3F1197-09, Provides Description of Control Complex Habitability Envelope Inleakage Testing & Revisions to Calculational Methodology Used for Determining CR Operator Dose at Plant, Unit 3.W/plan for Resolving Discrepancies & Commitments
| ML20198S902 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/10/1997 |
| From: | Rencheck M FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20198S910 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 3F1197-09, 3F1197-9, NUDOCS 9711140158 | |
| Download: ML20198S902 (43) | |
Text
-
l Florida Power 22".T,"
C.*'",* .?,i u. m n November 10,1997 3F1197-09 ,
U.S. Nuclear Regulatory Conunission Attn: Document Control Desk .
Washington, DC 20555-0001
Subject:
Control Room liabitability, NUREG-0737, item Ill.D.3.4
References:
Dear Sir:
The purpose of this letter is to provide a description of tir Coitrol Complex liabitability linvelope (CCllE)inleakage testing and the revisions to the calculational methodology used for detennining control room operator dose at Crystal River Unit 3 (CR-3). As discussed in the public meeting on October 31,1997, letween Florida Power Corporation (FPC) and the Nuclear Regulatory Commission (NRC), the inleakage testing and the revised calculational methodology constitute a change to the method of compliance with NUREG-0737, item lit.D.3.4, " Control Room liabitability," for CR-3. The results of the inleakage testing and the revised calculational methodology demonstrate compliance with the dose limits of NUREG-0737, item Ill.D.3.4 and will be used to demonstrate operability of the CCilE and the Control Room Emergency Ventilation System (CREVS) prior to restart from the current outage.
The infonnation provided in Att hment B to this letter describes FPC's overall plan far /
resolving the discrepancies reported in Licensee Event Reports (LER) 97-022-00 and J 97-022-01, perfonning inleakage testing of the CCllE, and for detennining the operability of the CCllE arid CREVS through dose analysis. This submittal includes a description of h physical modifications performeo '*nprove habitability (Attachment C), a summary of revised control room dose calculations (Attachment D), and a description of the test method j g())
and the test results of recent integrated inleakage tests of the CCllE (Attachments E and F, fl respectively).
- TA"= mL " lilB,8 EllVM01IJ P PDR CRYSTAL RIVER ENERGY COMPLEX
- 15760 W, Power une Street . Crystal Rrver, Florida 344284708 . (352)7954486 A Florida Progrese company
i ,
'* U.S. Nucle:r Regul: tory Conunission IF,1197-09 Page 2 of 3 As discussed in greater detail in Attaciunent C to this letter, FPC has made substantial hardware modifications which significantly improve the CCilE. This work reflects FPC's dedication to increasing design margins and addressing issues for the long term.
Specifically, the investigation of the identified problems has promoted a much better understanding of the integratx performance of the CCllE. FPC has perfonned critical reviews of CRiiVS orieration, and the interdependence of CREVS and the habitability envelope for protectioa of the operators. We are confident that the extensive modifications to the CREVS, and especially the installation of redundant bubble tight dampers in place of the originally installed equipment, have made a significant improvement in the ability of the habitability envelope to be reliably isolated from external radiological or toxic gas threats.
Additionally, die installation of vestibules at all habitability envelope boundary doors, creating a second barrier to door inleakage, provides an extra measure of defense in depth for inleakage control.
In die course of improving inleakage tightness of die habitability envelope, FPC has critically examined every wall of the Control Complex, as well as the floor and roof. The condition of each penetration in the envelope has been assessed. PeretMons have been scaled on all st.rfaces, climinating potential inleakage sites that may have n..sted pre Musly.
The remaining souces of inleakage have been identified and work is ongoing to improve the leaktightness of those sites. Testing has demonstrated that inleakage is extremely low for a structure the size of the CR 3 Control Complex.
As discussed at the October 31, 1997, public meeting, the changes in the control room habitability method of analysis constitute an unreviewed safety question (USQ), FPC will submit a License Amendment Request (LAR) to address this USQ and revise the existing Impmved Technical Specifications (ITS) for the CREVS by November 24, 1997. The proposed License Amendment will incorporate changes to the CR-3 design and licensing basis consistent with the information provided in this submittal. The changes will include:
- inclusion of Surveillance Requirements for the CCllE in ITS 3.7.12, e stipulation that CCllE integrity is required for CREVS operability in ITS B 3.7.12, and e revision of the Ventilation Filter Testing Program in ITS 5.6.2.12 to include current standards for charcoal filter testing and reduce allowable pressure drop.
Approval of these changes is not required prior to restmt from the current outage since the first perfonnance of the surveillance will be during Refuel Outage 11.
FPC is confident diat the major upgrade of the Control Complex liabitability Envelope and Control Room Emergency Ventilation System that will be completed prior to startup, will provide a high degree of protection to control room operators. We look forward to ;
I
U, S. Nucle:r Itegulatory Commission 3Fi !97-09 Page 3 of 3 ,
discussing our specific improvements and control room habitability calculational approach with you further at our meeting currently scheduled for November 14, 1997, in the NRC's offices.
Comminnents made in this letter are identified in Attachment A. If you have any questions conceming the information provided herein, please contact Mr. David Kunsemiller, Manager, Nuclear 1.icensing at (352) S63-4566.
Sincerely, JA%be,l-.
M. '. Itencheck, Director Nttticar Engineering and Projects MWit/kdw ,
xc: Regional Administrator, llegion 11 vNitR Project Manager Senior Resident inspector Attachments:
A.1.ist of Regulatory Commitments
- 11. Executive Summary C. Control Complex liabitability Modifications D. Summary Repoit of CR 3 CCllE Dose Calculations E. Control Complex Integrated inleakage Testing F. Summary Report Tracer Gas Test of CR-3 CCllE
ATTACHMENT A List of Regulatory Conunitraents
t U. S. Nuclear Regulatory Commission Attachment A 3FJ 197-09 Page1ofI ;
i i
ATTACilMENT A :
i List of Regulatory Commitments ;
i r
The following table identifies those actions committed to by Florida Power Corporation in this ,
document. Any other actions discussed in the submittal represent intended or planned actions i by Florida Power Corporation. They are described to the NRC for the NRC's information -
- and are not regulatory commitments. Please notify the Manager, Nuclear Iicensing of any questions regarding this document or any associated regulatory commitments, i
II) Nu,mber Commitment Commitment Date 3Fil97-09-1 Florida Power Corporation (FPC) will submit a Submit LAR by .
