ML20236V880
| ML20236V880 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/30/1998 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20236V876 | List: |
| References | |
| NUDOCS 9808040263 | |
| Download: ML20236V880 (46) | |
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Contrcl Room Habitability Report Florida Power Corporation Crystal River 3 Control Room Habitability Report Florida Power Corporation Crystal River- Unit 3 i
l July 30,1998 9808040263 980730 PDR ADOCK 05000302 .
l P PDR
Control Room Habitability Report Florids Power Corporation
' Crystal River 3 i l
1 Table of Contents i i
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- 1. Introduction and Summary 3 II. Control Room Emergency Ventilation System Description 5 IT.1 General Description 5 11.2 Technical Specification Requirements 6
III. Control Complex Habitability Envelope 6 111.1 General Description 6 III.2 Requirements Regarding CCHE Inleakage 7 i III.3 CCHE Inleakage Determination 8 111.4 Test Conditions 9 IV. Control Room Modifications and Improvements 10 IV.1 Control Room Modifications 10 l IV.2 Control Rcom Improvements V. Radiological Evaluation 11-
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i 11 V.1 Background 11 ;
V.2 . Technical Specification Requirements 12 I
VI. Maximum Hypothetical Accident _ 13 l VI.1 General MHA Analysis Methodology 13 ;
VI.l.1 Source Terms 14 i VI.1.2 Release Paths .
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VI.1.3 Control Room Model 16 VI. l .4 Additional Information on Wind Induced Leakage 17 VI.1.5 Consideration of Thermal Stack Effect 18 VI.1.6 Inleakage Induced by CREVS Operation and Imbalances 19 VI.2 Mh1A Without Loss of Offsite Power 20 VI.3 Hydrogen Purge 22 VI.4 Summary of MHA Results 30 VII. Other Design Basis Accidents 31 VII.1 Letdown Line Break Accident (LLA) Analysis 31 Vll.2 Steam Generator Tube Rupture (SGTR) Analysis 33 VII.3 Fuel Handling Accident (FHA) Analysis 35 VII.4 Summary of Other Analyses 36 l VIII. Hazardous Chemical Evaluation 36
' References 38 L l Attachments 40 A. Diagrams 40 B. SGTR Time Line/ Curies Released 43 C. Periodic CCHE Integrity Test Method 45 D. Verification Review 47 L
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Control Room Habitability Report Florida Power Corporation Crystal River 3
- 1. Introduction and Summary The Crystal River, Unit 3 (CR-3) Operating License contains a requirement to maintain control room habitability as specified in the post Three Mile Island (TMI) requirements of NUREG-0737. . System Readiness Reviews conducted in 1997 identified several issues which potentially impacted control room habitability. The primary issue identified was the inability of the Control Cornplex Ilabitability Envelope (CCHE) to maintain unfiltered inleakage to a value less than utilized in the original Control Room Ilabitability analysis.
. A number of actions have been undertaken to address the concems and to significantly i irnprove the level ofprotection provided for the control room operator. These include: !
. Modifications to reduce CCHE inleakage by improving the integrity of boundary elements, e Design changes to the Control Room Emergency Ventilation System (CREVS) to provide altemate means of mechanical equipment room ventilation and to improve system reliability, and
- ' Programmatic changes to ensure that the assigned efficiency of the Control L Complex charcoal filters is consistent with regulatory guidance and to ensure ,
periodic leak testing of the CCHE boundary is performed. j Re objective of these actions was to modify both the CCHE and the CREVS to maintam i the design and licensing basis (i.e., result in a Control Room unfiltered inleakage rate that is bounded by the value used in the habitability analysis). The changes to the CREVS, which are described in Section IV, included installing redundant bubble-tight dampers at all l system connections that penetrate the boundary of the Control Complen and the elimination 1 of the system which cooled the Ventilation Equipment Room with outside air. The changes l to the CCHE included an extensive sealing program and the addition of vestibules over all l CCHE boundary doors.
After the modifications were completed, a tracer gas test was performed to determine the l inteakage rate. The results of the tracer gas test showed that, n spite of the extensive sealing effort and system changes, the boundary leaks on the order of 20% higher than the
- value used in the existing design and licensing basis habitability analysis. Even though the measured inleakage is above the original value, the integrity improvement actions are categorized as successibl as they resulted in leak rates that ensure that calculated doses are L within NUREG 0737 guidelines (which are equivalent to 10CFR50 Appendix A GDC19 guidelines) for all analyzed accidents. Means to further reduce the remaining inleakage (a large portion of which is through the interstitial spaces in cable tray penetrations) are being evaluated.
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Control Room Habitability Report Florida Power Corporation '
Crystal River 3 Since the unfiltered inleakage rate exceeded the design and licensing basis value, the control room operator dose calculations were revised to align inputs and assumptions with plant design. The basic methodology used in these revised calculations is consistent with that found in regulatory guidance. However, the determination of CCHE inleakage and the application ofinleakage in dose calcuhtion differs significantly from previous
. methodologies. These differences have been determined to constitute an Unreviewed Safety Question (USQ), as stated in an informational report to the Nuclear Regulatory Commission on the subject of contml room habitability dated November 10,1997.
A Justification for Continued Operation (JCO) was prepared to address the safety significance of this USQ and ascertain the acceptability of restart in the interim per the guidance of Generic Letter 91-18, Revision 1. The specific issues addressed in the JCO are:-
e the operability of the CREVS e the integrity of the CCHE.
The JCO concludes that the CREVS operability and the CCHE integrity are adequate to:
e protect the control room operator in toxic gas events and Design Basis. Accidents (DBAs) for which CR-3 is licensed, such that regulatory limits are not exceeded e meet operability requirements defined in the Improved Technical Specifications (ITS) e not invalidate the assumptions and conclusions of the ITS bases.
Meeting these requirements was deemed adequate and appropriate basis for plant operation in any mode with regard to control room habitability.
Based on the revised design, the measured inleakage rates, and comments from the NRC, all of the control room doses were recalculated. The Maximum Hypothetical Accident (MHA), which is the large break LOCA with TID-14844 source terms, with no Lass of Offsite Power (LOOP) resulted in the highest calculated dose, but was still within the guidelines of 10CFR50 Appendix A, GDC19. Doses were also calculated for the Letdown Line Break, Stea.n Generator Tube Rupture (including a beyond licensing basis SRP scenario), and the Fuel Handling Accident. J f
The purpo's e of this control room habitability report is to present: )
e the latest design of CREVS and the CCHE e the inleakage measurement technique that is used and the means by which it is l~ converted to an inleakage rate for calculations l- e the methods, assumptions and results of the revised dose calculations I e the improvements made to toxic gas inventories and controls I
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f Control Room IIabitability Report Florida Power Corporation Crystal River 3 The report is being submitted with a Technical Specification change request. The Technical Specification changes support the assumptions made in the dose calculations.
( Upon NRC approval of the license amendment request, the JCO will be cleared.
1 This report will be a living document and will be referenced in the UFSAR. If any changes are made to the plant design or procedures that affect the assumptions or evaluations in this report, the report will be revised. Since this report will be referenced in j the UFSAR, any changes to this report will require an Unreviewed Safety Question l determination per 10CFR50.59.
II. Control Room Emergency Ventilation System Description 11.1 General Description
! A simplified schematic of the CREVS is provided as Attachment A. The CREVS consists L of two independent safety-related air recirculation trains that, in addition to the cooling l capability, have the ability to divert 100% of the recirculation flow through an l Emergency Filter Unit. Each Emergency Filter Unit contains, in the direction of flow, a
- i. roughing filter, a HEPA filter,2-inch activated carbon filter bank, and a safety-related recirculation booster fan. The Emergency Filter Units do not include means to lower the humidity of the air as it enters the adsorber bank, such as electric heating coils. Heaters
! are not required since the system only operates in a recirculation mode. The emergency recirculation fans are powered by separate safety-related power sources. The CREVS processes and filters air from the top five levels of the control complex.
Upon detection of either high reactor building pressure or high radiation in normal l Control Room ventilation ductwork, as detected by RMA-5, the redundant, bubble-tight l' boundary isolation dampers are automatically closed. The operation of the emergency l' fans and filters are manually initiated by the operator. The calculations conservatively assume 30 minutes for the time of manual initiation of the filters.
l~ The redundant, bubble-tight boundary isolation dampers are also automatically closed as
! a result of a Loss of Offsite Power (LOOP). The operation of the CREVS during a LOOP l is manually initiated by the operators. Although the diesel loading would allow starting l of the CREVS well before 30 minutes, the operation of the CREVS is conservatively not credited for 30 minutes.
The CREVS provides environmental control for personnel comfort and equipment operation and protection of control room personnel during radiological and toxic gas events. It provides habitability via zone isolation with filtered recirculation. The control ;
complex is not pressurized to limit inleakage. Leak 1:ghtness and filtration capability )
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Control Room IIabitability Report Floridz Power Corporation Crystal River 3 provide the necessary level of protection for the control room operator to ensure that exposure limits associated with DBAs and toxic gas events are not exceeded.
