ML20203L887

Forwards Interim Operation Plan for Auxiliary Bldg Ventilation Sys,Justification for Use of Plan,Results of Offsite & Control Room Dose Analyses Assuming Ventilation Sys Inoperable & Info Re FSAR & SER Dose Calculations

ML20203L887
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	08/26/1986
From:	Miosi A COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20203L889	List: ML20203L887 ML20203L954
References
2034K, NUDOCS 8609020031
Download: ML20203L887 (13)

Text

.

L~ Coenrnonwesitts Edison

- One First Nemonei Plaza. Chicago, luinois

!- Address Reply to: Poet Omco Box 757 Chicago,luinois 60800 076) .

r August 26, 1986 Mr. Harold R. Denton U.S. Nuclear _ Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC. 20555

Subject:

Braidwood Station Units 1 & 2 Interim Operation Plan for Auxiliary Building Ventilation System o NRC Docket Nos. 50-456 and 50-457

Reference:

April 1, 1986 A.D. Miosi letter to H.R. Denton

Dear Mr. Denton:

This letter transmits the interim operation plan for the Auxiliary Building Ventilation (VA) System at Braidwood. This plan is required since the VA system is common to Braidwood Units 1 and

2. Only portions of the system serving Unit 1 and common areas of the Auxiliary Building and the Fuel Handling Building are required to support Unit 1 operation during Unit 2 construction. Also, ventilation requirements for these buildings will be minimal at fuel load and will increase as Unit 1 operation approaches full power level. The VA System interim operation plan ensures adequate ventilation for the Auxiliary and Fuel Handling Buildings until Unit 2 construction is completed. At that time the entire VA System will become operable.

The VA System serves the Auxiliary and Fuel Handling Buildings by filtering radioactive contaminants from air exhausted from the buildings to ensure that offsite and control room radiation doses are within acceptable limits. The system provides balanced .

supply and exhaust airflows throughout the buildings to maintain them at negative pressures with respect to atmosphere. The system also cools the Unit 1 and Unit 2 engineered safety features (ESP) equipment cubicles as well as general areas of the buildings that are shared by both units.

Radiation levels inside the Auxiliary Building are directly proportional to the reactor's thermal power level and the Emergency Core Cooling System (ECCS) equipment leak rate. Prior to initial criticality and during plant operation at low power, high radiation levels will not exist since the production of fission products does not occur or is minimal. During higher power operation, radiation levels will increase if significant ECCS leakage occurs in the 8609020031 860826 i PDR ADOCK 05000456 V g(g A PDR t.- - - _ - - - - - _ _ - - _ . _ _ _ _ _ _ __

Auxiliary Building. In this situation the VA high efficiency particulate adsorber (HEPA) and charcoal filters are used to ensure that offsite and control room doses are within the limits specified respectively in 10CFR100 and General Design Criterion 19 of 10CFR50, Appendix A. Also, the building is maintained at negative pressure

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with respect to atmosphere to avoid uncontrolled, unmonitored radioactive releases.

The need for exhaust air filtration and negative pressure in the Nuxiliary Building at high power levels can be eliminated by reducing the ECCS equipment leak rate. Offsite and control room doses have been analyzed assuming the TUL System is inoperable (no HEPA and charcoal filtration). Maximum allowable reactor power levels have been determined for varying-ECCS equipment leak rates so that offsite and control room doses limits are not violated. The results of the analysis are shown in Figure 1 of the interim operation plan. This figure defines the operating limits for Braidwood Unit 1 without HEPA and charcoal filtration and negative pressure in the Auxiliary Building. Braidwood Unit 1 will operate within these limits until preoperational testing of the VA System is completed. At that time these limits will no longer apply.

Radioactive releases from the Fuel Handling Building will not occur without spent fuel in the fuel pool. Spent fuel will be present prior to Unit 2 completion only if the Unit 1 core is unloaded. In that situation HEPA and charcoal filtration and negative pressure is required in the' Fuel Handling Building.