License Amendment Request (LAR) to address 11/24/97 this Unteviewed Safety Question (USQ) and ;
revise the existing improved Technical Specifications (ITS) for the Control Room Emergency Ventilation System (CREVS). The proposed License Amendment willincorporate ,
i changes to the CR 3 design and licensing basis-for control room habitability. )
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ATTACIIMENT B Executive Summary
U. S. Nuclear Regulatory Commission Attachment B 3Fil97 09 Page 1 of 3 l
i ATTACHMENT 3 i
EXECUTIVE
SUMMARY
i As reported in Licensee Event Reports (LER) 97-022-00 and 97-022-01 dated August 18,1997 and August 27,1997, rewectively, Florida Power Corporation (FPC) identified errors in the -
current dcsign and licensing basis for control room habitability. The current licensing basis is !
principally based on the " Control Room IIabitability Evaluation Report," dateo June 30, i 1987, as approved by the NRC in its Safety Evaluation Report (SER) dated May 25, 1989 The LERs reported errors in the calculation of the Control Complex liabitability Envelope (CCllE) inleakage. During the October 31,1997, public restart meeting, FPC committed to revise the habitability analysis to correct these deficiencies. This submittal provides the description of the modi 0 cations, inleakage testing and calculational methodology for determining control room operator dose.
Findings on Control Room Operator Dose The CCilE at CR 3 is a five story, unpressurized envelope. This is different from many other ,
plants that typically have a one or two story pressurized envelope. FPC has made a number of ,
modifications to dampers and has sealed penetrations in an extensive effort to reduce air L
- inleakage into the CCllE.
1 The existing CR 3 licensing basis credits an analytical method for ensuring desiFa margin related to CCilE inleakage. This method for determining inleakage contained input
, assumptions that were not valid as reported in Licensee Event Reports (LER) 97-022-00 and 97 022 01 dated August 18, 1997 and August 27, 1997, respectively (see " Calculated inicakage" column of Table B 1). Therefore, FPC determined that a test of the CCilE should be performed to determine CCilE inleakage.
~
FPC performed a test to measure CCilE inleakage using tracer gases. This is an industry recognized approach and provides an overall assessment of the integrity of the CCilE. The results of the FPC evaluations are summarized in Table B 1. This table compares the existing licensing basis operator dose with the new test oriented results. The first column of Table B-1 summarizes key parameters impacting the control room dose. The results reflect an overall improvement in margin due to the extensive efforts performed as described in Attachment C.
The tracer gas test was conducted at a differential pressure of 0.171 inches water gauge (in. ,
wg). The test was based on measuring the inleakage across the CCllE. This condition is representative of the inleakage mechanism applicable to the design of the CR-3 CCllE (zone isolation system with filtered. recirculated air).
Since the CCllE is maintained at a neutral pressure by the Control Room Emergency Yentilation System.(CREVS), the only mechanism for developing an inleakage pressure under
- - U. S. Nuclear Regulatory Commission Attachment 11 3FJ 197-09 Page 2 of 3 Maximum flypothetical Accident (MilA) conditions is the effect of the outside wind on the building internal pressure. The specific wind speeds responsible for the building pressure were selected in accordance with the Murphy Campe assumptions regarding the distribution of wind speeds during an accident. This standard method for calculating oose selects the wind speeds to produce a conservative estimate of the control room operator dose. Using the wind speed to calculate the differential pressure at the four periods delined by Murphy-Campe during the accident, the total dose during the accident is 26.5 REM, including approximately 60 sq. in of breach margin in the envelope beyond the conservative inleakage measured during the tracer gas test (see "MilA" column of Table 11-1). FPC is using this mechanistic inleakage approach to demonstrate the operability of the CCllE and CREVS.
FPC is continuing to investigate several technical issues regarding the impact of the operation of non safety related ventilation systems and of the thermal and clevation driven " stack effect" on CCilE inleakage. it is expected thut the contribution of these effects to control room habitability will be bounded by the wind induced leakage mechanism.
FPC has determined that the changes in the control room habitability method of analysis constitute an unreviewed safety question (USQ). A License Amendment Request (LAR) to address this USQ and revise the existing ITS for CREVS will be submitted by November 24, 1997. The proposed License Amendment will include the changes to the CR-3 design and licensing basis consistent with the information provided in this submittal.
in summary, an overall improvement in the CCllE has been achieved through maintenance and modification activities. FPC is confident that the major upgrade of tne CCliE and CREVS, that will be completed prior to startup, will provide a high degree of protection to control room operators, t
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- - U. S. Nucicar Regulatory Commission Attachment B i 3FJ 197-09 Page 3 of 3 ;
i Table B 1 Crystal River Station Unit 3 - Control Room liabitability l Summary of Control Room Operator Dose Evaluation j i Licensing Basis Mechanistic Inleakage Evaluation Attribute Calwlated MiiA ,
Inleakage Concentration MilA with Murphy-Campe MilA with Murphy Campe method method Basis of inleakage 1/8 in. AP and calculation Wind driven AP based on based on component Murphy-Campe wind speed characteristics duration definition and tracer gas test leakage inleakage (cfm) '
- Tomi 355 80 to 435
- Unfiltered ' 285 80 to 435 Operator Dose (REM) 26.5 ,
26.5 Breach Margin (sq. in. 40 60 approximate)
Notes: ,
- 1. Operator dose within the Crystal River Unit 3 Control Complex liabitability Envelope t (CCilE) design (zone isolation with filtered, recirculated air) is primarily controlled by inleakage and concentration. ,
- 2. Tracer gas test measured CCilE leakage as a resul*. of 0.171 in, wg depressurization of Auxiliary Building with CCilE ventilation system operating in recirculation.
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r ATTACHMENT C Control Complex Habitability Modifications
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U. S. Nucle:r Regulatory Commission Attachment C :
3h 197-09 Page 1 of 5 i ATTACllMENT C ,
CONTROL COMPLEX 11 AHITAHil lTY MODIFICATIONS !
Hackground De Control Complex liabitability Envelope (CCilE) is defined as t e volume of the entire zone serviced by the mntrol room ventilation system. For CR 3 this zone includes all the areas in the Control Complex execpt the Controlled Access Area (CA) at floor elevation 95'-0". The CA is a ;
non-essential area that is isolated from CCilE by automatically closing danper AllD 12. The CCilB walls are a minimum 2 ft. of mnerete, with the roof and cast wall exposed to the outside environment. The CCilE is credited, in conjunction with the Control Room Emergency Ventilat!on System (CREVS), with protecting the Control Room operator in the event of a radiological or toxic gas accident.
CCllE inicakage is a significant factor in Standard Review Plan (SRP) 6.4 methodology for assessing operator radiological dose. A number of modifications have been implemented during the resolution of Restart issue R 12 to improve the lategrity of the CCllE and thereby the level of ,
operator protection provided. Generally, these modifications focused on decreasing inleakage through individual boundary elements. Issues pertaining to single failure and means for improving reliable ventilation to the Control Complex Mechanical Equipment Room are also addressed within this modification scope.
Nonnal Om ration - Prior to Modifications in the normal mode of operation, supply air is filtered through the normal filters (AliFle3A, B) and conditioned by the cooling coils (AllllE-SA, B) or by the electric heating coile (AllllE-4A, B). Conditioned air is supplied to all areas of the Control Complex by a normal supply fan (AliF-17A or B). Branch ducts to the various rooms and areas within the Control Complex are provided with thermostatically controlled in duct heaters to maintain the temperature during changes in heat i load. Supply air, except air supplied to the CA, is returned to the conditioning equipment by a return fan (AliF-19A or B). Outside air dampers (AllD 1, ID) are throttled to admit makeup air for ventilation. Recirculation damper AllD 3 is throttled to provide the desired mix ratio of outside air and return air prior to reconditioning. Relief damper AllD-2 is normally open.