II.2 Technical Specification Requirements The Improved Technical Specifications (ITS), prior to the reassessment of the Control Room Habitability issue, required that ITS Sections 3.7.12 and 5.6.2.12 be satisfied and that the bases for these technical specifications not be invalidated. The Surveillance Requirements to demonstrate CREVS operability, included the following criteria:
e operating each CREVS train for at least 15 minutes each month e satisfying the CREVS ventilation filter testing program e verifying that each CREVS train actuates to the emergency recirculation mode on an actual or simulated actuation signal every 24 months.
The second criterion refers to the requirements of the ventilation filter test program defined in ITS Section 5.6.2.12. ITS Section 5.6.2.12 defines requirements pertaining to the Ventilation Filter Testing Program at CR-3 and requires that the CREVS filtration units meet minimum performance standards regarding penetration, bypass, and adsorption. This program requires that in-place testing be conducted which verifies the performance of the CREVS filtration system at a flow rate of 43,500 cfm +/- 10% and that the pressure drop across the filtration unit be less than 6" water gauge when operating in this range. A License Amendment Request was submitted in December 1997 to change the filter testing to meet the 1989 ASTM D-3803 protocol and to revise the CREVS filtration system flow rate from 43,500 cfm +/- 10% to a range from 37,800 to 47,850 cfm. The amendment request also added the requirement to perform a periodic test of the CCllE integrity once per operating cycle. These proposed changes are included with the revised license amendment request submitted with this habitability report.
l Additionally, the revised license amendment request includes criteria for periodic tests of the Auxiliary Building Ventilation Exhaust System Filters and an allowance for short term small breaches in the CCHE.
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, III. Control Complex Habitability Envelope l
l IH.1 General Description The Control Complex is a six floor building located between the auxiliary building and the turbine building as shown in Attachment A. The Control Complex Habitability ;
Envelope is the top five floors of the Control Complex. The lower floor is isolated from l the CCHE under accident conditions. The top floor of the CCHE contains the control complex ventilation equipment, thus it is all internal to the CCHE. The control room is one floor below the ventilation equipment room. The CCHE, along with CREVS are designed to protect the operator in case of a radiological or toxic gas release.
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l Control Room Habitability Report Florida Power Corpor tion Crystal River 3 As is noted in the diagram in Attachment A, the only open surfaces to the environment are the upper levels of the east and west walls, which have no penetrations, and the roof.
The north wall adjoins the turbine building and has a number of penetrations through the CCllE and the south wall adjoins the auxiliary building and contains a number of penetrations. Thus, the CCHE is not subject to significant inleakage due to wind induced effects. This is due both to the fact that it is fairly well surrounded by higher structures which should minimize wind loading and because the open surfaces contain a limited number of penetrations. Most inleakage would be expected to occur from penetrations in the turbine building wall due to negative pressure in the auxiliary building providing a motive force for flow from the turbine building through the CCllE and the into the auxiliary building.
The detemiination of the unfiltered inleakage rate used in the CR-3 NUREG-0737, Item Ill.D.3.4 Control Room Habitability Evaluation Report, dated June 30,1987, was based on calculations. The unfiltered inleakage rate was calculated, for the neutral control complex, at 236 cfm. However, as discussed in the 1987 submittal, an unfiltered inleakage rate of 355 cfm (equal to 0.06 volume changes per hour) was used to avoid periodic testing (SRP 6.4 states that periodic testing should be performed if the assumed inleakage is less than 0.06 volume changes per hour).
The determination of the unfiltered inleakage presented in this Control Room Habitability Report was done by a rigorous test program using tracer gas technology. The improvements to the CCHE were completed or simulated prior to testing, such that the test reflects the current CCHE integrity. In addition, the CREVS was modified to improve the integrity of the CCHE (the changes to the CREVS included installing redundant bubble-tight dampers at all system connections that penetrate the boundary of the Control Complex and the elimination of the system which cooled the Ventilation Equipment Room with outside air). The changes to the CCHE also included the addition of vestibules over all CCHE boundary doors.
HI.2 Requirements Regarding CCilE Inleakage l
The CCHE is the physical barrier which separates the control room environment from the external environment. The integrity of these barriers (walls, doors, ceilings, floors, sealed penetrations, ventilation penetrations, etc.) directly affects the inleakage of radiation sources associated with Design Basis Accidents (DBAs). The ITS bases state that j CREVS ensures that the control room will remain habitable, following all postulated l design basis events, by maintaining exposures to control room operatocs within the limits of General Design Criteria (GDC) 19 of 10CFR50, Appendix A. There is currently no reference or commitment in the ITS or its Bases to Standard Review Plan ('SRP) 6.4,
" Standard Review Plan for Control Room Habitability." Also, there are no specific 7
Control Room Habitability Report Florida Power Corporation Crystal River 3 requirements regarding CCHE inleakage included in the current ITS or its Bases. Since the integrity of the CCllE has a direct influence on the ability of the CREVS to maintain I a habitable environment, there is an implicit requirement that the integrity of the CCHE be demonstrated as adequate to support CREVS in maintaining operator exposure within regulatory limits.
Therefore, the proposed license amendment adds the requirement to perform a periodic ;
test of the CCHE integrity once per operating cycle. Attachment C provides a description of the integrity test which will be performed.
HI.3 CCHE Inicakage Determination The infiltration of unfiltered air into a control room boundary is typically modeled as three paths: (1) through the zone boundary; (2) through the system components located outside of the emergency zone; and (3) through backflow at the zone boundary doors as a result j of personnel access or egress. ;
i The methodology used to determine zone boundary leakage at CR-3 is discussed below.
With respect to component leakage, all of the CREVS components and ductwork at CR-3 )
are contained within the emergency zone and as such do not contribute as an inleakage source. The leakage through the bubble-tight isolation dampers, which are installed in all CREVS boundary penetrations, were tested and are included in the total inleakage. With respect to backflow through the zone boundary doors, it is conservatively assumed that an additional 10 scfm ofinfiltration is induced by the opening and closing of doors, as recommended by SRP 6.4.
The integrity of the CCHE used in the radiological analysis was based on the "as-left" measured performance of the CCHE. The boundary performance was measured using tracer gas tests that were performed by a team of engineers from FPC and Lagus Applied Technology. The tracer gas test procedure was based on ASTM Standard E741-93,
" Standard Test Method for Determining Air Change Rate in a Single Zone by Means of a Tracer Gas Dilution." The tests were accomplished using an electronegative gas, sulfur hexafluoride (SF6), as a tracer. This gas was utilized since it is recognized as non toxic, non reactive, inert, and easily detectable in minute quantities by means of electron capture gas chromatography.
Although not prescribed by 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 purpose. By using tracer gas test methods, it is possible to measure inleakage under conditions which are representative of a specific postulated scenario. This methodology is expected to provide a more accurate prediction of inleakage.
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Control Room Hr.bitability Report Floride Power Corporation Crystal River 3 Tracer gas testing under post-accident conditions requires the Control Complex to be in its emergency recirculation rnode and treating the entire CCHE as a single volume. An additional penalty is not required at boundary damper locations since the dampers would be subject to the same pressures during testing as would be expected during post-accident operation. The additional penalty is inconsequential for CR-3 since two bubble-tight dampers have been installed in series at each boundary isolation location. The installed ;
dampers were factory tested to be bubble tight. Post installation testing demonstrated the dampers to have inconsequential leakage. Testing was performed by pressurizing i between each pair of series dampers; therefore, one damper was pressurized as it would be in service, in the closing direction while the other was pressurized in the opening direction. l l
Developing a test replicating post-accident conditions required that the scenario be postulated which provides both realistic conditions from the standpoint of exposure to source term and maximizing the differential pressure which drives inleakage. The l objective was to develop a test which determines inleakage as a function of pressure
[ differential. The inleakage for other differential pressures and modes of operation can
! then be determined by analysis using the empirically derived data.
The motive forces, which might induce a significant differential pressure across the CCHE, are taken as:
e the wind pressure (assuming a loss of offsite power (LOOP)) and l
e the running ventilation system in the adjoining structure (no LOOP).
l l A significantly higher differential pressure would be expected, assuming no LOOP, l however, the source term would be lower as in this scenario the Auxiliary Building l Filtration System would be in operation. Thus, to fully assess limiting post-accident conditions required that both scenarios be evaluated. This was accomplished by measuring inleakage at a known differential pressure using tracer gas methods, then l analytically adjusting the resulting values to correspond with postulated conditions.
!; III.4 Test Conditions
(' The following conditions were prescribed for the tracer gas testing which modeled CCHE l inleakage:
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- The CREVS was placed in emergency recirculation mode. Both " Toxic Gas" and the "High Radiation" recirculation lineups were tested. All CREVS I
boundary damper locations were sealed to duplicate post modification conditions.
(NOTE: The isolation dampers were tested in-place and the test boundary exhibited insignificant leakage.)
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Control Room Habitability Report Florida Power Corporation Crystal River 3
- All fans in the Turbine Building Ventilation System (TBVS) were secured. The turbine building normally remains well vented to atmosphere through normally open doors, roll out windows, and roof vents.
