ML20041D919

Forwards Draft FSAR Response to TMI Action Plan Item II.B.2, Including Dose Analysis for Vital Areas & Design & Implementation of Necessary Plant Changes to Reduce Doses. Info Sufficient to Close Out SER Confirmatory Item 40

ML20041D919
Person / Time
Site:	Clinton
Issue date:	03/03/1982
From:	Wuller G ILLINOIS POWER CO.
To:	John Miller Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-0853, RTR-NUREG-737, RTR-NUREG-853, TASK-2.B.2, TASK-TM L30-82(03-03)-6, L30-82(3-3)-6, U-0432, U-432, NUDOCS 8203090540
Download: ML20041D919 (10)

Text

.

ILL/NOIS POWER COMPANY gn3_93y_g l

i 33 g3 500 SOUTH 27TH STREET, DECATIm "'OtS G2525 March 3, 1982 7 4 -

Mr. James R. Miller, Chief Standardization & Special Projects Branch -

gCSN p. C Division of Licensing  ! p 09 r Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission 7,*

8 ansE p ig gps jkS a

Washington, D. C. 20555 s

Dear Mr. Miller:

4 C

Reference:

IP letter 12/1/81 (U-0358) J. D. Geier, IP to J. R. Miller, NRC.

Clinton Power Station Unit 1 f Docket No. 50-461 .. '

During meetings between NRC staff and Illinois Power Company representatives on December 1, 1981, to resolve issues for the Clinton Power Station SER, IP provided the radiation zone maps and identification of vital areas required by TMI Action Plan Item II.B.2 (referenced letter). The remaining in-formation required to close out this issue consisted of the dose analysis for the vital areas and the design and implementation of necessary plant changes to reduce doses. This information is now available and has been incorporated in the attached draft Clinton Power Station FSAR response to TMI Action Plan Item II.B.2.

It is believed that the information provided will be sufficient to close out Confirmatory Item #40 in Clinton Power Station SER(NUREG-0853).

Sincerely,

/ .

G. E. Wuller Supervisor - Licensing Nuclear Station Engineering DLH:mr Attachments cc: J. H. Williams, NRC Clinton Project Manager H. H. Livermore, NRC Resident Inspector M. LaMastra, NRC Radiological Assessment Branch (yool 8203090540 820303 PDR ADOCK 05000461 E pop hl!l

CPS-FSAR AMENDMENT 14 MARCH 1982 NRC ACTION PLAN (NUREG-0660 as clarified by NUREG-0737)

II.B.2 Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which May Be Used in Postaccident Operations NRC Position With the assumption of a postaccident release of radioactivity equivalent to that described in Regulatory Guide 1.3 and 1.4 (i.e., the equivalent of 50% of the core radioiodine, 100% of the core noble gas inventory, and 1% of the core solids are contained in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the control room, radwaste control stations, emergency power supplies, motor con-trol centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operation of these systems.

Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or postaccident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.

CPS Response A review of the design of shielding has been performed for the CPS in accordance with NUREG-0737, II.B.2. The radiation qualification of the safety-related equipment is discussed in Section 3.11.

! Accident Scenario The accident that forms the design basis for this review consists of damage to the fuel, resulting in the release of fission pro-ducts from ruptured fuel cladding. The cause or sequence of events leading to this condition is not strictly defined, but it is postulated that the accident can take place as a result of either a large pipe break or a loss of cooling water without a i break in a major pipe. For easy reference, the former type of accident can be called a line break accident and the latter a no-

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4 CPS-FSAR AMENDMENT 7 SEPTEMBER 1981 line-break accident. The distribution of fission products in

( various systems depends upon the accident type.

Line-Break Accident In this type of accident, the reactor coolant pressure boundary is ruptured, and the fission products are released immediately to the drywell and the suppression pool. The fission products are also instantly released to the primary containment atmosphere outside of the drywell if the fuel damage precedes the drywell pressure blowdown. Otherwise, the fission products initially stay in the drywell and are released to the containment over a longer period of time through the combustible gas control system and the suppression pool.