Air is supplied to the non-essential CA through dampers AllD 12 and exhauszd through fume hoods and CA exhaust fans (AliF-44A, B; -20A, B) to the Auxiliary Building Exhaust System.
The fume 1. cod auxiliary supply fan (A11F-30) provides additional makeup air (from the Turbine Building) required by the fume hoods. [
A Mechanical Equipment Roc.m exhaust fan (AliF-21A or B) is provided to ventilate the Equipment Room Elevator Machine Room, und the Control Room Lavatory. Isolation dampers ,
U. S. Nuclear Regulatory Commission Attochment C 7 311 197 09 Page 2 of 5
- AllD 24 and AllD 25, or AllD-26 and AllD 27 are normally open to provide an exhaust path to !
the outside. Makeup air for the Mechanical Equipment Room is supplied through damper AllD- l
- 99. ;
Emeryncy Operathm (Reclerulatkm Mode) - Prior to Modificatkms !
Upon detection of high reactor building pressure or toxic gas, all dampers that form the boundary :
of the CCllE are automatically closed, in addition, the mechanical equipment room exhaust fan, i CA fume hood exhaust fan, CA fume hood auxiliary supply fan, and CA exhaust fan are de- ;
energized and their corresponding isolation dampers close. The return fan, normal filters, normal supply fan, and the cooling (or heating) coils remain in operation and operate in a recirculating .
mode. The emergency fans and filters can be placed in service by the operator.
Upon detection of high ramation by RM AS, the dampers that form the boundary of the CCilE automatically close. The mechanical equipment room exhaust fan, CA fume hood exhaust fan, CA '
fume hood auxiliary supply fan CA exhaust fan, normal supply fan, and return fan are tripped and their corresponding dampers close. Manual action is then required to start an emergency fan and restart a return fan to operate in the recirculating mode using the emergency filters. The cooling (or heating) coils remain in operation.
Scope of Modifications The following modifications are being implemented to address the concerns associated with Restart issue R-12. Figures C 1 and C 2 provide a schematic of the pre- and post-modificadon configurations. Note that except as otherwise stated, pairs of dampers replacing a single damper receive the same control signals and act in unison, such that system logic is not changed.
e Damper AllD 99, which brings supply air to the Ventilation Equipment Room is being
, removed and a permanent blank installed. New supply and return registers shall be '
installed in the ductwork (164' elevation) which will now serve as the ventilation for this area, This will climinate AllD 99 as a potential source of inleakage.
- Existing damper AllD 12, located in the supply duct to the CA, is to be removed and replaced with two new bubble tight dampers, AllD-12 and AllD-12D.
- Existing damper AllD 2, located in the exhaust duct to the outside, is being locked <
open and abandoned in place. Two new bubble tight dampers, AllD-2C and AllD 2E, shall be installed in series in the exhaust path. AllD 2C will be normally closed.
.. The position of recirculation cir damper AllD-3 will be established during the process of b .ancing the system for the normal operating mode.
- Dampers AllD 1 and AllD-ID,-located in the air intake duct, are being disabled and abandoned in place. Two new bubble tight dampers, AllD-lC and AllD-lE, are being t
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U. S. Nuclear Regulatory Commission Attaciunent C 3D1197 09 page'3 of 5 installed in series on the inlet duct. Dampers AllD lC, AllD 2C and AllD-3 will retain positioners which provide a manual override feature. This feature allows operators to position these dampers to modulate the outside airflow as required for purging smoke or other contaminants from the CCll!!.
- Mechanical liquipment Room Ventilation Air llandling Fans, AllF-21 A/B and associated dampers AllD-24. AllD-25. AllD-26, and AllD 27 are being spared in place and the associated CCllii penetration scaled. This portion of the syr. tem originally exhausted air from the Mechanical liquipment Room, lilevator liquipment Room, lavatory, kitchen and toilet. This climinates another potential source of inleakage into the CCilli.
- New supply and return registers are being installed in tbc ductwork in the Mechanical
!!quipment Room. This will provide ventilation to this portion of the CCllli during both normal and recirculation modes.
- A skid mounted air handling unit consisting of a fan and a charcoal filtration unit will be installed to ventilate the lilevator I!quipment Room, lavatory, kitchen and toile:.
This system is non safety and non seismic and will vent approximately 1,0(X) efm by way of a field connection to a non safety related (NSR) duct.
- Small bore drain pipes penetrating the CCllii are being fitted with loop seals to prevent inleakage though the lines. These will be added to a queued work request in the work controls system which maintains CCilli drain line loop seals.
- Vestibules have been installed over all CCilli boundary doors, and have been scaled to provide maximum leaktightness. These vestibules provide a means to test individual CCllli boundary door leaktighmess, as well as reducing inleakage associated with CCilE access / egress, in addition to the above modifications, an extensive effort was undertaken to survey CCilli penetrations and seal as required to minimite inleakage. As a result of this work, it is concluded that conduit penetrations do not pose a significant liability to CCllii integrity.
Penetrations associated with electrical cable banks were inspected and scaled to the extent feasible with existing procedures and materials, but some leakage paths remain through the interstitial spaces between individual cables. Additional work is being planned to improve the sealing of penetrations with the most significant leakage.
U.S.'tjucl:or Regulatory Coramission Attachment C 3f1197-09 Page1 of.L l
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ATTACIIMENT D Summary Report of CR-3 CCHE Dose Calculations
U, S. Nuclear Regulatory Commission Attachment D 3F1197 09 Page 1 of 13 ATTACllMENT D .
CONTROL, ROOM POST-ACCIDENT DOSE CALCULATIONS ;
Purpose This attachment provides a discussion of calculations for post accident dose to the control ,
room operator. Specifically, iodine and beta skin doses are assessed. The Maximum-Ilypothetical Accident (MilA) is taken to be the limiting event from the perspective of Control Room habitability and is initially assessed. Ilowever, the Steam Generator Tube Rupture (SGTR) is also assessed to evaluate events which might require radiation monitor RM AS (for Control Complex liabitability Envelope (CCllE) isolation), and the Fuel flandling Accident (Fila) is assered to evaluate the source terms and model associated with this event. Inputs for these calculations are listed and compared against those in previous analyses. The bases and effect of selected changes are also discussed.
pcensing nasis - nackground i
System Readiness Reviews conducted in 1997 during the CR 3 Design Basis Outage identified ;
several issues which potentially impacted control room habitability. He predominant issue i pertained to the validity of assumptions for CCllE inleakage. Other notable items of concern included Control Room Emergency Ventilation System (CREVS) recirculation flow rate and charcoal filter efficiency. Restart issue R 12 was initiated to investigate these issues and provide resolution in support of the restart effort. Modifications were made to reduce CCllE inleakage by improdng the integrity of boundary elements. Existing boundary dampers were replaced with zero
-leakage models, and redundant dampers provided at all boundary damper kications. The mechanical equipment room exhaust duct was spared in place, and the CCllE penetration for this duct scaled. Minor CRFYS design changes were made to provide alternate means of mechanical equipment room ventilation and to improve system reliability. Programmatic changes were made to ensure that the assigned efficiency of the Control Complex charcoal filters was consistent with regulatory guidance.