- All fans in the Intermediate Building Ventilation System were secured. Note that conditions in the Intermediate Building are not deemed critical to the test in that relatively few penetrations are on the CCHE / Intermediate Building Wall.
. The Auxiliary Building Ventilation was operated at a test pressure of l approximately 0.171" water gauge negative pressure vs. the Turbine Building pressure. This va'ae is large enough to minimize test inaccuracies and external effects; the pressure was sustained for the duration of the test.
. The test was conducted on backshift when personnel traffic is minimized. Since a 10 cfm allowance for access / egress is added to measured inleakage applied, 1 minimizing traffic precludes counting this effect twice.
. Testing was conducted with vestibule doors blocked open. This conservatively assumes no credit for the additional integrity provided by vestibules.
- All loop seals penetrating the CCHE were verified to be filled prior to testing.
Controls are in place to ensure that these loop seals are maintained full during plant operation.
IV. Control Room Modifications and Improvements IV.1 Control Room Modifications The following modifications were implemented, during the 1997 outage, to address the concerns associated with Restart Issue R-12. The post-modification flow diagram is shown in Attachment A. 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.
. Damper AHD-99, which brought supply air to the Ventilation Equipment Room L was removed and a permanent blank installed. New supply and return registers were installed in the ductwork (164' elevation) which will now serve as the ventilation supply air source for this area. This modification eliminated AHD-99 as a potential source ofinleakage.
- Existing damper AHD-12, located in the supply duct to the Control Area, was removed and replaced with two new bubble tight dampers, AHD-12 and AHD-12D.
. Existing damper AHD-2, located in the exhaust duct to the outside, was locked open and abandoned in place. Two new bubble tight dampers, AHD-2C and 10 I
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Control R om Habitability Report Florida Power Corport. tion Crystal River 3 l AHD-2E, were installed in series in the exhaust path. AHD-2C will be normally closed.
l e Dampers AHD-1 and AHD-1D, located in the air intake duct, were disabled and i abandoned in place. Two new bubble tight dampers, AHD-lC and AHD-lE, were installed in series on the inlet duct. Dampers are provided with a manual j override feature. This feature allows operators to modulate the recirculation airflow as required for purging smoke or other contaminants from the CCHE.
. Mechanical Equipment Room Ventilation Air Handling Fans, AHF-21 A/B and associated dampers AHD-24, AHD-25, AHD-26, and AHD-27 were spared in place and the associated CCHE penetration sealed. This portion of the system originally exhausted air from the Mechanical Equipment Room, Elevator Equipment Room, lavatory, kitchen, and toilet. This additional modification eliminated another potential source ofinleakage into the CCHE.
. New supply and return registers were installed in the ductwork in the Mechanical Equipment Room. This provides ventilation to this portion of the CCHE, during both normal and recirculation modes.
- Small bore drain pipes penetrating the CCHE were fitted with loop seals to
- prevent inleakage though the lines. These were added to a queued work request i
in the work contml system which maintains CCHE drain line loop seais.
e Vestibules have been installed over all CCHE boundary doors and have been i sealed to provide maximum leak tightness. These vestibules provide a means to reduce inleakage associated with CCHE access / egress.
I IV.2 Control Room Improvements In addition to the above raodifications, an extensive effort was undertaken to survey CCHE penetrations and seal each of them, as required, to minimize inleakage. As a result of this work, conduit penetrations do not pose a significant leakage path through the CCHE. Penetrations associated with electrical cable trays were inspected and sealed to l the extent feasible with existing procedures and materials, but some leakage paths remain j through the interstitial spaces between individual cables. Additional work is planned to
- improve the sealing of those penetrations with the most significant leakage. )
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l V. Radiological Evaluation V.I Hackground 11 L_______ __ ____ _ . _ _ _ _
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l Control Room Habitability Report l Florido Power Corporation l Crystal River 3 The control room (CR) dose requirements are established by GDC-19 of Appendix A to l 10 CFR 50. The requirements of GDC-19 became a design basis for CR-3 as a result of i CR-3's commitmem to NUREG 0737, item Ill.D.3.4. This post-TMI action item l required a review of the CR ventilation system, which included an assessment of the system's ability to adequately protect the CR operators against the effects of accidental release of radioactive gases and verification that the plant can be safely operated or shut l down under design basis conditions, as required by GDC-19. In response to requirements pertaining to NUREG-0737, Item III.D.3.4, FPC performed a comprehensive habitability evaluation of the CR-3 Control Room and submitted its findings in the form of the revised CR-3 " Control Room Habitability Evalu: lion Report," dated June 30,1987.
Based on methodology which was considered consistent with SRP 6.4 guidance, the 1987 i
habitability evaluation found that the Maximum Hypothetical Accident (MHA) would l result in a thyroid dose of 26.5 REM.
i For the original radiological dose submitted in 1987, the most limiting accident with regard to control room habitability was determined to be the MHA As part of the effort for this revised habitability report, FPC has reviewed other design basis accidents that could potentially affect CR dose. The other accidents reviewed include the Fuel Handling Accident (FHA), the Letdown Line Break Accident (LLA) accidenh and the Steam Generator Tube Rupture (SGTR) accident. The assessment of the other accidents was performed to confirm that the MHA is the most limiting accident and as such was performed using the conservative NUREG-0800 approach as opposed to the methodologies currently described in the CR-3 UFSAR. In the case of the SGTR accident, the NUREG-0800 approach is significantly different than the CR-3 licensing basis accident analysis. Since CR-3 is not an SRP plant for this accident, the SGTR accident analysis scenario and methodology presented in this report, is not considered new licensing basis. The SRP calculation was only performed to demonstrate that the
!_ MHA would remain bounding even if NUREG-0800 SGTR assumptions were considered.
The Control Room Habitability analyses summarized in this report supersede the anaiyses l contained in previous submittals.
V.2 Technical Specification Requirements There are currently requirements for CREVS operability in the ITS which are taken credit for in the radiological calculations. However, there are currently no specific requirements regarding CCHE inleakage or integrity testing included in either the ITS or its Bases.
Such a requirement is proposed in the license amendment request submitted with this report. A discussion of the integrity test method to be used is included as Attachns nt C to this -eport.
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- 1 Control Room Habitzhility Report l Florida Power Corporation Crystal River 3 VI. -Maximum Hypothetical Accident l
l The existing licensing basis dose analysis has been examined and compared to NRC l
licensing guidance (primarily Regulatory Guides 1.4, and SRP 6.4,15.6.5) to confirm the i acceptability of the radiological analysis input parameters and methodology. Changes l have been made to the existing input parameters to correct deficiencies, to conform with NRC guidance, and to utilize acceptable methodologies to predict accident doses. The l
l significant changes involve the following: l t
- . utilized unfiltered inleakage rate based on results of tracer gas testing l
e since the Control Complex is a " neutral" design, considered that during an MHA with a LOOP, the wind pressure is the primary motive force that could induce j inleakage (resulted in aligning the inleakage rate with accident analysis) l l . considered other operating scenarios, such as an MilA without a LOOP, that could result in higher inleakage rates to the Control Complex due to the effects of adjoining buildings ventilation systems e added credit for ICRP-30 dose conversion factors
. included the direct dose contribution from activity buildup on the control room
! ventilation system filters - this was added to the direct dose contribution from i other sources such as the plume and the containment which had already been !
calculated but were recalculated based on the revised MHA assumptions The assumptions and input parameter values for the CR dose calculations for CR-3 are summarized in Table 1. Table 1 identifies the current licensing basis parameter, the l revised parameter and the basis for the parameter. The new methodology incorporates l l several less restrictive assumptions along with a number of more restrictive assumptions. l
! The net affect, however is a new calculation methodolo;;y that is still acceptably l conservative. Table 1 identifies some of the examples of the conservatism in the revised analysis. Key assumptions and input parameter values are described below, along with a l discussion of the principal quantitative and/or qualitative conservatism, as appropriate.
These conservatism are in addition to the extremely conservative methodology developed in TID-14844 to assess the adequacy of proposed reactor sites.
VL1 GeneralMHA Analysis Methodology The ar.alysis methodology for a design basis MHA done in accordance with the Regulatory Guides and Standard Review Plan can be segregated into three distinct areas:
. Source Terms
. Release Paths
. Control Room Model 13 I
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Control Room liabittbility Report Florida Power Corporr. tion Crystal River 3 For the CR-3 analysis, it was recognized that both the release paths and the control room model would be different depending on the availability of offsite power during an MHA.
A detailed analysis was prepared for both scenarios since the bounding scenario could not be determined qualitatively.
In previous submittals, the most limiting radiological consequences were assumed to occur with a LOOP coincident with a LOCA, as this was the standard assumption for public dose calculations. During this scenario, the dose contribution from the leakage in the Auxiliary Building is maximized since it neglects the filtration of the effluent released from the Auxiliary Building. The Auxiliary Building ventilation system is designed to maintain the building at a negative pressure a to includes a charcoal adsorber unit in the building exhaust stream. This filter unit was not credited in the MHA with LOOP j analysis since it is not powered from a diesel backed power source. Ilowever, durin<; this scenario, the inteakage to the Control Complex will be less since all of the buildings that surround the Control Complex will be at atmospheric pressure. In this scenario, the inleakage will primarily be based on the wind pressure, with a lesser contribution from th:rmal stack effect and local system imbalances.