Consequently, until Unit 2 completion, the Fuel Handling Building will be maintained at a negative prossure only if the Unit 1 core is unloaded. The Fuel Handling HEPA and charcoal filters will be tested concurrent with the Auxiliary Building filters unless the Unit 1 core is unloaded prior to the scheduled test completion.

The VA system interim operation plan will ensure that abnormal temperatures will not exist in the Auxiliary and Fuel Handling Buildings during the startup of Braidwood Unit 1. Prior to fuel-load, all Unit 1 cubicle coolers will become operable to ensure proper operation of ESF equipment. Prior to initial criticality, l the VA Main supply and exhaust fans will be available and Unit 1 and l common ductwork will be installed to distribute airflow throughout j the general areas of the buildings, if necessary. Airflow balancing of the Unit 1 and cor. mon ductwork will be completed during Unit 1 startup.

Justification for use of the interim operation plan is included as attachments to the plan. Attachment A provides results of offsite and control room dose analyses assuming the VA system is inoperable. Attachment B discusses several significant conservatisms in the conventional FSAR and SER dose calculations.

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Those analyses are shown to be conservative by at least a factor of 10 without consideration of the chemical state of principal radionuclides. This information, coupled with the fact that the VA System could be operated on short notice during the final testing phase, provide reasonable assurance that Braidwood 1.can be operated safely.

NRC concurrence is requested. Please direct any questions you may have to this office.

One signed original and fif. teen copies of this letter and enclosure are provided for your review.

Very truly yotyrs, A. D. Miosi Nuclear Licensing Administrator

/klj cc: J. Stevens 2034K

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j' BRAIDWOOD AUXILIARY BUILDING VENTILATION SYSTEM

!- . . INTERIM OPERATION PLAN 1

1.0 Introduction The Auxiliary Building Ventilation (VA) System at Braidwood is common to Units 1 and 2 serving the Auxiliary Building and the Fuel Handling Building. The system is designed to filter radioactive contaminants from the air exhausted from the buildings to ensure that offsite radiation l, levels are within acceptable limits. The system provides balanced supply and exhaust airflows throughout the buildings to maintain them at

! negative pressures with respect to atmosphere. Operation of the system directly affects radiation levels in the control room since the exhaust air can potentially leak into the control room boundary. The system also

cools the Unit 1 and Unit 2 engineered safety features (ESP) equipment l cubicles as well as general areas of the buildings that are shired by j both units. Proper temperature is maintained in each ESF cubicle by a

! cubicle cooler which operates independently of the VA System's main fans. The general areas of the building are cooled with air provided by the VA main supply fans. A complete description of the system can be found in Section 9.4.5.1 of the Byron /Braidwood Final Safety Analysis Report (FSAR). Preoperational testing require:nents for the system are l provided in TABLE 14.2-35 of the FSAR. These requirements call for the i entire system to be completed and preoperationally tested prior to Unit 1 l

Fuel Load.

The two units at Braidwood will not be completed simultaneously. The current FSAR testing requirements for the Auxiliary Building Ventilation f '

System do not account for this. Since the Auxiliary Building Ventilation j System serves both Unit 1 and Unit 2 areas of the Auxiliary and Fuel i Handling Buildings, the entire system is not required to function until

Unit 2 operation. Only those portions of the system serving Unit 1 and j common areas need to be functional for Unit-1 startup.

Ventilation requirements for the Auxiliary Building ~and Fuel Handling I

Building will vary with each stage of the Unit 1 startup. Radiation levels and temperatures inside the plant will be dependent upon the plant operating mode and reactor power level. Filtration and cooling needs of l Unit 1 and common areas will increase to maintain existing margins of

) safety as the. unit approaches full power operation. At Unit 1 Fuel Load,

! only a minimal portion of the Auxiliary Building Ventilation system needs to operate. Later, at higher power levels, all Unit 1 and common i portions of the system must operate to. satisfy ventilation requirements.