The findings and modifications arising from resolution of R 12 required that the Control Room operator dose calculations be revised to align inputs and assumptions with plant design. The basic ,
methodology used in these revised calculations is consistent with that foun, in regulatory guidance and utilized in previous calculations. Determination of CCilE inleakage, performed by tracer gas ,
testing, is different than the previous methodology and is described in detail in Attachment E to this submittal.
Analysis of the F'IA was performed using the post-accident model described in Regulatory Guide 1.4 and source terms derived from TID 14844. SRP 6.4 and the Murphy Campe Paper on Control -
Room Ventilation System Design were used as guidance documents. The following lists specific assumptions associated with control room dose calculations for the MilA. Additional information
1
- Attachment D U 3. Nuclear Regul: tory Commission 3Fil97-09 Page 2 of 13 pertaining to selected parameters is provided in the discussion that follows. A detailed set of inputs and assumptions is provided in Tables 1 and 2. ,
. Assumptions for the hill A This analysis uses 102% of the rated thermal power (2619 h1Wth).
- The containment free volume is 2,000,000 cubic feet. The sprayed volume is 1,304,000 ft' and the remainder is unsprayed.
- Containment air mixing rate is equal to 2 unsprayed volumes per hour between the sprayed and the unsprayed volumes.
The core fission products released to the free volume are 25% of the total iodine and 100% of the noble gases.
- The core fission products released ;o the sump water is 50% of the total iodine.
The iodine species fractions for the free volume are: 0.91 Elemental + 0.05 Particulate
+ 0.04 Organic.
- The post LOOP /LOCA containmecat design leakage rate is 0.25% for the first day and 0.125% for the remaining post accident recovery period. 3 Modeling includes continuous ECCS leakage outside the containment building.
- Modeling includes 30 minutes of 50 gpm ECCS leakage outside the containment building beginning 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA initiation event.
The vaporization fraction for the ECCS leakage is not less than 10%,
- The iodine removal ef0ciency for the 2*' thick recirculation charcoal filter is 95% based on meeting the testing requirements of Reg. Guide 1.52.
- Meteorology extrapolations and control rovm dose modeling and calculations (when equilibrium conditions are present) are based on the Murphy Campe methodology for meeting GDC 19 of 10CFR50.
- The containment spray elemantal iodine removal cut off is based on a decontamination factor (DF) of 100, and the particulate iodine spray removal constant is based on a DF of 50 for reducing the constant by a factor of 10.
The spray starts 124 seconds (0.03444 hours) after containment isolation which is assumed to occur instantaneously. This time is more conservative than full Dow time specified in Tech Spec 3.6.6.
During the first 30 minutes post LOOP /LOCA, there is no forced ventilation flow in the Control Complex.
Fission product solids that might be in the sump water are assumed to be non-volatile and are not released to the environment.
The sump water volume is assumed not to be reduced due to ECCS leakage.
Discussion of Selected MilA _ Radiological Dose Calculation Inputs
- 1) Dose Conversion Factors FPC applied revised dose conversion factors for Control Complex liabitability dose calculations. Specifically, FPC changed from using International Commission on Radiological Protection Publication 2 (ICRP-2), published 1959, to International Commission on Radiological Protection Publication 30 (ICRP-30), published 1979, and Y
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U,. S Nuclear Regulatory Commission Attcchment D 3F1197 09 Page 3 of 13 Federal Guidance Report #11, published 1988, for calculating thyroid dose to control room operators.
Dose calculations for internal organs such as the thyroid are performed using dosimetric and metabolic models contained in the ICRP publications. The current I NRC Safety Evaluation of FPC's control room habitability is based on the " Control Room liabitability Evaluation Report" submitted to the NRC on June 30, 1987. At that time, ICRP-2 methodology was used for internal dose calculations.
Revised methods for calculating organ dose and relating organ dose to whole body dose were published in ICRP-30, and endorsed for use in this country by the Environmental Protection Agency (EPA) in Federal Guidance Report #11. These documents changed the dose conversion factors that are used to convert a quantity of inhaled radioactive material to organ dose. For the radionuclides of concern, use of ICRP-30 / Federal Guidance Report #11 dose conversion factors results in the accident thyroid dose to bc l
~ 30% lower than previously calculated.
The statutory authority for the use of ICRP-30 and Federal Guidance Report #11 can be found in the Statements Of Consideration in the publication of revised 10 CFR 20 as a Final Rule, in Federal Register 56 Fl. 23360, published May 5,1991 and effective June 20, 1991, the use of ICRP-30 and Federal Guidance Report #11 for calculating internal doses is discussed. The NRC addressed a public comment regarding Section 20.1204, " Determination of Internal Exposure" by stating that:
" Appropriate parameters for calculating organ doses from radionucllde uptakes can be found in ICRP-30 and its supplements. Dose factors in Federal Guidance Report #11 are also acceptable for use in calculating occupational exposures for compliance with either 1120.1 - 20.601 or with ((20.1001 - 20.2401, except that the individual organ dose values must be used for {$20.1 - 20.601." (No:e: 1620.1 - 20.601 were the former sections of 10 CFR 20 that remained in effect concurrent with the revised sections, ss20.1001 - 20.2401, until January 1,1994, after which $520.1 - 20.601 were removed from federal law.)
CR 3 Improved Technical Specifications (ITS) include specific activity limits for primary and secondary coolant. Specific activity is measured and reported as DOSE EQUIVALENT l 131, which is a defined term in the ITS. The ITS definition of DOSE EQUIVALENT l-131 specifies that the thyroid dose conversion factors used for this calculation shall be those from ICRP-30.
Evaluations of postulated accidents include esti mation of offsite doses that could result from radioactive material releases. A standnJ assumption applied in determining the amount of radioactive material released is that the reactor coolant activity is equal to the Technical Speci0 cation limit. Since DOSE EQUIVALENT l-131 is used as a measure of the permissible concentration of radioactive iodine species in reactor ,
coolant, ICRP-30 is currently being used in the calculations of offsite dose consequences. Therefore, the use of ICRP-30 in revised control room dose calculations is consistent with the current licensing basis of CR 3.