On the other hand, for an MHA without a LOOP, the Auxiliary Building fLns could be in operation which would result in significantly higher inleakage to the Control Complex due to the fairly high negative pressure maintained in the Auxiliary Building. If the Auxiliary Building Ventilation Exhaust System (ABVES) is in operation, a constant i' inleakage path would exist with the activity entering the Turbine Building where it would leak into the Control Complex and would then leak from the Control Complex to the Auxiliary Building. During this scenario, the activity released in the A.uxiliary Building would be filtered prior to being released to the environment (the ABVES is seismically qualified and would be intact fbliowing an MHA). The Control Complex is located l between the Auxiliary Building and the Turbine Building, so that the worst alignment would entail the Auxiliary Building being et a negative pressure compared to the Turbine Building. In order to ensure credit for filtration when the ABVES is operating, the License Amendment Request submitted with this report includes ABVES filter testing criteria.
VL1.1 Source Terms As shown in Table 1, the analysis was performed using the source terms from the US Atomic Energy Commission Technical Information Document (TID) 14844 at a power level of 102%. The power rating used in the analysis is actually 102% of the tentative power uprate value. This power uprate action has not been completed, but the post accident source term associated with the higher power rating has been incorporated into dose calculations. Since the source term is determined based on a per megawatt basis in 14 w__________________.
Control Ro:m Habitchility Report FloridI Power Corporation Crystal River 3 accordance with TID-14844, the use of the larger MWth rating results in a source term slightly higher than that which would be predicted with the lower power rating. This is a conservative position since the plant is currently licensed to the lower power rating.
The removal mechanisms of the activity in the containment considered the initial plateout of 50% of the airborne iodine, spray removal for iodine, mixing between de sprayed and unsprayed regions, decay, and activity leakage. Based on NRC comments on the JCO, the analysis was revised. to use a maximum elemental spray removal rate of 20 per hour.
VLI.2 Release Paths The MHA radiological analyses consider the following three release paths:
. Containment Leakage
. Continuous Emergency Safeguards Features (ESP) Fluid Leakage Outside of Containment
. Short Term ESF Fluid Leakage Outside of Containment due to Component Failure (50 gpm teak for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the Accident per the SRFj For the Containment leakage path, the analysis considered a leak rate at the value incorporated in the Technica; Specifications for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (0.25% per day) and a leak rate of 50% of the value incorporated in the Technical Specification for the next 29 <
days. The source term for the ESF fluid leakage was conservatively based on 50% of the core iodine activity being initially mixed in the containment sump and available for release.
For the MHA with a LOOP analysis, all of the leakage paths were treated as unfiltered ground-level releases. For the MHA without a LOOP analysis, the ABVES was conservatively credited with an efficiency of 75% for all forms ofiodine (the proposed license amendment request requires that the ABVES be tes:ed at levels that would justify this removal rate). For the MllA without LOOP analysis, the analysis does not consider the failure of an ESF fluid component at 50 gpm for 30 minutes at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident since the ABVES is in operation. This is consistent with the SRP which states that the radiological consequences of an assumed 30 minute leak do not have to be evaluated if the leak is into a filtered secondary building.
The release path for the MilA without a LOOP scenario is based on the activity being released from the Containment and ABVES as a ground level release and subject to initial dispersion as it travels to the Turbine Building Ventilation System intakes and into the l
Turbine Building. From that point, the released activity ultimately enters the control room as unfiltered inleakage as a result of the differential pressure induced across the Control Complex by the ABVES. This release path model accounts for the time l
15 i
I L _. _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
, Control Room II bitability Report Florids Power Corporation l l Crystal River 3 dependent rate of change of activity in the Turbine Building volume, which includes a minor effect due to the holdup of the activity as it passes through the Turbine Building.
VI.l.3 Control Room Model i The CREVS provides protection to the operator from the consequences of design basis accidents through zone isolation and filtered recirculation. As previously discussed, the Control Room model is also affected by the availability ofoffsite power. During a LOOP, j the " neutral" Control Room was assumed to leak primarily as a function of the wind speed.
The analysis also considered the possibility of additional leakage caused by the thermal stack effect and as causea by local system imbalances. During an MHA without a LOOP, the Control Complex is assumed to leak at a constant rate due to the negative pressure in the Auxiliary Building.
For both scenarios, the analysis credits the intake being automatically isolated at the time of I the accident due to the high reactor building pressure signal, and that the filtered recirculation system is manually staned at 30 minutes after the accident.
Integrated inleakage testing has been performed to provide an assessment of CCHE integrity. This integrated testing used tracer gas methods to directly measure building inleakage while operating in the post-accident alignment. To calculate dose based on test results, FPC developed a model which predicts inleakage under various differential pressure conditions which was used to mechanistically determine the inleakage under postulated post-accident conditions. The measurement ofinleakage and n mechanistic application in dose calculations is a significant change from previous methodology and is expected to produce a result which is more realistic.
Previous dose calculations were based on a model defined in SRP 6.4 which determines inleakage at a Control Complex pressure of 1/8" water gauge, then applies 50% of this inleakage as a constant value in dose calculations. This model stemmed from the common control room designs which rely on takine contaminated air from outside the envelope, filtering it, and then using this filtered air to pressurize the control room. In this case, pressurization of the habitability envelope to a nominal value of 1/8" water gauge is used to prevent inleakage and is generally accepted as being high enough to overcome the effects of wind, thermal effects, and operation of ventilation systems in adjacent structures. This model does not correlate well to the CR-3 habitability envelope, which is a filtered recirculation system with isolation (no makeup).
FPC used a more realistic, but still conservative, model for determining CCHE inleakage
, for the purpose ofinput into control room operator dose calculations. This model is based f on considerations of the actual motive forces which would exist for driving inleakage under
( postulated post-accident conditions. The use of this model is a departure from the previous methodology and is not described in regulatory guidance. As such, the application of this 16
l .
l Control Room Habitability Report Florids Power Corporation Crystal River 3 l methodology needs to be reviewed by the NRC. As a part of the restart effort, FPC l presented this methodology to the NRC in public meetings and through the restart JCO.
l As detailed within the JCO, FPC and its contractors have reviewed the treatment of CCHE
! inleakage in control room habitability calculations and concluded that the change in l methodology is safe, that CREVS operability is not compromised, and that CCHE integrity l is maintained.
i The tracer gas test model was established based on consideration of site layout, source l terms, and possible plant operating conditions. In the event of a MHA:
e With a LOOP, the ventilation systems in all of the areas adjacent to the Control Gomplex will be out-of-service. During this scenario, given that the vast majority of penetrations are either on the north (Turbine Building) or south (Auxiliary Building) walls of the control complex, it follows that north or south wind directions would tend to maximize CCHE inleakage.
- Without a LOOP, the Auxiliary Building supply fans would be secured by radiation moritor RM-A2, causmg the Auxiliary Building to develop a negative pressure and induce leakage through the CCHE in that direction.
In either case, the tracer gas test (which models these conditions) would utilize the Auxiliary Building Ventilation Exhaust System (ABVES) to induce CCHE inleakage by creating a negative pressure in the Auxiliary Building.
Tracer gas testing conducted under these conditions measured an inleakage of 462 cfm in "High Radiation" recirculation mode at a differential pressure of 0.171" water gauge.
Using this information, the inleakage at other differential pressures can be predicted by the use of the equation: ;
Q = C P" !
where:
Q = air flow in efm C = inleakage coefficient l P = differential pressure, and n = flow exponent.
According to ASHRAE, values of n for building penetrations are typically between 0.6 and 0.7. The extrapolation to differential pressures less than the test pressure (i.e., 0.171 water gauge) were estimated by using n = 0.5, which gives conservative results for estimating inleakage at pressures less than the test value. On the other hand, the l extrapolation to differential pressures above the test pressure (i.e.,0.171" water gauge) were estimated by using n = 0.65, which gives more realistic yet conservative results for estimating inleakage at pressures greater than the test value.
VI.l.4 AdditionalInformation on Wind Induced Leakage 17 l
Control Room II:bitability Report Florids Power Corporation Crystal River 3 Since wind pressures are assumed to be the primary motive force under MHA with LOOP conditions, inleakage for this scenario is determined from meteorological conditions associated with event analysis. 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 i
from the containment building to the Control Complex. The methodology then allows for three incremental increases in wind speed and directional variance over the duration of the accident due to the extreme improbability that these initial wind conditions would be sustained over an extended period of time. Based on these considerations, inleakage i values are derived for each of the four time intervals over which x/Q values vary by l correcting inleakage at the test differential pressure to the differential pressure induced by the wind speed associated with that interval. These wind induced differential pressures
( were conservatively calculated using ASHRAE methods. Each of these inleakage values l 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. j i
(
l For the MHA with LOOP, it is noted that the use oflow wind speeds provides a relatively '
small differential pressure for inducing leakage through the CCHE although the dispersion at low wind speeds is at a minimum. At higher wind speeds the differential ;
l pressure and thus the inleakage is increased. The increased inleakage is offset by the increased dispersion of the higher wind speeds. A parametric study showed that, over the range ofinterest, increased wind speeds will tend to lower control room dose when it is applied uniformly to both x/Q values and building differential pressure.
