ML20210P132

Forwards Rev 0 to Calculation BYR97-332, CR Dose for MSLB Accident, Per 970806 Telcon W/Nrc Re Licensee 970131 Amend Application for Rev to TS on Reduction in Dose Equivalent Iodine

ML20210P132
Person / Time
Site:	Byron
Issue date:	08/22/1997
From:	Hosmer J COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20210P137	List: ML20210P132 ML20210P139
References
NUDOCS 9708260449
Download: ML20210P132 (1)

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1400 Opm 11x e i I Am nen Gnnt. Il Wt t voi August 22,1997 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D C. 20555

Subject:

Additional Information Pertaining to the Technical Speci0 cation Amendment for the Reduction in in Dose Equivalent lodme flyron Units I and 2 J NRC3nckcLNumbers: 50 454/4g

References:

1) J. Ilosmer letter to the Nuclear Regulatory Commission dated January 31, 1997 transmitting Technical Specification Amendment Request.

2) Teleconference dated August 6,1997, between the Commomvealth Edison Company and the Nuclear Regulatorv Commission.

3) J. Ilosmer letter to the Nucleai Regulatory Commission dated August 21, 1997, transmitting Additional Information for the Reduction in Dose Equivalent hxline The referenced letter transmitted the Commonwealth Edison Company (Comed) request to amend the Technical Specification for the reactor coolant system Dose Equivalent lodine 131 for Byron Unit 1. Subsequent to that traitsmittal, teleconferences were held to clarify information pertaining to the dose commitment in the unlikely event of a main steam line bre.ak. In the referenced teleconference, the Nuclear Regulatory Commission requested the dose calculations for the control room, exclusion arca boundary and low population zone Reference 3) transmitted the exclusion area boundary and low population zonc calculations. Attached is the final control room calculation Comed welcomes meeting with the Staff on August 26th to further discuss this calculation along with any additional issues related to the Technical Specification Amendment.

If you have any questions, please contact this office Sinccrcl- 1 1 y '

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7 . QG \ I John it ilosmer Engineering Vice PresiGnt Attadiment ec: A. Deach, Regional Administrator - Rlli G. Dick, D) ton Project Manager NRR S. Burgess, Senior Resident inspector - Dyron

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.' Eshibit D NEP.13 02 Revisiou 5 COMMONWEALTH EDISON COMPANY CALCULATION TABLE OF CONTENTS PROJECT NO. 9044457 CALCULATION NO. BYR97 332 REV,NO.O PAGE NO. 3 SECTION PAGE NO. SUB PAGE NO.

TITLE PAGE /

REVISION

SUMMARY

TAllLE OF CONTENTS 3 PURPOSIUOBJECTIVE 4 4

METil0DOLOGY AND ACCEPTANCE CRITERIA ASSUMPTIONS 4 DESIGN INPUT-4 REFERENCES CALCULATIONS 5

SUMMARY

AND CONCLUSIONS 7

A*ITACllMENTS A UFSAR Table 6.4.1 4, y B UFSAR Figurc 7.21 gj C UFSAR Table 15.6.9 gj

.' Exhibit E NEP-12-02 Revision 5 COMMONWEALTil EDISON COMPANY l CALCULATION NO. : ilYR97 332 PROJECT NG. 903057 PAGE NO. 4 l PURPOSE / OBJECTIVE:

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i The purpose of this calculation is to assess the post accident radiological dose to control room i

occupants following a main steam line break (MSLB) accompanied by primary-to secondary coolant leakage.

(_ M! P'9DOLOGY AND ACCEPTANCE CRITERIA:

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4 The approacn is to demonstrate that the consequences of the MSLB accident will be less than tne design basis loss of coolant accident (LOCA) assessment and are, therefore, bounded by the assessment included in the Updated Final Safety Analysis Report (UFSAR).

ASSUMP flONS:

The control room ventilation is assumed to be in operation following the MSLB. The basis for this assumption is discussed in CALCULATIONS section. ,

DESIGN INPUT:

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The iodine releases and the maximum acceptable steam generator leak rates are from calculation ATD 0410, Rev.1 (Reference 1). The releases from the containment an(, the resulting control room dose are from the UFSAR (References 3 and 7).

REFERENCES:

. 1. Comed Calculation ATD-0410, Rev.1,"Aliowable Leakrate Calculation for Steam Generator Interim Plugging Criteria," August 21,1997.

2. USNRC, Standard Review Plan WREG-0800, Section 15.6.3," Radiological Consequences of Steam Generator Tube Failure (hVR),"Section II, Acceptance Criteria, Rev. 2, July 1981.

3. Byron UFSAR, Table 15.6-9 (copy attached)

4. USAEC Technical Information Doct. ment TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites, March 1962.

5. Sargent & Lundy Calculation BY-VC-07, Rev. O, " Radiation Habitability for the Control Room -

Byron," October 3,1985.

