ML20127A697
| ML20127A697 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/17/1985 |
| From: | Wilson R GENERAL PUBLIC UTILITIES CORP. |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-3.D.3.4, TASK-TM NUDOCS 8506210228 | |
| Download: ML20127A697 (5) | |
Text
GPU Nuclear Corporation NNOIM7 loo lnterpace Parkway Parsippany, New Jersey 07054-1149 (201)263-6500 TELEX 136-482 Writer's Direct Dial Number:
June 17, 1985 Mr. John A. Zwolinski, Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Zwolinski:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Control Room Habitability (NUREG-0737 Item III.D.3.4)
Results of Whole Body and Beta Skin Dose Analysis By letter dated June 4, 1985, GPU Nuclear Corporation committed to submit to the NRC staff the results of calculations and analysis, as well as the assumptions and models used in the analysis for whole body and beta skin doses using Regulatory Guide 1.3 source term.
Enclosed is a summary which provides the general methodology, assumptions, data, and results from the calculations performed by Stone & Webster Engineering Corporation (SWEC) to determine the 30-day integrated gamma and beta doses to the control room operators following a design basis LOCA. This enclosure fulfills the requirements contained in item #8 of Attachment I for the Interim System Upgrades for Control Room Habitability at the Oyster Creek Nuclear Generating Station as stated in our June 4, 1985 submittal.
The calculated doses are below the SRP 6.4 limits of 5 rem (gamma) and 30 rem (beta) as required for compliance with Item III.D.3.4 of NUREG-0737. As agreed during the March 19, 1985 meeting with the NRC, the thyroid dose will not be addressed at this time. The NRC is currently evaluating recent industry data on the LOCA iodine source term in order to determine the requirements for licensee response. The thyroid dose will be addressed after completion of this NRC review and independent of the commitments contained in our June 4, 1985 submittal.
ADOCK050g29 ghh 8506210228 850617
/
DR GPU Nuclear Corporation is a subsidiary of General Public Utihties Corporation
. If yta have any questions, please contact M. W. Laggart, Manager, BWR Licensing at-(201) 299-2341.
Ve tr'aly yours, In Director Technical Functions 1r cc: Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pa.
19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N. J.
08731 1943f
ENCLOSURE
SUMMARY
OF POST-LOCA CONTROL ROOM RADIOLOGICAL HABITABILITY 30 DAY GAMMA AND BETA DOSES GENERAL METHODOLOGY Stone and Webster Engineering Corporation (SWEC)%/Qs, leak rates, an computer code DRAGON 4 is used along with the appropriate HVAC operating data, Regulatory Guide 1.3 source terms to determine the 30-day integrated gamma and beta doses in the control room due to the airborne source within the control room HVAC pressure envelope.
The gamma dose contribution from the radioactive cloud surrounding the control room is calculated using the isotopic concentrations versus time outside the control room and the appropriate dose rate converstion factors for an infinite cloud source model. One-foot thick concrete shielding is assumed.
The gamma dose contribution from post-LOCA piping sources and airborne radioactivity in the reactor building is taken from the September 1984 report for OCNGS on NUREG-0737, Item II.B.2 compliance.
The internal and external airborne source doses are reduced to account for less than 100 percent control room occupancy (see Assumptions and Data, Item 13). The piping and containment shine dose is not reduced for occupancy.
This is conservative.
The three' airborne fission product release paths considered are:
1.
MSIV Bypass Leakage (SRP 6.2.3 - Branch Technical Position CSB 6-3).
2.
Containment Leakage (Appendix A to SRP 15.6.5).
3.
ESF Leakage (Appendix B to SRP 15.6.5).
ASSUMPTIONS AND DATA 1.
The reactor is operating at 102-percent rated thermal power at the time of the accident (reference SRP 15.6.5).
2.
Twenty-five percent of the core halogen inventory and 100 percent of the core noble gas inventory becomes airborne and immediately(reference available for release from primary containment at the time of the LOCA Regulatory Guide 1.3).
3.
Fifty percent of the core halogen inventory is released to the water being circulated through the ESF equipment in the reactor building (reference SRP 15.6.5).
4.
One of the four MSIVs is assumed to have failed in the open position "
the time of the LOCA.