VI.1.5 Consideration of Thermal Stack Effect At relatively low wind speeds, the effects of thermally induced inleakage becomes
! significant. Differential pressure across walls that result from differences between inside and outside temperatures (i.e., stack efTect) can be pronounced in tall structures. The magnitude is a function of the difference in temperatures across a wall and the difference in height from a specific penetration to the building's neutral pressure level. This effect is quantified and combined with the wind driven infiltration using the methodology L established by ASHRAE. The following table lists the parameters used in calculating the I l " stack effect".
1 I
i l Value l Basis l Conservatism Winter Conditions Control Envelope 75"F hot weather design
, Temperature temperature Exterior Temperature 31 F 99% Design Dry Bulb occurs for only 88 including Adjacent Buildings (DDB) Temperature for hours annually (1%
l
! 18
t l Control Room IIabitability Report Florida Power Corporation Crystal River 3 Ocala, Fla. of the time), whereas this scenario maintains it for 720 l hours.
l Summer Conditions l Control Envelope 75 F hot weather design i
Temperature temperature Exterior Temperature 118 F Turbine Building Auxiliary building is including Adjacent Buildings summer transient at a lower L temperature at the start temperature. The of LOOP is 20 F temperature was higher than the summer used for the entire 30 time outside design days of accident temperature of 95 F l (i.e. I15 F) l l The winter conditions were used because they result in higher differential pressure and
! inleakage principally due to the higher air density. Such conditions result in an effective thermal pressure differential of 0.0237 inches of water. The inleakage associated with l this value was determined by application of relationship between building differential pressure and inleakage derived from tracer gas test results. This value was then combined l with wind induced inleakage.
f l VI.l.6 Inicakage Induced by CREVS Operation and Imbalances l The CCHE is a five story structure. The control room is located on the fourth floor. The CREVS recirculation filter system is located on the top floor. Return ducts draw air from all five of the floors. Although supply air is provided to each flocr, substantially more supply air is typically supplied to the continuously occupied control room area of the
' fourth floor such that the air supplied to the control room area is essentially filtered air and creates a localized region of positive pressure. No credit was taken for the filtered positive pressure in the control room in the radiological calculations.
Localized areas of negative pressure could develop where the return flow exceeds the supply flow in the building. This localized negative pressare can induce leakage from the external environment or other floors inside the CCHE, especially at periods oflow wind speed. Any inleakage that occurs due to localized negative prc sures would flow directly to the return ducts and would pass through the CREVS charcoal fihcr prior to returning to an occupied area. Thus the induced in leakage is filtered prior to entering an occupied area.
- A direct determination of the amount ofleakage that could be induced by localized negative pressures would be difficult to achieve and has not been done. In the tracer gas 19
1 -
l Control Room Itabitability Report l Florido Power Corporation Crystal River 3 testing of the CCllE, the CREVS was operating. The effect of any CREVS induced leakage was included as part of the total 462 cfm at 0.171 in. water gauge of measured inleakage.
Total = Differential + CREVS Induced Inleakage Pressure Inleakage Inleakage The leakage resulting from CREVS operation is independent of the other CCHE inleakages, since it is a characteristic of the ventilation system. The maximum possible CREVS induced inleakage would result if all of the inleakage measured during the testing were the result of CREVS operation. (i.e.,462 cfm).
As modeled by the Murphy-Campe guidance, the CREVS induced inleakage is analogous to return ductwork inleakage and is filtered by the CREVS before it contributes to the control mom dose. Therefore the maximum possible effective contribution to the thyroid dose, in terms of unfiltered inleakage, from the CREVS operation using the 95% filter effectiveness of the CREVS charcoal filter would be:
462 cfm x 0.05 = 23.1 cfm Some location specific mixing of the CREVS operation induced inleakage within the CCHE could occur prior to it being filtered. This would have an increased impact on the operator dose. Therefore, to address this effect and produce a more conservative dose assessment, a value for unfiltered inleakage due to CREVS operation was assumed to be 125 cfm, approximately 25% of the total inleakage measured by the testing.
This value was directly added to the wind induced inleakage in the dose calculation (i.e.,
the inleakage determined from the total inleakage measured during the test). In determining the wind induced inleakage, the total inleakage during the test was not reduced by the assumed CREVS induced inleakage. This approach ensured that the CREVS operation impact is bounded by the dose calculation.
VI.2 MIIA without Loss of Offsite Power Given the occurrence of the MilA without a LOOP, the ventilation systems in adjacent
! buildings are assumed to continue to operate during and after the accident. Increasing I
levels of radiation in the Auxiliary Building, as sensed by radiation monitor RM-A2, would result in a trip of the ABVS supply fans; thus, resulting in a significantly larger i negative pressure in the Auxiliary Building. The Turbine Buildmg is considered to be essentially at atmospheric pressure due to the numerous large openings in that structure.
l l 20 l l
w__ _ _ __ _
l i Control Room Hnbit:bility Report Florid: Power Corporation
- Crystal River 3 l Under these conditions, the post-accident leakage into the control room could be higher I
(especially during the 0-8 hour and 8-24 hour time steps when wind speed is postulated to be low) than that determined on the basis of wind pressures (i.e., MIIA with LOOP).
The release path for this scenario is based on the activity being released from the Containment and subject to initial dispersion as it travels to the TBVS intakes and into the Turbine Building. From that point, it ultimately enters the Control Complex as unfiltered inleakage driven by the differential pressure induced across the Control !
Complex to Turbine Building boundary. This release path model accounts for the time dependent rate of change of activity in the Turbine Building volume, as well as minor I decay and holdup while the activity is in the Turbine Building.
The evaluation of MHA without LOOP has five distinct changes from the version of the event which assumes LOOP; e given that the ABVES must be in operation to induce the differential pressures of concern, then filtration by the ABVES charcoal filters occurs and, per the SRP, !
there is no requirement to assume an ECCS pump seel failure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after )
the accident with a leak rate of 50 gpm for 30 minutes, e the normal ECCS leakage which does occur is assumed to be filtered at 75%
efficiency, '
. the activi'y will enter the control room via the Turbine Building and as such will be subject to some delay due to the buildup and decay in the volume of the Turbine Building, and e inleakage will be constant for the duration of the accident and will not be affected by the wind speed used in the dose analysis. This is conservative in that the wind direction necessary for transport towards the Turbine Building would tend to oppose inleakage through the CCllE towards the Auxiliary Building.
. inleakage induced by CREVS operation does not have to be included as an addition to the wind induced inleakage, as it is already included in the measured inleakage. (Both ABVES and CREVS are operating during the tracer gas test).
Temperature effects in this scenario are assumed to be insignificant given that continued operation of adjacent ventilation systems minimizes the temperature differentials between these areas and the Control Complex. The 75% efficiency assumed for the Auxiliary Building charcoal filters is consistent with the proposed license amendment for the filter testing.
i As with the MiIA with LOOP, analysis of this scenario assumes that inleakage is distributed evenly throughout the CCHE volume, even though the location of the penetrations and the design of CREVS ensures that very little inleakage is introduced directly into the control room without being subject to filtration. Given these and other 21 L______-___
l I
1 Control Room thbitsbility Report Florida Power Corpor tion j Crystal River 3 conservative considerations in the analysis, the treatment of MilA without LOOP described above is considered to be an acceptably conservative treatment of this scenario.
VI.3 IIydrogen Purge I
(
i The impact on the control room operators of purging the containment, for the purpose of combustible gas control, has also been evaluated. A conservative hydrogen generation i analysis shows that the containment would not require purging for two weeks after the occurrence of an accident. After two weeks, the containment would be purged to maintain combustible gas mixtures within acceptable limits. The emergency procedures identify that the preferred approach is to periodically purge during favorable meteorological conditions. Favorable meteorological conditions, based on public exposure, are defined j to be when the wind is blowing into an offshore direction. Fortunately, such wind directions are also away from the control complex.
l Although purging could most likely be coordinated with offshore winds to minimize dose, the calculation of the potential control room operator dose due to purging was l performed using the 30 day control room x/Q.11ydrogen purging is achieved by pressurizing the containment with an air compressor to a small positive pressure and allowing the containment to vent at a rate sufficient to maintain the hydrogen
)
concentration below the flammability limit. The release path for the purging is through the hydrogen purge filter system, which was conservatively assigned a removal efficiency of 90% for all forms ofiodines. This system is tested in accordance with the CR-3 Ventilation Filter Test Program to 99% iodine removal efficiency. The requirements fbr ,
the hydrogen purge filters were in the CR-3 Standard Technical Specifications and were !
moved to control under plant administrative procedures. These are identified as former STS requirements that require a 50.59 evaluation for any changes. ]
I i
The results of the hydrogen purge analysis, using TID-14844 source terms, are doses of l 2.9 REM to the thyroid,0.3 REM to the whole body and 11.2 REM to the skin. When i added to the dose from the MilA, the total doses remain within the limits of GDC19.