( 2.0 Interim Operatina Plan i

! An interim operation plan for the Auxiliary Building Ventilation System has been established. This plan ensures adequate ventilation for the Auxiliary Building and Fuel Handling Buildings during Unit 1 startup and operation and Unit 2 construction. The plan follows:

i A. Initial Startup Phase

! During the period between reactor fuel load and initial criticality, i fission products will not be produced. Radiation levels inside the l

' Auxiliary Building and Fuel Handling Building will be at natural background levels and there will be essential'ly no risk of releasing radioactive contaminants to the environment. Consequently, there will be no need for HEPA and charcoal filtration of exhaust air at this time. Also the buildings do not need to be maintained at negative pressure,s with respect to atmosphere.

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7 Excessive heat loads will normally not exist in the general _ areas cf the Auxiliary and Fuel Handling Buildings during this period.

However, high temperatures may occur in the engineered safety features (ESP) equipment cubicles when operating the ESF equipment.

Cooling of the ESF equipment cubicles is mandatory, if high temperatures occur.

. To provide the necessary cooling during the initial startup phase, the following cubicle coolers will be operable orior to Unit 1 Fuel Load:

1. Residual Heat Removal Pumps lA and 1B Cubicle Coolers

2. Safety Injection Pumps lA and 1B Cubicle Coolers

, 3. Centrifugal Charging Pumps lA and 1B Cubicle Coolers

4. Essential Service Water Pumps lA and 1B Cubicle Coolers

5. Unit 1 Auxiliary Feedwater Diesel-Driven Pump Cubicle Cooler

6. Containment Spray Pump 1A and 1B Cubicle Coolers

7. Unit 1 Positive Displacement Charging Pump Cubicle Cooler

0. Unit 1 Spent Fuel Pit Pump Room Cubicle Cooler Operation of the coolers listed in Items 1 through 6 will provide the necessary heat removal in any plant operating event. Items 7 and 8 l will also be operational prior to fuel load, although their function is not necessary in an emergency situation.

B. Low Power Operation ,

Beyond initial reactor criticality, fission products are produced at a rate directly proportional to the reactor's thermal power level.

The potential for radioactive releases from the-Auxiliary Building is due mainly to fission products from Emergency Core Cooling System (ECCS) equipment leakage. The potential for radioactive releases from the Fuel Handling Building is dependent upon the presence of spent fuel in the building. Spent fuel will be present only if core l

unloading is required.

i l During normal plant operation at all power levels, ECCS equipment leakage will be minimal since most ECCS equipment will not be operating. Therefore, radiation and airborne radioactivity levels in the Auxiliary Building will be normal and will not affect offsite and control room doses. However, during an accident condition, more ECCS l equipment may operate and leakage may increase. If this condition

occurs, the Auxiliary Building Ventilation System's HEPA and charcoal filters may be used to ensure that offsite and control room doses remain within the limits of the Code of Federal Regulations Title 10, Part 100 (10CFR100) and General Design Criterion 14, Appendix A of 10CFR50.

1 1

4 The radiological consequences of Energency Core Cooling System (ECCS) component leakage in the Auxiliary Building are discussed in Section 15.6.5.6 of the Byron end Braidwood FSAR. This analysis was based on methods described in Regulatory Guide 1.4. and assumes component leakage of approximately 1 gallon per hour. Analys,is of offsit'e and control room doses for Braidwood Station can be'found respectively in  ;

FSAR Sections 15.0 and 6.4. The FSAR dose analysis assumes the l Auxiliary Building Ventilation System to be operation'a1. Expected doses are provided in FSAR Tables 15.0.12 and 6.4.1.

Offsiteland control room doses have been reanalyzed assuming the Auxiliary Building Ventilation System inoperable (i.e. without HEPA ,

and charcoal filtration). Details of this reanalysis are provided in Attachment I entitled "Offsite and Control Room Doses - Braidwood t Station". In the reanalysis the assumed ECCS equipment leakage is  ;

i that required by Standard Review Plan 15.6.5. Appendix B and is

larger than the leakage values used.in the FSAR analyses. Under <

1 these conditions, the reanalysis demonstrates that Braidwood Unit 1 can operate at a maximum power level of 20% with the Auxiliary

. Building Ventilation System inoperable without violation of offsite and control room dose limits.