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Natural phenomena are not considered as the initiating event. Therefore, all plant structures are assumed to remain intact for the duration of the.
accident (in accordance with NRC Minutes, dated April 16, 1985, of-the March 19, 1985 meeting held at the Oyster Creek Nuclear Generating Station).
- 6. -
Upon MSIV closure from the LOCA signal, the steam supply to the steam jet air ejectors is cut off. This causes the vacuum to be lost in the turbine / condensers. Therefore, the steam leaking through the MSIVs
- travels through the main steam piping and escapes to the turbine building. Due to buoyancy, the steam rises to the top of the turbine building for release to the environment. No credit is taken for holdup
.or dilution of the. activity in the turbine building.
7.
A delayed release credit is taken to account for the time it takes the MSIV leakage to travel from the outboard MSIV to the turbine / condensers.
The delay times calculated are: 8.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> for the line with single isolation and 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> for the line with double isolation.
- 8.
The worst meterological conditions are assumed to occur at the onset of the MSIV leakage release to the environment (i.e., the 0 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7/Q is applied for the 8.7 to 16.7 hour8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> time period, etc.). Refer to assumption 19.
9.
' Primary containment is assumed to leak at its technical specification limit of 1.0 percent per day for the duration of the accident (reference
' technical specifications page 4.5-1).
-10.
The ESF equipment in the reactor building is assumed to leak at a total rate of I gpm for the duration of the accident (in accordance with the NRC assumption - see Crutchfield/Fiedler letter on SEP Topic XV-19, dated
. September 2, 1982).
. _1 1. No credit is taken for hold up or dilution of containment leakage or ESF leakage in the reactor building. However, credit is taken for 90 percent halogen filtration by the SGTS (see Crutchfield/Fiedler letter dated September 2, 1982).
~
'12.
In response to a LOCA signal, the control room operator will switch the control room HVAC system to.the recirculation mode with the outside air intake flow rate at the minimum required to maintain a positive pressure within the HVAC envelope (450 cfm was used - to be confirmed by test).
The HVAC system is assumed to remain in this mode of operation for the duration of the accident.
F
- 13. The crew of operators in the control room at the time of the accident is assumed to remain there for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A four crew, eight hour shift rotation is assumed thereafter.
- 14. Drywell free volume = 180,000 ft.3 (FSAR Table 6.2-2).
- 15. Torus minimum water volume = 82,000 ft.3 (technical specifications page 3.5-1).
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- 16. One hundred-percent rated thermal power = 1930 MWt (technical specifications page 1.0-3).
- 17. The gross volume-served by the control room HVAC system is calculated to be 64,192 ft.3 This volume is reduced by 20 percent to account for furniture, equipment, people, etc., to yield a net control room envelope volume = 51,354 ft.J.
- 18. The core inventory'is calculated using the shutdown isotopic activities generated in GE Computer Run No. SNUMB=7007S dated November 1979.
T
- 19. The following /Qs were used in this analysis:
-Time Period Stack A/Q (Note 1)
(sec/mjneduilding%
.Turb
/Q (Note 2) 3 (hr)
(sec/m )
)
0-8 1.80.x 10-4 5.18 x 10-3 8-24 9.67 x 10-5 3.94 x 10-3 24-96 2.50 x 10-5 2.75 x 10-3 96-720 3.60 x 10-6 1.66 x 10-3 Notes:
1.
Containment leakage and ESF leakage 2.
MSIV bypass leakage These % s are calculated using Murphy and Campe methodology with
/Q site-specific meteorological data. Occupancy factors are not incorporated into these values.
20.~
Although the MSIVs are designed to provide a leak-tight barrier, it is recognized that some leakage through the valves will occur. The typical BWR Technical Specification allowable MSIV leakage is 11.5 scfh.
However, a value of 11.9 scfh has been calculated for Oyster Creek Nuclear Generating Station based upon its Technical Specification test pressure of 20 psig. Therefore, this more conservative value for MSIV leakage of 11.9 scfh was utilized in the Oyster Creek dose analysis.
RESULTS 30-day integrated gamma dose = 2.61 rem (limit = 5 rem) 30-day integrated beta dose = 25.8 rem (limit = 30 rem)
The calculated doses have margins of safety of 48 percent for the gamma dose and 14 percent for the beta dose with respect to the SRP 6.4 limits. Therefore, OCNGS complies with Item III.D.3.4 of NUREG-0737 for gamma and beta doses.
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