CR-3 will not include this additional purge dose in the calculation of breach margin. This is com.aered acceptable as corrective actions, such as the closure of all breaches, would be taken within the minimum two week period available before purging is required.
22
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' l Control Room Habitability Report '
l Florida Pow; Corporation Crystal Rivt+ j VI.4 Summary of MHA Results Using the dose analysis methodology described herein, along with RG 1.4, SRP 6.4, SRP i 15.6.5, and the input parameters identified in Tablel, the CR-3 30-day integmted MHA l Control Room operator doses were calculated to be: 4 Control Room 30-day Integrated Doses for the MHA l
L Scenario Thyroid (REM) Whole Body Skin (REM)
(REM) .
MHA with LOOP 18.6 0,4 15 MHA without LOOP 19.6 0.7 18 As can be seen above, the integrity of the CCHE and the configuration and operation of the CREVS is adequate to ensure that the consequences of a design basis MHA is in i compliance with the dose limits of GDC-19 of 10 CFR 50, Appendix A. The existence -
other mitigation features, (i.e., vestibules at CCHE boundary doors, maintained water seals), source term reduction by the Auxiliary Building filters, and Radiological l Emergency Response Plan dose management procedures provide additional assurance of improved control room personnel protection. The inputs, assumptions, and methodology
[ utilized in the MHA analysis are conservative both individually and as a whole. The actual consequences of a design basis MHA are expected to be well below that determined in the analysis.
l l The results of these analyses show that the bounding version of the MHA is that which is associated with the accident occurring without a LOOP, however, the results are very l'
- similar. The MHA accident analysis used TID-14844 source terms and assumed that
)
CREVS boundary dampers were isolated from the outset of the event by virtue of the 4 psi Reactor Building High Pressure ES signal. The MHA, as well as the other accidents l j
analyzed and discussed below, do not rely on the automatic isolation from the control room ventilation radiation monitor.
The analyses also consider the concept of" breach margin." The breach margin is the area l that can be opened in the CCHE envelope and still result in 30-day post accident MHA doses within the limits. The breaches are controlled and logged to ensure that the integrity of the CCHE is maintained within the limits through normal operation. The '
2 MHA without LOOP analysis calculates an allowable breach area of 35.5 in in addition l
to that area determined by the tracer gas leak test. Considering that CCHE breach areas are maintained below the value of 35.5 in2, it is concluded that the level of CCHE l integrity is sufficient to meet operability requirements pertaining to radiological j consequences of the MHA. !
)- 30 4
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Control Rcom Habitability Report Florida Power Corporation Crystal River 3 VIL Other Design Basis Accidents Other design bcsis accidents were reviewed to ensure that the MHA accident is the most limiting / bounding accident regarding Habitability of the Control Room. The accidents analyzed include the Letdown Gie Dreak Accident, Steam Generator Tube Rupture, and Fuel Handling Accident. CR-3 is not an SRP plant. Therefore, the licensing basis SGTR for CR-3 does not include some cf the conservative assumptions of the S~RP, such as a LOOP, consideration of the worst single failure, and iodine spiking. However, for the purposes of this report and to confirm that the MHA would still be the bounding accident, a SGTR analysis consistent with SRP assumptions was performed.
This effort included a single failure analysis, detailed (i.e., RELAPS) thermal hydraulic amilysis, as well as validation of the assumed operator response times through simulator runs. A LOOP was taken at the time of reactor trip, thus eliminating the condenser as a significant iodine removal mechanism. The initial coolant activity was assumed to be at
' the pre-accident spie limit of 60 pCi/gm DE l-131. The inputs, assumptions, and methodologies used in the Control Room Habitability SGTR evaluation described herein were only perfonned to demonstrate that the MHA remains bounding. CR-3 still considers the SGTR analysis presented in Chapter 14 of the FSAR as the licensing basis SGTR.
VH.1 Letdown Line Accident (LLA) l The Letdown Line Accident is discussed in Chapter 14 of the FSAR. The accident consists of an outside containment line break releasing primary coolant into the Auxiliary.
Building during full power operation. The resulting dose received by control room operators has been evaluated for the accident occurring with a LOOP and without a j LOOP. In addition, two source terms were considered for each case: a pre-accident iodine spike source and an accident initiated iodine spike source.
At time zero, the letdown line outside containment is postulated to break releasing primary coolant in the Auxiliary Building during full power operation. The control room is operating in full makeup (5700 cfm) and recirculation (37800 cfm) without filtration.
For the first six minutes, control room personnel try to keep the reactor at full power.
l During this time period, the ABVES filter is removing 75% of the iodine from the air as
[ it is exhausted to the environment. The control room is still assume'l to be in nomial t makeup, with 5700 cfm of outside air. No credit is taken for the automatic radiation monitor isolation.
At six minutes, the reactor trips and simultaneously, for one case, there is an assumed J LOOP. As a result of the LOOP, control room ventilation trips resulting in no makeup or recirculation of the air within the control room. Because there is a LOOP, the Auxiliary 31
Control Room IIabitability Report Florida Power Corporation Ceystal River 3 Buiding Ventilation System loses power; therefore, credit for the ABVES filter stops. At l this time, unfiltered inleakage into the control room begins. For the LLA with LOOP, the l unfiltared inleakage is 207 cfm. This inleakage, which includes the 10 cfm inleakage due l
to opeping and closing of doors to the control room, is based on the inleakage value that corres; onds to the worst atmospheric dispersion factor (i.e., low wind speed).
l The a;cidem continues at these conditions until 19.5 minutes. At this time, control room l r.asonnel recognize and isolate the letdown line break ending the release. The unfiltered l inleakage into the control room is reduced to 90 cfm to minimize outleakage. This was don'e to increase the residence time of activity in the control room resulting in conservative doses.
It was conservatively assumed that the control room recirculation filters are not started for 30 minutes after LOOP; therefore, at 36 minutes into the accident, the control room recirculation is brought back on line with the filters. The filters have a 95% iodine removal efficiency. From this time, all conditions remain the same until the end of the accident scenario (30 days).
i
( For the LLA without LOOP, there are two significant changes from the LLA with LOOP.
l Since there is no LOOP, the ABV Filter continues to filter the air released into the
[ environment from the Auxiliary Building for the duration of the accident. Second, the
[. control room makeup is operating at 5700 cfm until the isolation of the letdown line at 19.5 minutes into the accident. Af er 19.5 minutes, the unfiltered inleakage is again reduced to 90 cfm to generate conservative control room doses.
The release rate calculations evaluated both the accident initiated spike and the pre-l accident spike cases. The total activity released was much greater for the pre-accident spike case, so that case was used for the dose calculations as the bounding letdown line
[ rupture case. The results are presented in the follo' wing table:
L Control Room 30-day Integrated Doses for the Letdown Line Accident Using the Pre-Accident lodine Spike i
Thyroid (REM) Whole Body (REM) Skin (REMJ l
l LLA With LOOP 6.61 0.0224 1.62 LLA Without LOOP 12.4 0.0558 4.04 t
32 l
- Control Room IIabitability Report
' Florido Power Corporation Crystal River 3
\
1 VII.2 Steam GeneratorTube Rupture (SGTR) Analysis The current licensing basis accident analysis for the SGTR accident was not based on the methodology described in the SRP. Specifically, the current analysis was based on the followm' g: I e does not consider a LOOP, e does not consider a single failure, e' considered a constant 435 gpm leak for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, with I gpm leak in the unaffected OTSG, I
I e considered 1% defective fuel which correlates to a specific activity of i approximately 7 Ci/gm dose equivalent 1"' and did not consider iodine spiking, 1
4 considered an iodine decontamination factor of 10 for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the condenser. J l
Using the current licensing basis SGTR accident analysis methodology, the 30-day thyroid dose was calculated at 1.6 REM. l l
As a part of the restart review with the NRC, FPC agreed to analyze the habitability of the Control Room using an SRP approach for the SGTR accident. The primary elements of the SRP approach that differ from the licensing basis is that the analysis considers a L LOOP, considers the most-limiting single failure, and considers both a pre-existing iodine spike at the maximum value allowed in the Technical Specification (for CR-3,60 pCi/gm dose equivalent I"') and an iodine spike 500 times the equilibrium release rate coincident p with the operation value of I Cilgm dose equivalent I*.
l In response to the NRC's request, a detailed SRP-like analysis was performed. The analysis is referred to ns an SRP-like analysis since credit is taken foi some non-safety-related components, specifically, the analysis credits the use of the ADVs for couldown, l and the use of the PORV for depressurization.
l' l: . A conservative and detailed analysis /model was prepared that included a RELAPS L thermal hydraulic analysis, a single failure analysis, and included simulator validation to confirm the adequacy of the operator response time assumptions. ' The operator response times and sequence of events for the SRP like SGTR analysis are presented in Attachment B. The analysis conservatively considered a LOOP coincident with the j' reactor trip. The results of the single failure analysis, which included both qualitative and L quantitative sensitivity studies, revealed that the most limiting single failure is the failure
! of the Atmospheric Dump Valve (ADV) on the unaffected steam generator to automatically open in response to the post trip pressure control signal. This forces the operator to cool down using the affected generator's ADV until the unaffected steam generator's ADV can be opened. This results in an extended cooldown time and hence time until the affected generator can be isolated.
