235. The addition clarifies the storage requirements and is acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

These amendments involve a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendments involve n'o 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 these amendments involve no significant hazards consideration and there has been no public comment on such finding.

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 environmental impact statement nor environmental assessment need be prepared in connection with the issuance of the amendments.

4.0 CONCLUSION

We have 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, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defense and security nor to the health and safety of the public.

Principal Contributor:

C. Willis and E. McKenna Dated: October 19, 1987

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Table 1 SEQUOYAH Fuel.HandlingLAccident' Analysis

' Power, MWt' 3582 Assemblies in the core.

193 i

i Assemblies.that. fail 1

Radial' peaking. factor _

1.65 Fraction of activity.in the gap 0.1 Pool decontamination factor 100 Iodine fractions released from the pool

-Elemental 75 percent l

1 Organic.

'25 percent

'I Filter Efficiencies for Tedine Removal Elemental 95%

Organic 95%

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Hours decay before release 100 i

Atmospheric dispersion factor 0.0014 Thyroid dose, rem 19.1 l

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DOSE RATE v ns TIME f

Time, hours Dose rate, Rem /Hr a

0.OOE+00

(

O.00 Min) 4.3CE+03

)

1.OOE-01

(

6.00 Min) 2.49E+03 2.OOE-01

( 12.00 Min) 2.06E+03 3.00E-01

( 18.00 Mtn) 1.84E+03 1

5.OOE-01

( 30.00 Min) 1.58E+03 l

7.OOE-01

( 42.00 Min) 1.42E+03 J

1.OOE400

(

1.QD Hours) 1.27E+03 2 OOE+00

(

2.00 Hours) 9.84E+02 3.OOE+00

(

3.00 Hours) 8.09E+02 5.QOE+00

(

5.00 Hours) 5.83E+02

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7.OOE+00

(

7.00 Hours) 4.31E+02 1.OOE+01

( 10.00 Hours) 3.40E+02 2.OOE+01

( 20.00 Hours) 2.OSE+02 y

2.40E+01

(

1.00 Days) 1.86E+02 7.OOE+01 i

1.25 Days) 1.64E+02 5.OOE+01

(

2.08 Days) 1.27E+02 7.OOE+01

(

2.92 Days) 1.10E+02 1.OOE+02

(

4.17 Days) 9.18E+01 1.50E+02

(

6.25 Days) 6.94E+01 1

2.OOE+02

(

1.19 Weekt) 5.26E+01 2.SOE+02

(

1.49 Weeks) 4.OOE+01 3.QOE+02

(

1.79 Weeks) 3.04E+01

]

5.00E+02

(

2.98 Weeks) 1.02E+01 7.OOE+02

(

4.17 Weeks) 3.48E+00 1.00E+03

(

1.37 Month) 7.26E-Oi l

)

1.SCE+03

(

2.08 Month) 7.51E-02 2.OOE+03

(

2.78 Month) 1.98E-02 2.50E+03

(

3.47 Month) 1.22E-02 3.OOE+03

(

4.17 Month) 1.04E-02 5.OOE+03

(

6.94 Month) 9.64E-03 7.OOE+03

(

9.72 Month) 9.50E-03 1.OOE+04

(

1.14 Years) 9.29E-03 1.50E+04

(

1.71 Years) 8.95E-03 Gamma dose rate 35 feet from noble gases from 3.1 MWt.

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GAMMA' AND I OD Il4E' TIME, HR GAMMA DOSE IODINE ~ DOSE RATIO

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mrom/ hour

' rem' thyroid O.OOE+00 4.30E+06

-8.22E+02 5.23E+03 i

1.OOE 2.49E+06 8.20E+02 3.04E+03 j

.2.00E-01 2.06E+06 8.19E+02 2.52E+03 J

13.OCE-01 1.94E+06-8.17E+02 2.25E+03 5.00E-01 1.5GE+06 8.14E+02 1.94E+03

)

7,00E-01

.1.42E+06 8.10E+02 1.76E+03 l

1.00E+00-11.27E+06-8.06E+02 1.57E+03' 2.OGE+00 9,.84E+05 7.92E+02 1.24E+03

.3.-00E+00

.8.09E+05 7.78E+02 1.04E+03 i

5.00E+00 5.83E+05 7.54E+02 7.74E+02 7.00E+00 4.51E+05 7.32E+02 6.'16E+02 1.'00E+01 3.40E+05 7.03E+02 4.83E+02

' 2. 00E+01-2.08E+05 6.26E+02 3.32E+02 j

2.40E+01-

>1.86E+05 6.02E+02

.3.09E+02 3.OGE+01.

1.64E+05 5.71E+02 2.87E+02 1

3.00E+01 1.27E+05 4.93E+02 2.58E+02 7.00E+01 1.10E+05 4.40E+02 2.50E402 1.00E+02 9.18E+04 3.82E+02 2.40E+02 1

1.50E+02 6.94E+04 3.13E+02 2.21E+02 2.00E+v2

'5.26E+04 2.60E+02 2.02E+02 2.50E+02 4.00E+04 2.17E+02 1.84E+02 3.00E+02 3.04E+04 1.81E+02 1.68E+02 5.00E+02 1.02E+04 8.82E+01 1.16E+02 i

7.00E+02 3.48E+03 4.29E+01 S.11E+01 1.00E+03

-7.2eE+02 1.46E+01 4.97E+01 1.50E+03 7.51E+01 2.42E+00 3.10E+01-2.00E+03 1.98E+01 4.02E-01 4.92E*01 2.50E+03-1.22E+01 6.66E-02 1.52E+02 3.00E+03 1.04E+01 1.11E-02 9.44E+02 i

5;OOE+03 9.64E+00 3.36E-06 1.15E+06 1

7.00E+03 9.50E+00 6.33E-09 1.50E+09 1.00E+04 9.29E+00 1.32E-13 7.05E+13 1.50E+04 8.95E+00 2.07E-21 4.31E+21 Half lives are from DOE / TIC-11026 Yields are from NEDO-12154-1 (1974)

Iodine dose factors are from Regulatory Guide 1.109 Gamma dose rate constants are from ORNL-RSIC-0045 XOO is.0014 seconds per cubic meter Gamme dese rate is at 35 feet lodsne penetration of the charcoal is 100 %

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