-Considerable conservatism is introduced into the FSAR analysis of

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ECCS leakage by assuming that elemental iodine becomes a,irborne after release from the core. This conservatism results in radiological celeases which are much larger than expected. Attachment II is an analysis entitled " Iodine and Cesium Releases due to ECCS Leakage" performed by Fauske and Associates, Inc. This analysis addresses the concentration of iodine and cesium within reactor cooling water and its diffusion'into the Auxiliary Building atmosphere at Byron i

Station. The analysis is also applicable to Braidwood Station since

! the Byron.and Braidwood plant system de, signs are essentially l identical. The analysis generates more realistic values of the I

radiological release which would exist during a loss-of-coolant accident based on component leakages of 1 gpm for 72O_ hours and 50 gpm for 30 minutes. These more realistic numbarc are small compared to those generated by the FSAR analysis.

, The Fauske analysis demonstrates that with the conservatism removed, the radiological releases from the Auxiliary Building can.be reduced at leact by a factor of 10. This reduction in source term is equivalent to that resulting f rom the- operation of the Auxiliary Building filters. As a result it should be possible to operate the ,

reactor at 100% power with~the Auxiliary Building Ventilation System inoperable (i.e. no HEPA and charcoal filtration) without exceeding offuite and control room dose limitations.

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1 Negative pressure will not be mandatory in the Auxiliary Building until a minimum reactor power level of 20% based on the analyses discussed above. The need for negative pressure in the Fuel Handling Building is discussed below in Item D.

Abnormal temperatures are not expected in the general areas of the Braidwood Auxiliary and Fuel Handling Buildings during the operation of -Braidwood Unit 1. These expectations are based on Byron Unit 1 '

startup experience. The Auxiliary Building' Ventilation System at Byron did not become operable until the final phase of the Byron Unit 1 startup (75% reactor power). Operating and safety problems due to abnormal temperatures in these buildings did not occur at Byron.

Heat loads in the general areas will not significantly change during operation at power levels between 75% and 100%. Consequently, cooling requirements in general areas of the Braidwood Auxiliary and

, Fuel Handling Buildings will be minimal during operation at any power i

level. -

i To meet the ventilation requirements in tue Braidwood Auxiliary and Fuel Handling Buildings during low power operation, the following will be accomplished prior to initial reactor criticality:

1. Unit 1 and common area ductwork will be installed:

, _ 2. Unit 2 exhaust ductwork may be isolated from Unit 1 exhaust ductwork and Unit 1 areas of the Auxiliary Building may be isolated from Unit 2 areas as necessary;

3. Main supply and exhaust fans will be available to run.

The installation of ductwork and fan availability will allow airflow to be distributed through the Auxiliary and Fuel Handling Buildings for cooling, if necessary. Because contaminated exhaust air will not be a problem, the system's HEPA and charcoal may not be operable and airflows may not be balanced.

3 C. Operation at Higher Power Levels Per the offsite and control room dose analysis discussed above, HEPA and charcoal filtration and negative pressure are not required in the Braidwood Auxiliary and Fuel Handling Buildings during Unit 1

operation up to 20% reactor power. This analysis assumes continuous

, ECCS leakage of 1 gallon per minute. If this ECCS leakage is reduced, Braidwood Unit 1 can safely operate at higher reactor power levels without HEPA and charcoal filtration and negative pressure.

The onclosed Figure 1 indicates the allowable power level at any given Auxiliary Building ECCS leakage rate without filtration of the Auxiliary Building' air and with acceptable low population zone,(LPZ),

exclusion area boundary (EAB) and control room doses. These corves are based on the assumptions noted on the figure.

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t It is clear from Figure 1 that the control room dose is most restrictive of high power operation. Based on the control rdom dose curve. Braidwood Unit 1 can potentially operate at a reactor power level of 97% with the Auxiliary Building Ventilation System inoperable without exceeding offsite and control room dose limitations.