1
U,. S. Nuclear Regulatory Commission Attachment D 3FJ 197-09 Page 4 of 13
- 2) Accident Analysis Software (POSTDBA)
Computer program POSTDBA is Sargent & Lundy proprietary software which performs radiological dose calculations and related analyses for the LOCA in a PWR or a IlWR. POSTDBA was originally developed to calculate PWR control room (CR) and offsite doses in accordance with icquirements and recommendations of Regulatory Guide (RG) 1.4, Standard Review Plan (SRP) Section 6.4, and SRP 6.5.2. POSTDBA was revised and revalidated in 1994. This program handles containment leakage, additional gaseous leakage (purge and MSIV), and can model liquid lenkape (constant and intermittent reactor coolant (RC) boundary releases outside containment) as a separate case in the same computer run, in addition to the dose evaluation, POSTDBA calculates the time dependent airborne concentrations of iodine and noble gases for the containment at the CR inlet and in the CR volume. The control room's potential outside air intrusions, iodine Oltration, and gamma body, beta skin, and thyroid dose computations are based on the Murphy Campe approach using time dependent integration techniques. The site meteorology can be entered as predetermined 5th percentile x/Q values or as effective wind speeds as determined by the Murphy Campe methodology. POSTDBA can calculate the 5th percentile x/Qs et the CR inlet using annual or annual average joint frequency meteorology data as input instead of the parameters mentioned above.
POSTDilA is constructed to allow the user to selec' the time steps and to controi variable parameters for each time step. The variables melude containment spray iodine removal rates; post accident source release rates (iodine and noble gases) and anv iodine filtration; x/Q changes; CR parameters (makeup, micakage, iodine remow, breathing rates, and occupancy factors); plus the fractions of elemental, particulate, and organic iodine released to the environment. The first and the following time steps can be used to vary most of the variables, and if needed, the first time step can be used to model a delayed release. This high degree of user control allows other types of accidents to be analyzed.
initial DBA isotopic iodine and noble gas sources (starting at time = 0.0 Seconds) are individually entered, element family release fractions can be applied, and the iodine can be mhdivided into specinc fractions of chemically different types. Any one or all of the three release pathway rates and the containment spray removal factors can be specified for each time step. Control room ventilation inputs include outside air makeup, recirculation, and unfiltered inleakage. Separate filter efficiencies can be specified for the makeup and recirculation filters,
- 3) Reactor Building Spray System The reactor building spray system was analyzed conservatively using only the spray pump recirculation rate of 1112 gpm for the entire time spray removal credit for iodine is permitted. For calculation purposes the spray mode of operation is one header at 1112 gpm for the permitt ed duration of iodine removal credit. The volume of the sprayed region is 65.2% of the total building free volume and the unsprayed volume is 34.8 % .
U.. S. Nuclear Regulatory Commission Attachment D 3FJ 197-09 Page 5 of 13 ,
l l
The SPIRT Computer Code was used to evaluate the spray removal constants for elemental iodine. SRP 6.5.2 calculational methods were used to evaluate the spray removal constants for particulate iodine. Organic iodine remains airborne for the duration of the accident. lodine removal constants derived in CR-3 caleviation # l 86-0002, Revision 5 was used as input to this analysis. Additional inputs and assumptions
, for Ril Spray analysis are found in Tables 3 and 4.
- 4) Control Room Emergency Ventilation System Recirculation Flow Rate 1.ER 97 022 identified that past modifications were implemented which added resistance to Control Complex Ventilation System ductwork without fully assessing the impact on recirculation How rate. As a result, the system Dow rate with clean Glters is now somewhat lower than the 43,500 cfm nommal design How rate. Previous calculations have assessed Control Room dose at values as low as 39,150 cfm (43,500 -
10%). A Dow evaluation of current system performance has established a calcult,ted minimum recirculation flow rate of 37,800 with 4" wg across fouled Glters. This lower bound is conservatively tak:n forward for use in Operator dose calculations.
The 4" Gltet fouling value is less than the 6" currently reflected in the TS and taken as the combined (llEPA and charcoal) filter fouling limit. Current procedures constrain operation within 43,500 cfm +/- 10% and ensure that the 37,800 cfm minimum now rate requirement for dose calculations is met.
- 5) Inicakage Parameters SRP 6.4 identifies that locally high differential pressures at boundary damper locations within the control room ventilation system can have the effect of inducing significant ,
leakage across the habitability boundary. The redundant dampers being installed at all CREVS boundary locations are tested to be bubble-tight at 15" wg, and effectively eliminates this concern for the CR-3 CCllE. Subsequently, the only mechanism of get:ing unfiltered outside air now into the controlled environment following a LOCA with a LOOP would be to induce it by virtue of differential pressure across outside walls such as would be induced by wind pressure.
Wind speeds can be converted into differential pressure (Ap) in the fallowing relationship:
Pv = Ap = 0.00642 p U3 inches of water column, where U is the wind speed in mph.
Air density "p" for this equation is conservatively tuen as that at a temperature of 15 F, and is (pn = pto x Tvo / To = 0.075 lb/ft' x 530 R / 475 R =) 0.0837 lb/ft' .
U,. S. Nuclect Regulatory Commission Attachment D 3F1197-09 Page 6 ef 13 The relationship between inleakage and differential pressure for a fixed resistance can be conservatively expressed as Q = C (Ap)" , where Q is the inleakage flow rate in cfm with the flow cor.fficient C taken as 1 and the flow exponern "n" assumed to be 0.5. From this relauonship the following equation can be obtained which determines corrected inleakage Qc at any differential pressure Ape based on test inleakage Qr at the corresponding test differential pressure Apt:
Qc = Qt / (Ape /Aprf' .
Finally, substituting the results from tracer gas testing (465 cfm at 0.171" wg) and incorporating an additional 10 cfm allowance for access / egress yields the equation which predicts inleakage of the CR 3 CCilE at my differential pressure:
Q = 465 x (Apr / 0.171)" + 10 liased on an initial wind speed of 2.68 mph (1.2 m/sec) associated with the 5% x/Q, Murphy-Campe methodology takes the initial wind speed as that associated with the 5%
x/Q value, then allows for adjustment of the x/Q value in three increments based on the likelihood that initial conditions would not be sustained for extended periods.
Conservatively combining the Murphy-Campe wind direction and wind speed adjustmer.t factors for each interval and applying the result as an effective wind speed adjustment yicids the following wind speeds for the four time intervals: 0 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> @
2.68 mph, 8-24 hours @ 4.55 mph, 24-96 hours @ 7.16 mph, and 96-720 hours @
16.27 mph. Using the equations given above, these values can be used to determine the differential pressure associated with a given time interval, and the differential pressure value used as an input for calculating the inleakage associated with that interval. The following inleakages wr ' determined for each of the four time intervals based on this manner: 0-8 hours @ J , cfm, 8-24 hours @ 129.0 cfm, 24-96 hours
@ 197 cfm, and 96-720 hours @ 435 cfm. These inleakage values were input into POSTDBA to determine the dose for each interval, and the dose summed over all four intervals to yield the total dose for the event. This analysis found a Control Room thyroid dose of 13.93 REM.
