Text

Rev 5 to Justification for Continued Operation for CR Emergency Ventilation Sys & Control Complex Habitability Envelope
	ML20198H507
Person / Time
Site:	Crystal River
Issue date:	12/31/1997
From:	; FLORIDA POWER CORP.
To:	;
Shared Package
ML20198H500	List: 3F0198-20, Addresses Issues Re Emergency Ventilation Sys & Control Complex Habitability Envelope Raised by NRC in; ML20198H507;
References
	NUDOCS 9801130322
	Download: ML20198H507 (43)
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REVISION S j

l JUSTIFICATION FOR CONTINUED OPERATION FOR THE CONTROL ROOM EMERGENCY VENTILATION SYSTEM AND THE CONTROL COMPLEX HABITABILITY ENVELOPE (For MODES 3,4,5, and 6 ONLY)

Inital Issue: 11/26/97 Change la 11/28/97 Change 2:

11/29/97 Change 3:

12/06/97 Change 4:

12/28/97 Change 5:

12/31/97 9801130322 980107 yDR ADOCK 05000302 I

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.l-l DEFICIENCY REPORT INSTRUCTIONS-Precursor Numbert PC 97-435$

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4 DEFICIENCY REPORT Re: PC 97-4355 CRANGE RECORD CHANGE:

N/A REASON: Original DATE:

11/26/97 Issue ORIGINA1DR VERIFIER:

SUPERVISOR:

Robert.J. Lane Jack Wilkinson

Harry B. Oates CHANGES Page Description N/A

see original cover sheet (attached) for signatures.

t 1

i Change:

1 REASON:

Comments DATE:

11/28/97 from 11/26/97 PRC Meeting ORIGINATOR:

VERIFIER:

SUPERVISOR:

3

~*

M. Clary Steven D. McMahan Harry P. Oates CHANGES Page Description 4

added... "(corresponding to a filter dP of 4" wg) "

19 item 2) added "... In addition to satisfying current I"!!S testing requirements..."

item 9) changed NGRC to PRC 21 & 22 added information regarding spray flow rates:

Calculation I86-0002 Rev.

5, 1/16/96, determined containment spray removal constants using the new instrument error corrected flow values of 1397 gpm (injection phase) and 1112 gpm (recirculation phase).

Spray constants associated with the lower value of 1112 gpm are used in revised dose calculations.

The instrument loop uncertainties for spray flow indication and control were being reviewed concurrent with performing the revised dose calculations. As a contingency, the revised dose calculation looked at a containment spray flow rate of 1000 gpm and found that it was essentially the same as the 1112 gpm case. The calculation concludes that containment spray rate of 1000 gpm can be tolerated.

Balliet to Widell ltr ser NOE97-2311 dtd 11/11/97,- shows~that when spray is being supplied from the RB Sump, the actual flow may be 121 gpm below the indicated flow of 1200 gpm. Thus, the lowest value may be 1079gpm.

2

CHANGE:

2 REASON: Incorporate DATE:-11/29/97 Licensing Comments ORIGINATOR:

VERIFIER:

SUPERVISOR:

Steven D.

McMahan M. Clary K. Anderson for Harry B. Oates CHANGES Page Description 3

added..." 3)

CR-3 Operating License The CR

3 Operating License contains a

requirement to maintain Control Room habitability as specified in the post-TMI requirements of NUREG-0737.

However, there-is no requirement for the measurement or evaluation of inleakage in accordance with specific requirements.

4)

FSAR discussion..."

6 Added "...See page 6A, that follows..."

6a Inserted page that read...

If a radiological accident were to occur which involved the release of radioactive material from the reactor or spent fuel storage area, the CR-3 Radiological Emergency Response Plan would be implemented. The plan provides for staffing the emergency response organization and establishing emergency response actions commensurate with the severity of the event.