Prior to Exceedinq 20% reactor power operation of Braidwood Unit 1, the leak rate from ECCS equipment in the Auxiliary Building will be determined. This leak rate wil'1 be used to determine, from Figure 1, the available range of reactor power operation with the Auxiliary Building, Ventilation System inoperable. Braidwood Unit 1 will operate within the available power range until the Auxiliary Building Ventilation system becomes operable. At that time these operating limits w.ill no longer apply. Because of the expected short duration of these limitations, no ongoing surveillance will be conducted once the leak rate has been determined for a given power level.

Prior to operPJ'on of Braidwood Unit 1 outside the limits determined from Ficure 1 the following activities will be completed:

1. The airflow through main supply ductwork, accessible and nonaccessibl.e areas exhaust ductwork and the Fuel Handling Building exhaust ductwork will be balanced to support Unit 1 operation.

2. Accessible areas, nonac'cessible areas and Fuel Handling Building exhaust filter plenums will be filter tested as follows:

2.1 Prior to the balance of airflow through the ductwork the following tests will be performed:

2.1.1 Visual Inspection 2.1.2 Housing Leak Test 2.1.3 Mounting Frame Pressure Leak Test 2.1.4 Airflow Capacity and Distribution Test 2.1.5 Air-Aerosol Mixing Uniformity Test For items 2.1.4 and 2.1.5 artificial resistance, in lieu of actual filters, may be used to achieve required pressure drops.

2.2 After the balance of airflow through the ductwork, the following tests will be performed. However, prior to these tests, it will be verified that the airflow rates through each plenum are approximately the airflow rates established in the previous tests.

2.2.1 In-Place Leak Test, HEPA Filters 2.2.2 In-Place Leak Test, Adsorbers

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3. A preoperational test to verify proper functioning will be performed on the Unit 1 and common portions of the system.

After completion of the above activities, the Auxiliary Building o Ventilation System will function to ensure safe operation of Braidwood Unit 1 at all power levels.

D. Fuel Handling Building Pressure Section 6.5.1.3 (Item h) of the Byron /Braidwood FSAR states that the Fuel Handling Building Exhaust System is designed to maintain the Fuel Handling Building at a negative pressure of 1/4 inch water gauge with respect to atnosphere. A negative pressure is required in the Fuel Handling Building to prevent uncontrolled, unmonitored airborne L radioactive releases.

During the normal operation of Unit 1 prior to Unit 2 completion, it

- will be very difficult for the ventilation system to perform this 4 function. The Fuel Handling Building is located between the Unit 1 and Unit 2 containment buildings and the equipment hatches from each containment open directly into it. During Unit'2 construction the Unit 2 hatch will remain open. ,

Radioactive releases from the Fuel Handling Building will not occur without spent fuel in the fuel pool. Only as a result of a Unit 1 '

core unloading event will there be spent fuel in the building prior s to Unit 2 completion. Only under these circumstances will the Fuel Handling Building be maintained at a negative pressure with respect to the atmosphere.

3.0 Summary The entire Auxiliary Building Ventilation System at Braidwood does not l need to be operable to support Unit 1 operation. Construction and testing of Unit 1 and common portions of the system will be completed at various stages in the Unit i startup program. Adequate filtration and cooling provisions will exist in the Auxiliary and Fuel Handling Buildings to support Unit 1 operation during Unit 2 construction. After Unit 2 completion, the entire Auxiliary Building Ventilation System will be integrally balanced and preoperationally tested as required.