' 6. Byron UFS AR, Figure 7.2-1 (Sheet 8) (copy attached) 7 Byron UFSAR, Table 6.4-1 (copy attached)

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l REVIE ON NO.: 0- _

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l Exhibit D NEP-12-02 Resision 5 COMMONWEALTII EDISON COMPANY l CALCULATION NO. i BYR97 332 PROJECT NO. 9044-057 PAGE NO. 5 ]

CALCULATIONS:

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Comparison of Releases:

In order to determine control room doses for he MSLB case from doses calculated for the LOCA case.

the radionuclide releases in each case must first be identified. Since the thyroid dose is the governing limit, releases of radioactive iodines will be compared.

) e MSLB Release Reference I considers three release sources:

a. Primary coolant with a pre-accident iodme spike

• b. Primary coolant with a post-accident iodine spike

c. Secondary coolant with iodine concentrations at a predetermined limit Reference 1, then, evaluates these to obtain a limiting primary to secondary leakrate as follows. For a one gallon per minute primary to secondary leak, we have:

Dose Equivalent Exclusion Area Source 1-131 Release Boundary Does a 15.9 Ci 3.40 rem b 10.6 Ci 2.25 rem  :

c 1.58 Ci 1.14 rem l NRC acceptance criteria considers the following combinations of sources and acceptance criteria:

Source Acceptable Dose Basis, Reference 2 at EAB a+c 300 rem 10CFR100 limit b+c 30 rem 10% of 10CFR100 limit Using this criteria, Reference 1 calculates a limiting primery to swondary leak rate of 12.8 gallons per minute (Scenario b+c).

l REVISION NO.: 0 l

, ,. .' Exhibit D NEP-12-02 Revision 5 COMMONWEALTH EDISON COMPANY l CALCULATION NO. : BYR97 332 PROJECT NO. 9044-057 PAGENO. 6 l Using this leakrate, the iodine releases for the difl'erent sources are as follows:

Source Dose Equivalent 1-131 Release a 204 Ci b 136 Ci c 1.6 Ci 7

Under these conditions, the worst case combination of(a+c) or (b+c) is (a+c), which results in a release of 206 curies of dose equivalent 1-131,

• LOCA Release The releases of radioactive iodines in the design LOCA case are given in Reference 3. These may be convened to dose equivalent I-131. In order to compare fairly with the control room dose assessment (because oflong term X/Q and ether parameter variations), consider only the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of release (where all control room habitability parameters are constant).

Release, Curies kem per Curie 0-2hr 2-8 Total Dose Total X hr Conversion DCF Factor (Ref 4) 1-131 214 545 759 1.48E6 1.12 E9-I-132 257 217 474 5.35E4 2.54E7 l-133- 469 1060 1529 4.00E5 6.12E8 l-134 320 62 382 2.50E4 9.56E6 1-135 401 684 1085 1.24E5 1.35E8 ,

Sum 1.90E9 l The total dose equis alent I-131 release is then 1.90E9/1.48E6 or 1290 curies. Note that these are ICRP-2 dose conversion factors rather than the ICRP-30 dose conversion factors used for the MSLB dose calculation. This is done to be consistent with the LOCA control room dose, which also uses ICRP-2 dose conversion factors, l REVISION NO.: 0 l 1

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l l Eshibit D NEP 12-02 Resision 5 COMMONWEALTil EDISON COMPANY l CALCULATION NO. : BYR97-332 PROJECT NO. 9044-057 PAGE NO. 7 l T

Evaluation of Control Room Ventilation System Configuration:

The proposed control room dose estimation is valid only if the control room emergency filtration system (VC system) is in op : ration. The VC system is assumed to be in operation for the LOCA assessment (Reference 5). A review of applicable functional diagrams (Reference 6) shows that, in the event of a low steam line pressure, the VC system is activated. His assures that the proposed dose estimation is valid.

Estimate of Control Room Dose:

Reference 7 gives the control room dose calculated for a LOCA. This calcalation is documented in Reference 5.

From intermediate results included in Reference 5, one can obtain the cumulative dose for the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the LOCA case is 8.27 rem. His would be the resulting dose from the 1,290 curies of dose equivalent I 131 released in 8 hc. irs as noted above. A normalized dose for this period can be calculated as:

8.27 rem /1290 Ci = 6.41E-3 rem /Ci 1 131 equivalent Using the above LOCA results, then, one can say that, all other things being equal, the control room dose for the MSLB is in proportional to the radiciodine release and would be:

206 Ci x 6.41E-3 rem /Ci = 1.3 rem Since the ICRP 30 dose conversion factors are smaller than the ICRP-2 dose conversion factors, use of the normalized dose based on ICRP 2 is conservative.

SUMMARY

AND CONCLUSION:

The post accident control room dose due to the MSLB accident is less than that due to the design basis LOCA already included in the UFSAR (Reference 7).

FINAL l REVISION NO.: 0 l

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