Some of the conservatism used in the analysis are as follows:
33
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Control Room Habitability Report Florida Power Corporation Crystal River 3 I L
- treated al! leakage as a ground level release, neglected any buoyancy affects of the steam, ,
l
- minimized the steaming capacity since less steaming is bounding, l
e maximized the high pressure injection (HPI) since more HPI is bounding, e utilized a RCS Low Pressure Trip Setpoint based on nominal setting plus uncertainty to obtain the earliest LOOP and loss of condenser, e assumed 20% steam generator tube plugging to minimize cooldown rate, e assumed an HPI temperature of 100*F, e did not credit purification by the MU&P system,
[
- i l did not credit decontamination ofiodines after the reactor trip in the OTSG (i.e.,
used a partition factor of 1.0),
e used a decontamination factor of only 100 in the condenser prior to reactor trip,
- and l . used a constant release rate from the intact OTSG at 150 gpd.
l l In addition to the conservative system parameters and model, the analysis used conservative operator response times. The detailed sequence of events that is described in Attachment B was compared against simulator testing. Margin was added to the observed simulator times.
The Control Room Model that was used to analyze the SGTR accident is as follows: At the time that the tube in one of the steam generators is postulated to break, it was assumed that the control room is operating in full makeup (5700 cfm) and recirculation l (37800 cfm) without filtration. The control room was modeled in this configuration until i the reactor trips at eight minutes (i.e., no credit for isolation of the intake by the L ventilation radiation monitor). Coincident with the reactor trip, and the simultaneous LOOP, the auxiliary building ventilation stops, control complex dampers fail closed, and control complex intake and recirculation fans stop. Unfiltered inleakage into the control room begins. With the auxiliary building ventilation off, the unfiltered inleakage is 207 cfm. This inleakage, which includes the 10 cfm inleakage due to opening and closing of doors to the control room, is based on the inleakage value that corresponds to the worst L atmospheric dispersion factor (i.e., low wind speed).
L L lt was conservatively assumed that the control room recirculation filters are not started L for 30 minutes after LOOP; therefore, at 38 minutes into the accident, the control room recirculation is brought back on line with the filters. The filters have a 95% iodine removal efficiency. From this time, all conditions remain the same until the end of the I accident scenario (30 days). ' After the filter recirculation system is in operation, it is l conservatively assumed that there is an addit 5nal 125 cfm filtered inleakage into the control room which could be caused by local system imbalances.
34
Control Room IIabitability Report Florids Power Corporation
- Crystal River 3 Similar to the letdown line rupture, the release rate calculations were performed for both the accident initiated iodine spike and the pre-accident spike cases. The pre-accident l
spike case was used for the dose calculation based on the significantly higher amount of activity released. The curies released for the pre-accident spike case are provided in Attachment B. The dose calculation results are presented in the following table:
Control Room 30-day Integrated Doses for the SGTR Accident Using the Pre-Accident Iodine Spike l
Thyroid (REM) Whole Body (REM) Skin (REM)
SRP-like SGTR 9.6 0.006 0.35 l VH.3 Fuel Handling Accident (FHA) Analysis In the FHA, it is assumed that the cladding of all fuel reds in one assembly experiences mechanical damage such that the entire quantity of fission gases trapped within the gap l are released to the fuel storage pool or the refueling canal. Since the radioactive material i released from the damaged fuel assembly must pass through water reaching the Fu21 Huilding or Reactor Building atmosphere, credit is taken for retention ofiodine in the water.
l The radioactive material that escapes from the spent fuel pool or from the refueling canal to the Fuel Building or Reactor Building will be picked up by the building ventilation systems and processed prior to release to the environment. The spent fuel pool r ventilation system provides a continuous sweep of air across the top of the pool and the l cask loading pit. All exhaust flow is directed to the main ABVES or the Reactor l Building Purge System where it passes through roughing, HEPA and charcoal filters before being discharged to the environment.
l Two cases for the FHA were evaluated:
. A conservative case, in which the Building Ventilation System (either auxiliary building or reactor purge) is assumed inoperative. Hence, no credit is taken for tiltration from the Building Ventilation System.
- A realistic case,in which the Building Ventilation System (either auxiliary building or reactor ptuge) is assumed to be operating along with its HEPA and charcoal filters in this case, credit is taken for a 75% removal rate for all species ofiodine.
In the FHA's puff release model, the activity is released from the building very rapidly.
Essentially all of the activity is released in a matter of seconds which simulates a puff release. The effect of scrubbing in the spent fuel pool on iodine activity is determined by 35 L-_--__---_------------_ - .-
Control Room IIabitability Report i Florida Power Corporation Crystal River 3 using the Regulatory Guide 1.25. An effective DF of 100 is assumed. Procedural !
requirements ensure that the CREVS is placed in the recirculation mode prior to and during any irradiated fuel movement. Therefore, for the FII A, the Control Room Ventilation System is initially operating with an unfiltered recirculation flow rate of 37,800 cfm, all bubble tight dampers closed and an inleakage rate of 524 cfm. After a thirty minute delay, the recirculation filter is manually actuated. It is assumed the recirculation filter efficiency for all iodine species is 95% .
The results of the two FHA cases considered are the following:
Control Room Integrated Doses for a Fuel llandling Accident Thyroid (REM) Whole Body (REM) Total Skin (REM)
Fila Analysis with No Filtration of Releases 11.9 0.049 3.48 4 FIIA with Filtration of Releases 2.98 0.049 3.48 VII.4 Summary of Other Analyses The results of the Habitability analyses for the other design basis accidents demonstrate that the integrity of the CCHE and the configuration and operation of the CREVS is adequate to ensure that the consequences of the spectrum of design basis accidents is in I compliance with the dose limits of GDC19 of 10 CFR 50, Appendix A. The results of the analyses also show that the bounding accident with regard to Control Room l Habitability is the MHA with the accident occurring without a LOOP.
VIII. Ilazardous Chemical Evaluation i CCHE integrity is also required to provide protection to the control room operator in the event of a toxic gas accident. Regulatory Guide 1.78 provides infomtation and assumptions for assessing toxic gas accidents with regard to control room habitability.
From this document comes the basic criteria that, in the event of a toxic gas accident, appropriate toxicity limits not be exceeded in the control room two minutes after initial detection in order to allow the operator adequate time to take action (i.e., don an air pack) prior to being overcome. The Regulatory Guide allows for detection to be accomplished by personal means (nasal detection) or with automatic detection equipment. CREVS isolation, if required, can be attained either by operator action or by an automatic signal from toxic gas detectors. At CR-3, procedures provide the appropriate instructions for the l
operator in the event of a toxic gas accident and for the use of air packs.
1 l 36 l
Control Room H:bitibility Report Florida Power Corporation Crystal River 3 1
Based on previous evaluations, the locations and quantities of toxic gas storage sites at the Crystal River site which posed a potential liability to CR-3 control room habitability are listed below:
Toxic Gas container Size and Location Toxic Gas Helper CR-1/CR-2 CR-4/ CR-5 Cooling Towers Chlorine -- Cl, 17 tons none 1 ton cylinders Sulfur Dioxide --SO2 50 tons 45 tons 1 ton cylinders The most limiting source of toxic gas was an SO2 tank at CR-1 which had been administratively limited to storage of 30 tons. Automatic detection and isolation was required as a result of that tank. That tank has recently been replaced with a system that uses solid pellets which are converted to SO2 as needed. The tank has been emptied ofits contents and will no longer be used. As such, it is no longer the hmiting source.
The next most limiting toxic gas source is the Helper Cooling Towers. Currently, there is no SO2 or Cl stored 2 at this location and this is ensured by a CR-1 " Red Tag" clearance No.1997-01543. Prior to releasing this clearance, administrative controls will be in place to limit the Helper Cooling Towers to 8 tons of Cl2and 30 tons of SO2 . 189-0053 Revision 3," Control Room Habitability Helper Cooling Tower Project", is the current calculation of record that analyzes ruptures of these tanks. The calculation analyzes the 17 ton Cl 2and 50 ton SO; tank ruptures and found that automatic isolation was required only to meet the two minute C12 toxicity limits. FPC has used revised calculations to evaluate the lower quantities: 8 tons of Cl2 and 30 tons of SO2 . This analysis combined the wind tunnel results from the CR-1 SO2 tank model and traditional atmospheric dispersion mathematical modeling techniques to conclude that CREVS could remain in .
its normal alignment (i.e., no CCHE isolation required) without exceeding Control Room !
toxicity limits if up to 9 tons of Cl 2or 50 Tons of SO 2were released. Thus, CCHE l inleakage would be of no consequence given the new limits of 8 tons and 30 tons for Cl2 j and SO2 respectively.
i A revised calculation also analyzed the Cl2 and SO2 stored at the CR-4/CR-5 site. This calculation allows for a 4 ton Cl 2release at the CR-4 and CR-5 cooling towers located 3600 feet from the CR-3 control complex intake. There are eight one ton tanks, with four in service on a single header at a time. The assumed accident has one tank fail and the i other three leak out though the common piping. With no automatic detection or isolation, l the calculated control complex concentration is 7.5 ppm, well below the 15 ppm toxicity !
limit for this source. The one ton SO2tanks were not analyzed due to the SO2 at the Helper Cooling Towers being more limiting, both in terms of volume and dispersion.