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FIGURE 1. MAXIMUM POWER LEVEL VS. ECCS EQUIPMENT LEAKAGE-Assumptions: - 25 CFM Control Room Unfiltered Inleakage

- Thyroid Dose Calculations Include Contributions From:

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' 2. Massive ECCS Leakage (50gpm for 30 min. at I

T=24 hours);

3. Thirty (30) day ECCS Continuous Leakage as l

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____ _=

. . ATTACHMENT I OFFSITE AND CONTROL ROOM DO3ES BRAIDWOOD STATION

Reference:

Letter from A.D. Miosi to H.R. Denton dated April 1, 1986 The operation of the Auxiliary Building Ventilation (VA) System directly affects the radiological doses both offsite and in the Control

' Room. Air exhausted from the Auxiliary Building is fiJ tered and released to the atmosphere and can potentially leak into the control room boundary through normal mode and purge mode intake dampers which are closed during emergency operation.

The radiation levels of the air exhausted from the Auxiliary Building is directly proportional to the power level of reactor operation and the assumed ECCS equipment leak rate. The HEPA and charcoal filters in the Auxiliary Building Ventilation System will operate to maintain radiological releases at a minimum during a loss-of-coolant accident at high power levels. It is not necessary for these filters to operate at low power levels due to reduced fission products. Calculations using regulatory source terms have demonstrated that control room dose and offsite dose can be maintained within acceptable limits without the Auxiliary Building Ventilation System operable at low reactor power levels. The following is a summary of those calculations:

4 A. Offsite Doses The offsite radiological consequences of a design basis loss-of-coolant accident have been calculated using the methodology described in the NRC Standard Review Plan, Section 15.6.5, Appendices A and B.

The results of the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) doses were given in FSAR Table 15.0-12 and for as-designed station. The meteorology of Table 15.0-14 was used.

If the VA system is not tested and certified to remove postaccident radioiodine, Appendix B of SRP 15.6.5 requires different ECCS leakages to be considered. The leakages are as follows:

Applicant assumption 3910 cc/hr or about (FEAR Table 15.6-15a) 1 gallon / hour NRC assumption with VA 1 gallon / minute non-operable: continuous leakage, 30 days Massive leakage 50 gpm for 30 minutes at I t = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> -

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The resulting thyroid doses at the EAB and LPZ are as follows:

Thyroid dose, rem EAB LPZ With Applicant's values and no VA 164.5 81.1 With NRC assumptions and no VA 784 576 10CFR100 limit 300 300 Thus according to the NRC assumptions, the power level would be limited to 300/784 or about 38% to meet the 10CFR300 limit of 300 rem at the EAB.

B. Control Room Dose Analysis of the control room dose during a loss-of-coolant accident is discussed in Section 6.4.4.1 of the Byron /Braidwood FSAR. The results of our recont reanalysis of the Braidwood control room habitability were transmitted to the NRC for review with the referenced letter. This analysis assumed all engineered safety systems to be operable and is the base case for NaC safety

~ evaluation.

If the auxi.11ary building ventilation system (VA) is assumed to be inoperable, then the above analysis is repeated with the following assumptions:

1. Containment leakage as identified in the Byron /Braidwood FSAR:

2. Auxiliary Building Ventilation System inoperable; no filtration is provided;

3. 10% of iodine released from ECCS equipment leakage is assumed airborne:

4. Fluid leakage from the Emergency Core Cooling System (ECCS) is assumed as follows:

- 1 gallon per minute (qpm) leak for 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (continuous leakage);

- 50 gpm leak-for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (massive leakage).

(If the VA system is operable, the only leakage term is a continuous 1 gallon per hour.)

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All calculations were performed per the Murphy-Campe paper. The results are shown in Figure 1. These figures present thirty day control room dose at 100% power as a function of unfiltered inleakage for each of the contributing source terms: containment leakage, continuous ECCS leakage (both 1 gph and 1 gpm are shown),

and massive ECCS leakage. Figure 1 combines the doses for the three source terms (containment leakage, continuous ECCS leakage and massive leakage) and, in addition, shows maximum allowable power level that can be achieved without exceeding the 30 rem dose limit.

Figure 1 shows that with unfiltered inleakage of 25 cfm and continuous leakage of 1 gpm (unfiltered), the reactor is limited to 20% power.

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FIGURE 1. POSTACCIDENT CONTROL ROOM DOSE (calculated using USNRC assumptions) 1 10 4 _

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