Determination of the allowable breach margin is conservatively determined by setting the allowable dose value at the licensing commitment of 26.5 REM and then iterating to determine the inleakage factor which, when applied to the dose calculations described above, will result in this value. Multiplying this inleakage factor, determined to be 1.91, by the previously determined inleakage rates results in the following limiting inleakage values: 0-8 hours @ 152.99 cfm, 8-24 hours @ 246.39 cfm. 24 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> @ 376.27 cfm, and 96 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> @ 831.04 cfm, These inleakages
f 11, S. Nuclear Regulatory Commission Attachment D j 3F1197-09 Page 7 of 13 were then rounded up to the next whole number (i.e.,153 cfm,247 cfm,377 cfm and 832 cfm), and then input into the following equation to determine the breach area.
Q = C Co A (2 Ap/p)" cfm, where Ci = 776 (a conversion factor),
Co = orifice coefficle'nt = 0.65, A = area in sq. ft., Ap is differential pressure in inches water, and p is density at 15'F = 0.0837 lb/ft .
The results of this enalysis show that a breach area of up to 61.3 in' could be incurred in any interval whhout exceeding the limiting inleakage value (and therefore dose contribution) for that interval, based on the 26.5 REM dos: limit. Table 5 provides a detailed list of inputs pertaining to post-accident meteorology and the Control Room model used in this analysis.
Analysis of Other 1.OCA Scenarios FPC is examining alternate LOCA scenarios which could generate leakage into the envelope.
The fans in adjacent buildings that could create differential pressures across the habitability boundary are all non safety related, and not powered during LOOP renditions. Therefore, for these systems to create leakage into the habitability envelope, it must be postulated that offsite power remains available throughout the accident scenario. The Auxiliary Building Ventilation (ABV) System has the potential to induce the largest differential pressure, but ABV operation would be accompanied by the filtration effects of the system's charcoal filters. Studies of CCilE inleakage and operator dose consequences associated with the operation of these systems are ongoing.
Analysis of other DBAs Additiontd analyses were performed for DBAs which might not generate an ES signal, and therefore may require isolation of the Control Complex either by the acti on of R.M-A5 or manual operation action. Based on a comparison of source terms, the Steam Generator Tube Rupture Accident was chosen to evaluate the requirements far CCilE isolation either by radiation monitor RM-A5 or operator action. An analysis was also performed for the Fuel llandiing Accident to evaluate CCllE inleakage requirements for this event (assumed CC Ventilation initially in recirc via administrative comrols). The result of these studies show that, assuming CCllE isolation by either the RM-A5 or operator action in the SGTR, the MilA remains the limit ng event with regard to Control Room !!abitability.
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'o U. S. Nuclear Regulatory Comr.11ssion Attachment D 3h !97-09 Page 8 of 13 Table 1 Significant Core lodine and Noble Gas Fission Products at the Start of a DB-LOCA(MIIA) 2619 MWT ISOTOPE FISSION C1/MWT REACTOR BUILDING YlELD T=0 AIRBORNE INVENTORY In Cl 1131 0.0,29 2.508E + 4 6.5685 E +7 l-132 0.044 3.806E+ 4 9.%79E +7 l133 0.065 5.6220 + 4 1.4724 E + 8 l134 _ 0.076 6.575 E + 4 1.7220E+ 8 l135 0.059 5.103E + 4 1.3365_E + 8 KR-83M 0.0048 4.152 E +3 1.0874E+7 KR 85 _0.0029 4.102 E +2 1.0743E +6 KR 85M 0.015 1.297E+ 4 3.3%8E+7 KR-87 0.027 2.335 E+ 4 6. l l54E+7 KR 88 0.037 3.200E +4 8.3808E+7 KR 89 0.046 3.979E + 4 1.0421 E +8 XE-31M 0.0003 2.595 E +2 6.7%3 E+5 XE-33 M 0.0016 1.384E+3 3.6247E+6 XE-133 0.065 5.622 E +4 1.4724E+G XE 35M 0.018 1.557E+4 4.0778E +7
_ X E-135 0.062 5.363E+4 1.4046E + 8 XE-137 0.059 5.103E +4 1.33651i+8 XE-138 0.0552 4.775 E+4 1.2506E+ 8
--.-___m
y l L . U. S. Nuclear RegJlatory Commission Attachment D 3k1197-09 Page 9 of 13 -
< TABLE 2 List of Assumptions and Parameters to Model the Maximum flypothetical Accident for Control Room flahitability Analysis [
Parameter "
Value Thermal Power (MWt) 2619 Containment Free Volurie (ft') 2.0 x 10* ;
% Sprayed Volume,(ft') 65.2(1,304 x 10') ,
% Unsprayed Volume (ft') 34.8(6.% x 10') l lodine Fraction initially Dispersed in Sgrayed Volume 0.652 lodine Fraction initial Dispersed in Unsprayed Volume 0.348 Air Turnover Unsprayed to Sprayed Volumes 23.200 cfm Air Turnover Sprayed to Unspra) .:d Volumes 23,200 cfm .
Fraction of Airborne lodine Activity Released From the Core 0.25 Fraction of Airborne Noble Gases Released From the Core 1.0 Fraction of Sump lodines Released From_the Core 0.5 Elemental lodine Species (%) 91 Organic lodine Species (%) 4 Particulate lodine Species (%) 5 Maximum Decontamination Factor For Removal of Elemental 100 -
lodines by Sprays Maximum Decontamination Factor For Removal of Particulates 50 Maximum Decontamination Factor For Removal of Organics O Containment Spray Flow Rate-One licader (gpm) _
1112 Spray System Actuation Time Post LOCA(Seconds) 124 lodine Removal Cutoff (hr) 4.4 _ _ _
Time to Sump Recirulation (Min) 29.95 Elemental lodine Removal Constant br 20.46 (To a DF of 100)
Particulate Removal Constant hr 2.21 (To a DF of 50) 8 c Particulate Removal Constant hr '
0.221 (After a DF of 50 for 2.01 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) ,
Containment Leak Rate (%/ Day) 0 24 hr 0.25 Containment Leak Rateg/ Day) 130 Days t 0.125 Recirculation Loop Leakage-Operational Leakage (cc/hr) 4510 cc/hr for duration of the accident Recirculation Loop SRP Assumed Leakage 50 ppm for 30 Minutes Starting 24 llours After Accident Sump Liquid Volume Post LOCA ft' 45.902 Fraction of Recirculation Loop Leakage Flashing to Steam (%) 10 L
i
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Attachment D
. U. S. Nuclear Regulatory Commission !