No Line-Break Accident It is assumed that the reactor coolant pressure boundary is intact, cooling.

yet fuel cladding is damaged due to the loss of adequate It would appear that in such a case all the released fission products are confined to the reactor coolant and the steam in the reactor vessel dome. NRC has stipalated that this assumption be made. However, such a conditior is not very probable in a BWR. Loss of adequate cooling ts expected to increase the pressure in the vessel, which is relieved through the safety-relief valve operation. Thus, most of the released fission products are expected to be dumped into the suppression k' pool in a short time.

If the fission products are assumed to be confined to the vessel, the systems affected by this assumption will be the RHR, the postaccident sampling and the steam side of the.RCIC.

Radiation Source Assumotions The radioactive nuclides released from the core due to the accident are distributed into the reactor coolant, suppression pool water and the air in the drywell and the primary containment. They are then carried to different areas and components by the systems which are put into operation after the accident. The components that receive these isotopes and the pipes that carry them can be treated as individual sources of radiation. They are located in various parts of the station, and thus affect the radiation environment throughou; the station.

Certain assumptions are made in order to quantify these sources.

The radiation sources for this report were calculated based on the guidance of NUREG-0737, and upon the accident scenarios discussed above. The sources were calculated as a function of time, with due accounting for the radioactive decay and migration of nuclides. The calculations were performed using the computer codes RACER and RUNT (References 22 and 23). The basic k

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4 CPS-FSAR AMENDMENT 14 MARCH 1982 assumptions made and the parameters used in the calculations are listed in Table D-1.

Systems in Postaccident Use Clinton Power Station systems that can be used in postaccident operations and which affect the radiation levels in the station are the High Pressure Core Spray, Low Pressure Core Spray, Residual Heat Removal, Reactor Core Isolation Cooling, Main Steam Isolation Valve Leakage Control, Standby Gas Treatment, Combustible Gas Control, Control Room HVAC, Sampling, Radiation Monitoring, and Floor Drain and Equipment Drain Systems. These systems are described in detail in various sections of the FSAR.

Vital Areas The areas of the station where access will be required following an accident have been identified as the control room, technical support center (TSC) and the sampling station. The safety systems and components are designed to have redundancy, and to be operated remotely so that access near their locations will not be required.

Further, the safety components have been qualified to withstand the radiation environment that they will be subjected to in their respective locations. The environmental zones and compor ent qualification are discussed in Section 3.11.

Vital Area Dose Analysis Radiation doses to personnel occupying the vital areas have been calculated based upon the considerations of postaccident radiation sources and the occupancy requirements.

a. Control Room The control room dose analysis has been presented in Section 15.6.5. The control room shielding design and ventilation system design are discussed in Section 6.4. No additional shielding or other protective features are required to meet the dose criteria of NUREG-0737.  ;

b. Technical Support Center Technical support center has been established in the same protected envelope as the control room. Hence, the occupancy doses in the TSC are the same as those in the control room.

c. Sampling Station ,

l  ! ;

In order to collect samples of postaccident fluids, a l

, postaccident sampling system is being installed at CPS.

l This sytem is discussed in Section II.B.3 of this appendix. j i

D-33 i

CPS-FSAR AMENDMENT 14 MARCH 1982 The postaccident sampling panel is being located in the Control /

Diesel Generator building. The panel location has been chosen to enable routing of the sample lines through shielded pipe tunnels, to provide easy and safe access from the control room and TSC, and to enable convenient transport of the samples to the labora-tories. The chosen location of the sampling panel is in the sane building as the control room and TSC and on the same floor as the laboratores. The absence of any postaccident sources betueen the control room /TSC and the sampling station ensures that the access path to the sampling panel will have low radiation background.

The panel is equipped with lead and steel shielding in the front.

Concrete walls and ceiling are being added around the panel as well for shielding. The sample and return lines will be routed through shielded pipe tunnels. Details of the sampling station chielding are shown in Figure D-3.