The amount at CR-4 and CR-5 is less (one ton versus fifty tons), farther away (3600 feet versus 3400 feet), and has a htrger building wake value (3 versus 2L Since the calculation allows > 50 tons at the Helper Cooling Towers without automatic detection 37
Control Room Habitability Report Florida Power Corporation Crystal River 3 and isolation, then the CR-4 and CR-5 SO2 amount will also not require automatic detection and isolation.
Based on the above discussion, it can be seen that the current level of CCHE integrity provides adequate protection for the control room operator for postulated toxic gas events. It is also noted that the updated analyses would support operation without crediting the existing toxic gas detectors.
References
- 3. FPC SA/USQ For Sargent & Lundy Calculation SL-9929-M-009 R1
- 4. NRC letter 3N0589-25 to FPC dated 5/25/1989," Crystal River Unit 3 - Control Room Habitability Evaluation (NUREG-0737 Item III.D.3.4) ( TAC No. 64805)"
- 5. License Amendment Request (LAR) #222 R0, " Control Room Emergency Ventilation System and Control Room Habitability"
- 7. FPC IOC Ser NL97-0424 dated 12/20/97, Fm D. Kunsemiller to M. Rencheck
- 8. FPC Calculation No. M97-0110R/4 (SL-9929-M-0009 R/4), Control Room Dose Analysis and Maximum Infiltration Following a LOCA.
- 9. FPC Calculation No M97-0137 R/4 (FPC-CED-M-01 IU4), Control Room Habitability Analysis Considering LOCA without LOOP.
- 10. FPC Calculation No. M97-0111 R/4 (SL-9929-M-0010 R/4), Control Room Dose Analysis for a SGTR & FHA 4
comments on validation ofinputs to SL-9929-M-0009 and 125 cfm of CREVS n Induced Filtered Inleakage
Clary FPC, Re: Third party assessment of thermal stack effect presented in SL-9929-M-0009.
- 13. FPC Calculation No. M97-0109 R/l (SL-9929-M-0008 R/l), Toxic Gas Analysis
Line Break" l 15. NRC letter Ser 3N1297-19, dated 12/24/97, Fm L. Raghavan (UNNRC) to R. l l Anderson (FPC), Re: Interim Assessment Results of Justification for Continued Operation
Control Desk
- Control Desk 38 l
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I Control Room Habitchility Report l Florida Power Corporation Crystal River 3
- 18. EM-103 Revision 11, dated 01/21/98," Emergency Plan Implementing Procedure, Operation and Staffing of the CR-3 Control Room During Emergency Classifications" l 19. FPC Calculation M98-0018 R/0 (SL-9929-M-0021 R/0)- Control Room Dose
! Analysis for a Letdown Line Break Analysis.
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Control Rcom 111bitsbility Report Florida Power Corporation l Crystal River 3 l
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, i Control Room Hrbitsbility Report Florid 2 Power Corporation i Crystal River 3 l Attachment B Steam Generator Tube Rupture j 1. Tim'e Line i
i T=0 Tube Rupture T = 8 min. Rx trip, LOOP, loss of condenser, isolation of CCHE with LOOP T = 38 min. CR ventilation on recire T = 53 min. Operators initiate cooldown with ADV's - fails on unaffected generator l
T = 55 min. Initiate RCS depressurization with PORV's
- T = 78 min. Manually open failed ADV 1
T = 108 min. Isolate affected OTSG
! T = 8 hrs ' Terminate release from unaffected generator I
I l~ ,
43
Co trol Ro:m Hibitxhility Report Florids Power Corpor tion Crystal River 3 l Attachment B (continued) i
- 2. Curies released for pre-accident spike case of the SGTR CURIES l Isotope 0 to 480 sec 480-2280 sec 2280-28800 sec l 1131 6.485E0 5.130E2 2.024E3 1132 2.493E0 1.972E2 7.780E2 I133 7.784E0 6.155E2 2.429E3 t.
l 1134 8.629E-1 6.825El 2.693E2 1135 3.164E0 2.503E2 9.875E2 l Xel31m 2.974E0 9.678E0 1.753E1 Xe133m 5.081E0 1.654El 2.996El Xel33 4.787E2 1.558E3 2.823E3 l
Xel35m 5.179E-1 1.685E0 3.054E0 l
Xel35 1.182E1 3.848El 6.972El Xel38 8.370E-1 2.724E0 4.936E0 Kr83m 5.505E-1 1.792E0 3.246E0 l Kr85m 2.726E0 8.871E0 1.607E1 Kr85 2.130El 6.933El 1.256E2 l Kr87 1.407E0 4.579E0 8.297E0 ,
Kr88 4.365E0 1.420El 2.573El
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l 44
Control Room IIrbitsbility Rcport Florid: Power Corporation !
Crystal River 3 Attachment C Periodic CCHE Integrity Test Method i
l i
The proposed license amendment request requires that an integrity test of the control l
complex habitability envelope be performed at least once each operating cycle. This Attachment provides the basic method by which the required integrity test will be performd. I i
l The integrity of the CC11E will be determined using tracer gas tests. The tracer gas test procedure will be based on ASTM Standard E741-93, " Standard Test Method for Determining Air Change Rate in a Single Zone by Means of a Tracer Gas Dilution." Tae tests will use an electronegative gas, sulfur hexafluoride (SF6), as a tracer. i Tracer gas testing under post-accic'ent conditions requires the Control Complex to be in its emergency recirculation mode and treating the entire CCIIE as a single volume. An additional penalty is not required at boundary damper locations since the dampers would be subject to the same pressures during testing as would be expected during post-accident operation. The additional penalty is inconsequential for CR-3 since two bubble-tight dampers have been installed in series at each boundary isolation location. The installed dampers were factory tested to be bubble tight. Post installation testing demonstrated the dampers to have inconsequential leakage. Testing was performed by pressurizing between each pair of series dampers; therefore, one damper wa9 pressurized as it would be in service, in the closing direction while the other was pressurized in the opening direction. Individual damper testing will not normally be performed as part of the periodic integrity test, as the dampers are included as part of the CCllE boundary during the tracer gas test.
The following test conditions will be prescribed for the tracer gas testing:
. The CREVS will be placed in emergency recirculation mode. Both " Toxic Gas" and the "lligh Radiation" recirculation lineups will be tested.
l . All fans in the Turbine Building Ventilation System (TBVS) will be secured. The turbine building normally remains well vented to atmosphere through normally open doors, roll out windows, and roof vents.
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- l Control Room Habitability Report Florida Power Corporation Crystal River 3
. All fans in the Intermediate Building Ventilation System will be secured.
Conditions in the intermediate Building are not critical to the test in that relatively few penetrations are on the CCHE / Intermediate Building wall.
e The Auxiliary Building Ventilation will be operated to maintain an auxiliary building pressure differential of at least 0.125 inches water gauge vs. the turbine building pressure. This value is large enough to minimize test inaccuracies and external effects. The pressure will be measured and sustained for the duration of L the test.
i e The test will be conducted when personnel traffic is minimized. Since a 10 cfm allowance for access / egress is added to measured inleakage, minimizing traffic precludes counting this effect twice. )
e Testing will be conducted with vestibule doors blocked open. This l
conservatively assumes no credit for the additional integrity provided by vestibules.
. All loop seals penetrating the CCHE will be verified to be filled prior to testing.
Controls are in place to ensure that these loop seals are maintained full during plant operation.
l The test will use the tracer concentration decay method of ASTM Standard E741-93. Due l to the large size of the CR-3, approximately 40 sample locations are used. Statistical data l analysis is applied to the measured results.
Based on the test measurements, the inleakage rate can be determined for the pressure differential established during the test. The inleakage rate at other pressure differentials will be determined mathematically. The calculated inleakage rates will then be compared against acceptance criteria. The acceptance cri: e 1 be based on the inleakage rates assumed in the dose calculations for the various combinations of ventilation system operation and meteorology.
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Control Room IIabitability Report Florids Power Corporation Crystal River 3 l
! Attachment D l
Control Room Habitability Report Verification Review Report Prepared By:
Raymond A. Crandall-- Safety Allalysis
- 8. buld 7/13/9e
- Report Reviewed By
- 7[81fff Kevin R. Dampbell- Operations
% U 9E Kelme~tli L. Anderson - System Engineer AttC Sidney C. Powell- Licensing Mn ' '
4-7/w/W Robert W. Knoll - Safety Analysis
! Report Approved By: h 7!@fjS erome A. Ranalli - Manauer. Nuclekt S fety Management I
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