3h 197-09 Page 10 of 13 l t
i TAllLE 3 Summary of Input Parameters Used for lodine Spray Removal Analysis Input Parameter Value Total Containment Free Volume 2.0 x 106 ft' Sprayed Containment Free Volume 65.2% (1.304 x 10' ft') i 5
Unsprayed Containment Free Volume 34.8% (6 96 x 10 ft') i Spray Nozzle Type SPRAYCO Model 1713A Spray Distribution See Table 2 Number of Drop Sizes See Table 2 Mean Spray Fall licight-One licader Model 110.5 ft Spray Flow Rate- One licader Model 1112 gpm Collection Drop Efficiency 1 Elemental lodine Partition Coefficient Standard Review Plan 6.5.2 Normal Temperature at Which Spray Water is Stored (40-100)'F Maximum Post -Accident Sump Temperature 275 F 12minar Boundary Layer Surface Area 4084 ft 2 Turbulent Boundary Layer Surface Area-One lleader Model 57,708 ft' Water Wall Flow Fraction 0.1 A T Across Wall / Gas Boundary 1.0 F Liquid Volume of Containment Sump 45,902 ft' Containment Wall Surface Area Impacted by Sprays-One 37,900 ft' licader Model Containment Radius 05 ft f
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U,. S. Nuclear Regulatory Commission - Attachment D -
3F1197 Page 11 of 13 TABLE 4 /-
Spray Distribution for SPRAYCO MODEL 1713A Nonle Data Point No. Drop Size (cm) Relative F[.Muency (fraction) 1 3.75-3 .011 2 6.25-3 .027 3 E.75-3 .056 4 1.125-2 .105 5 1.375-2 .095 j 6 1.625-2 .080 7 1.875-2 .070 8 2.125-2 .051 9- 2.375 2 ,
.066 10 2.625 2 .044 11 2.875 2 .026 12 3.125-2 .022 13 3.375-2 .017 14 3.625 2 .020 15 '3.875 2 .023
, 16 4.125-2 .011
^
17 _
4.375-2 .011
~
18 4.625-2 .015 19 4.875-2 .012 1 20 5.125-2 .011 21 5.375-2 .011 22 5.625-2 .016 23 5.875-2 .012 24 6.125-2 .008 25 _ 6.375-2 .008 26 6.625-2 .007 27 6.875-2 .011 28 7.125-2 .009 29 7.375-2 _
.011 30' 7.625 2 .009 31 7.875-2 .008 32 8.125-2 .007 33- 8.375-2 .006 34 8.625-2 .006 35 8.875-2 .008 36 9.125-5 .006 37 9.375-2 .005
- U,. S. Nuclear Regulatory Commission Attachment D 3FJ 197-09 Page !? of 13
/
TABLE 4 - (continued) -
Spray Distribution for SPRAYCO MODEL 1713A Nozzle Data Point No. Drop Size (cm) Relative Frequency (fraction) 38 9.625 2 .005 39 9.875-2 .005 40 1.013-1 .004 _
41 1.038-1 .005 ;
42 1.063-1 .004 43 1.088-1 .005 _
44 1.i13-1 .005 45 1.138-1 _ 005 46 1.163 1 -
.004 47 1.188-1 .005 48 1.213 1 .005 49 1.238-1 .007 50 1.288 1 .005 51 1.313-1 .002 _
52 1.338-1 .002 ,
53 1.413 1 .001 54 1.438-1 .001 55 1.613-1 .001 56 1.738 1 .002 4
- 0,. S/ Nuclear Rtj;alitory Commission Attachment D -
3FJ 197-09 Page 13 ef 13 4
. TABLE 5 List of Assumptions and Parameters Used to Model the Control Room for the Control Room Dose liabitability Analysis Parameter Value Mode of Operation ' Zone Isolation With Filtered Recirculated Air After 30 Minutes i liabitability Envelope Free Volume (ft') 364,92.7 Control Room Free Volume (ft') 88,000 Unfiltered Infiltration Rate (SCFM) '
0-8 hrs 153.0 8-24 hrs 247.0 1-4 days 377.0 4-30 days 832.0 Filtered Recirculation Flow Rate (SCFM) 37,800 Recirculation Charcoal Filter Bed Depth (Inch) _
2 Filter Efficiency for lodines(%) 95 Control Room X/Q values (sec/m')
0-8 hrs 9.00 x 10-*
8-24 hrs 5,31 x 10
l-4 days 2.03 x 10
4-30 days 5.94 x 10'S Thyroid Dose Conversion Factors ~
ICRP-30 CR Breathing Rate m*/sec 3.47 x 10~*
4
'4 ATTACHMENT E Control Complex Integrated Inleakage Testing
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U,. S. Nuclear Regul: tory Commission Attachment E 3FJ 197 09 Page 1 of 13 ATTACllMENT E CONTROL COMPLEX INTEGRATED INLEAKAGE TESTING Purpose This attachment provides a discussion of the conditions and parameters associated with testing the Control Complex liabitability Envelope integrity. A comparison is also provided which examines these test conditions with those assumed in the current liabitability Report for determining CCliE inleakage.
Licensing Basis - Background The CR-3 CC11E design is categorized according to SRD 6.4 as a zone isolation system with filtered, recirculated air. As such, it does not rely on pressurization to limit inleakage, but rather on leaktightness and filtration capability to provide the necessary level of protection. Preti?usly, CCllE inleakage has been determined on the basis of standard methodology as provided in SRP 6.4. This method provides a standard means by which to derive a " baseline" inleakage. The SRP 6.4 method is not based on any particular accident scenario, and contains assumptions outlined below which may not be relevant under post-accident conditions at CR-3.
Although not described in regulatory guidance, application of tracer gas technology is recognized as a means to accurately measure building inleakage, and is being increasingly utilized in the nuclear industry for this purnose. Using tracer gas test methods, it is possible to set up a test to measure inteakage under conditions which are representative of a specific postulated scenario.
Using an inleakage value derived from testing under simulated post-accident conditions is a departure from the existing licensing basis for CCllE inleakage, but justifiable given that CR-3 is not committed to SRP 6.4 except as a guidance document. The following is a comparison of parameters and inputs associated with accident specific tracer gas test methods vs. SRP 6.4 methodology:
- 1) Control Complex Ventilation System Operation SRP 6.4 methodology for determining base inleakage in a recirculation system is to measure (or calculate) the air flow required to pressurize the habitability envelope to 1/8" wg, then divide the result by two. This method ensures that all penetrations are subject to test pressure. No differentiation is made between leakage which may be remote from sources of activity and subject to little differential pressure under post-accident conditions, and those which are adjacent to areas of high activity and subject to higher differential pressures due to wind or fan _ operation. -The SRP 6.4 ,est would include additional enhancements for inleakage through boundary dampers which may be subject to unusually high differential pressures, and a 10 cfm allowance for personnel access / egress daring an accident.
'o U,. S. Nuclear Regulatory Ccmmission Attachment E 3FJ 197-09 Page 2 of 13 Tracer gas testing under post-accident conditions would han the Control Complex in its emergency recirculation mode, and treat the entire CCHE as a lumped, well-mixed volume.
Any local pressures induced by the operation of the Control Complex ventilation system would also exist under post-accident conditions. No additional penalty would be required at boundary damper locations because these dampers are subject to the same pressures during the test and during post-accident operation. (This is inconsequential for CR-3 given that bubble-tight dampers will be installed at all boundary isolation locations.) The application of a 10 cfm penalty for access / egress is applicable, and is incorporated into the final result for CR-3.