Using the highest anticipated radiation sources, as described in Table D-1, item 2, at a time instant of 1 hr. after the onset of the accident, the dose rate at 3 ft. from the front of the panel has been conservatively calculated to be approximately 500 mrem /hr. The total time required for taking a liquid sample is estimated to be 32 minutes, starting with purging of the in-coming lines to backflushing of the panel after taking a sampic.

The dose rate in front of the panel will vary with the sampling step that is underway, with the maximum of 500 mrem /hr value e::pected when the tubes in the panel are filled with undiluted reactor coolant. The integrated dose to the operator while taking the highest radioactivity sample is estimated to be approximately 100 mrem. The integrated dose while taking other types of camples and/or while taking samples later on in the course of the accident is expected to be much smaller. The sample is taken directly into a' shielded cask which is designed to minimize dose to the operator while transporting the sample.

The sampling panel is also equipped to perform on-line radioiso-topic analysis. Use of the on-line analysis will further reduce overall dose to the operator.

Radiation Qualification of Safety Related Eauipment Radiation Qualification of Safety-related equipment is an integral part of the environmental equipment qualification, program, which is addressed in FSAR Section 3.11.

Design Modifications

, As a result of the postaccident radiation and shielding design l review, the following design modifications are being implemented to reduce radiation doses:

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. CPS-FSAR AMENDMENT 14 MARCH 1982

a. The shielding design review has indicated that the only shielding design modification required is the addition of shiciding around the postaccident sampling panel, which is being implemented, as discussed above.

b. One of the significant contributors to radiation sources in the secondary containment was found to be the exhaust of the MSIV leakage control system which was routed into an RHR cubicle. Routing this exhaust to a suction header of the SGTS via a pipe connection would eliminate this source contributor from the secondary containment, and would significantly improve the postaccident radiation environment there. A design modification is being implemented to accom-plish the above.

l D-33b

CPS-FSAR AMENDMENT 16 MARCH 1982 TABLE D-1 RADIOACTIVE SOURCE ASSUMPTIONS Parameters and Assumptions Source Medium Used in Source Calculations

1. Reactor Core o Power level - 2,894 MWt o Fuel irradiation time - 3 years o Thggmal,peutron 1 flux - 3.97 x 10 cm .sec

2. Reactor Coolant o 100% noble gases *, 50% halogens, 1% solids mixed uniformly in the reac5 7,520 ft rc nt v lum nt f

3. Suppression Pool o 0% noble gases, 50% halogens, 1% solids mixed uniformly in the reactor coolant plus suppressiog pool volume of 144,300 ft

4. Drywell Air o 100% noble gases, 25% halogens mixed uniformly in dgywell volume of 246,500 ft

5. Primary Containment Air o 100% noble gases, 25% halogens mixed uniformly in drywell plus containgent volume of 1,550,800 ft

6. Secondary Containment o Secondary cogtainment volume -

Air 1,981,000 ft o 0.65%/ day primary containment leak o 30 sefh/line MSIV leak o 1500 gal leak from ECCS o 4000 cfm exhaust via SGTS (a) o 100% mixing, for source concen-trations in the secondary con-tainment air (b) o 0% mixing, for releases from the secondary containment

7. SGTS Filter o Flow rate - 4,000 cfm o 100% particulate, 99% iodine removal efficiency o Sources in air as in 6 (b) above D-34

- __ __. -~.

CPS-FSAR AMENDMENT 7 SEPTEMBER 1981 f_ Table D-1, Cont.

, Parameters and Assumptions i Source Medium Used in Source Calculations

8. Plume

• Releases from the SGTS, with parameters as in Item 7 above

.!

• Complete mixing in the wake of the containment building i *1 m/sec wind speed

9. Control Room o Releases per Item 7 above HVAC Filter

• x/Q as given in FSAR

• Dual, separated air intake locations

• Flow rate - 3,000 cfm

• 100% particulate, 99% iodine removal efficiency i

• Fission product activities are listed as percentages of ccre inventory.

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l n L I l SECTION "A- A' PA S SYS T E M Page D-35b

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