- 2) Test Pressures SRP 6.4 methodology would pressurize the habitability envelope to 1/8" wg and then divide the result by two. The 1/8" value is not based on any particular post-accident conditions, but is large enough to minimize the impact of test inaccuracies and local pressure effects. The wind speed necessary to generate this differential pressure is on the order of 15 - 20 mph, much larger than the low wind speeds associa'ed with the 5% worst x/Q value used in SRP 6.4 to minimize source dispersion. Division by two accounts for conservation of mass in recognizing that approximately half the penetration area must be leaking inward (the other half outward) but fails to consMer that this leakage must in reality be induced across the pressure drop associated with two penetrations, not just one.
Given that the Control Room Emergency Ventilation System is a 100% recirculation system with no boundary damper leakage, the CCilE must, by definition, be " neutral" while operating in this mode. This is not to be interpreted as a design requirement regarding local pressures in the CCllE with respect to adjoining areas, but simply recognizing that a feature of this design is that bulk building pressure is controlled by interaction with its surroundings.
Developing a test replicating post-accident conditions requires that the scenario be postulated which provides both realistic and challenging conditions from the standpoint of source term release location and differential pressure to drive inleakage. The postulated conditions may result in a variance of differential pressure across the Control Complex, and are not intended to necessarily test all penetrations to a certain minimum differential pressure. Rather, the objective is to develop a test which accurately gauges the overall consequence associated with the limiting post-accident scenario.
Post-accident Scenario Evaluation For CR-3, the limiting scenario is that associated with the Maximum Hypothetical Accident (MHA), which is taken to be a largely unmitigated LOCA concurrent with a loss of off-site power.
Under these conditions the Auxiliary Building Ventilation System charcoal filters would be inoperative, and not provide filtration for the portion of source term originating in the Auxiliary Building due to ECC$ leakage outside containment. Differential building pressure induced by wind would be the primary motive force for driving inleakage across the CCHE boundary. Given that
'a U,. S. Nucle:r Regulatory Commission '
Attachment E 3FJ 197-09 Page 3 of 13 i
the vast majority of penetrations are either on the north (Turbine Building) or. south (Auxiliary Building) walls of the control complex, north / south wind directions would tend to maximize CCHE inleakage.
Since it is undesirable to pressurize the Auxiliary Building, the tracer gas test which replicates these conditions would utilize the Auxiliary Building Ventilation System to induce a negative differential pressure in the opposite direction. The following conditions were prescribed for tracer gas testing of CCilE integrated inleakage:
- The Control Room Emergency Ventilation System (CREVS) placed in emergency recirculation mode and operating normally. The system will be tested in both the " Toxic Gas" and the "High Radiation" recirculation lineups. All CREVS boundary damper locations shall be sealed " bubble tight" to duplicate the post modification conditions as shown in Figure C-2.
Turbine Building Ventilation System - All fans secured. The turbine building normally remains well vented to atmosphere through normally open doors, roll out windows and roof vents.
- Intermediate Building Ventilation System - All fans secured. Note that conditions in the Intermediate Building are not deemed critical to the test in that few penetrations are on the CCHE / Intermediate Building Wall.
- Aux. Building Ventilation - Operate fans to attain a test pressure of at least 1/8" negative pressure vs. the Turbine Building (See Figure E-1 for building orientations). This value is large enough to minimize test inaccuracies and external effects and should be sustained for the duration of the test.
. - Testing is to be conducted at a time when personnel traffic is minimized.
Access / egress shall be coordinated such that no more than one habitability door be opened at any one time,
- Testing is to be conducted with vestibule doors blocked open.
All loop seals penetrating the CCHE shall be verified filled prior to testing.
3 Controls are in place to ensure that these loop seals are periodically filled during plant operation.
Testing is to be avoided during periods of high wind conditions. Prevailing.
wind speed and direction are to be recorded as part of the test effort.
r Interpretation of test results SRP 6.4 methodology assumes post-accident meteorological conditions corresponding to the 5% x/Q value during the critical initial stages of the event in order to minimize dispersion of the radioactive plume as it is carried from the containment building to the Control Complex.
The methodology then allows for three incremental increases in wind speed and direction over l
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~ - _ _ _ . - _ - - . _ _ , , . .
U,. S. Nuclear Regulatory Commission Attachment E 3FJ 197-09 Page 4 of 13 the duration of the accider,i due to the extreme improbability that these initial wind conditions wauld be sustained over an extended period of time.
Based on these considerations, inleakage values will be derived for each of the four time intervals over which x/Q values vary by correcting inleakage at the test differential pressure to the differential pressure induced by the wind speed associated with that interval. Each of these inleakage values is an input into the appropriate interval in the revised radiological dose calculations such that the wind speed associated with plume dispersion corresponds to that which drives inleakage through the Control Complex boundary.
Test results will be corrected to differential pressures associated with wind speeds utilized in control room post accident dose calculations. The calculated radiological consequences of a design basis LOCA are a function of both meteorological conditions and unfiltered inleakage.
Accordingly, the differential pressures associated with the wind speeds used in post-accident meteorology are input into revised radiological calculations to determine allowable CCHE inleakage. It is noted that the use of low wind speeds provide relatively small motive force for inducing leakage through the CCilE. Ilowever, parametric studies show that, over the range of interest, increased wind speeds will tend to iower Control Room dose when it is applied unilaterally to X /Q values and building differential pressure.
Other Pressure Effects Differential pressures across individual walls induced by differences in inside and outside temperatures (i.e., stack effect) provides an additional motive force not discussed in SRP 6.4 and not addressed solely by consideration of wind pressures. This effect can be pronounced in tall structures, and its magnitude is basically a function of the difference in temperatures across a wall and the difference in height from a given penetration to the building's neutral pressure level. Although the CCilE is relatively tall, the following considerations tend to minimize the effects of this phenomenon at CR-3:
- The temperature gradient between *e Control Complex and adjacent areas at the outset of an accident is relatively small, Given a source term model wherein the majority of exposure occurs during the initial stages of the event, differential pressures induced by the stack effect would by relauvely small over this critical period.
- The majority of CCilE penetrations are at or near the elevation of the cable spreading room, which is itselfjust below vertical center of the Control Complex clevation. (See Figures E-1 through E-9 for locations of penetrations.) The neutral pressure level of a building wall tends to be at the elevation containing the largest leakage area, or in the case of uniform leakage, at the vertical center of the building. Since the stack effect results in no appreciable differential pressure at the neutral pressure level and increasing differential pressures as distance from the neutral pressure level increases, the distribution of CCHE penetrations would tend to minimize the inleakage due to stack effect.
- U,. S. Nuclear Regulatory Commission Attachment E 3FJ 197-09 Page 5 of 13
- At higher wind speeds, *he differential pressure induced by wind pressure overcomes the stack effect, such that leakage across a given wall cannot develop b: tween the top and bottom of that wall.
Calculation of the stack effect is not included in current analyses, however, examination of the impact of this effect is ongoing.
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3F1197-09 P0ge 2 of 13 FIGURE - E-4 I
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