Actions required in the Emergency Plan Implementing Procedures include dispatching a Health Physics Technician to the Control Complex to monitor radiological conditions, and to provide radiological and meteorological data to the Dose Assessment Conrdinator. The Health Physics Technician will perform radiological surveys within the Control Complex, including surveys for airborne radioactivity. Dose Assessment personnel use data collected from surveys to project expected personnel doses. Provisions exist in the Emergency Plan Implementing Procedures for considering administration of potassium iodide (KI) to personnel based on projected dose. A projected dose of 25 REM to an individual has been established as the threshold for considering administration of KI.

Control room dose calculations contain very conservative assumptions 3

regarding operator presence in the control room.

For example, in accordance with the Murphy-Campe methodology it is assumed that the operator is present in the control room continuously for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then 60% of the time for the next 3 days, and then 40% of his time for the remaining 26 days. Similarly, atmospheric conditions are assumed which channel.the release of radioactivity toward the control _ complex at conditions which maximize the plume concentration, particularly in the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the scenario. Based on the conservatisms that exists throughout control room habitability dose calculations, no doses approaching-GDC 19 limits are anticipated in a _ realistic accident.

However, provisions exist in currently approved procedures to monitor actual and projected doses based on measured 'and observed conditions, and to control personnel exposure. Through the protective features of the control complex evaluated in the control room habitability calculations, and the established emergency response procedures, protection' of the control room operators is assured.

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CHANGE: 3 REASON: Incorporate DATE:

NGRC Comments i

SUPERVISOR:

VERIFIER; I W. 2.ba %A 4 ORIGINATOR:

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Page Description 1

Changed " multitude" to " number" 6a Moved contents of page 6A into a new section in under Just'fication for Continued Operation.

Now located under Additional Protective Features section 3. Dose Management.

Deleted page 6A 19 Added a discussion regarding vestibules under new heading Additional Protective Features section

1. Vestibules. New section reads as follows:

1. Vestibules CCHE boundary doors represent a significant source ofleakage into the CC. There are three double doors and three single doors. Vestibules were added to the three CCHE double doors in 1996, and have proven to be effective in reducing differential pressure across the existing doors.

Reducing the differential pressure exerted on the boundary doors has two benefits. First, lower differential pressure reduces leakage around the doors, and second, lower differential pressure allows the door closers to perform more reliably. During this outage, vestibules were added to the single CCHE boundary doors. All of the vestibules have additionally been sealed at interfaces with the CCHE boundary making the enclosures more effective in reducing CCHE boundary door leakage.

Control Complex tracer gas' testing was performed with the vestibule doors blocked open to assure that the test was conservative. Blocking open the vestibule doors increased conservatism in two ways. First, normal access and egress was permitted during the tracer gas leakage testing which contributed to the measured leakage. Dose calculations 5

performed in accordance with standard methods include a factor of 10 cfm of continuous leakage to account for access and egress during an accident. This factor was added to the measured inleakage in performing CR-3 dose cakulations. Therefore, the effect ofleakage during access and egress was applied twice. In a real event the vestibules would not be blocked open. Since they function similar to an airlock, they would be effective in reducing leakage during access and egress.

The second manner in which maintaining the vestibules open during the test was conservative is that the existing boundary door was the only barrier to leakage during the test. In an actual event the vestibule doors would be in their normal closed positions, and would be effective in reducing infiltration into the CCHE through the doors. This feature would be particularly effective in reducing operator dose during the MHA without LOOP scenario where leakage is induced from the Turbine Building into the CCHE due to operation of the ABVS.

19 Added a discussion regarding Auxilary Building filtration under new heading Additional Protective Features section 2. Auxilary Building-Filtration. New section reads as follows:

2. Auxiliary Buildinn Filtration Performing dose calculations in accordance with the requirements of NUREG-0737 Item III.D.3.4, Control Room Habitability, includes assuming source terms spccified in Standard Review Plan 15.6.5, One aspect of this for a plant that does not have an "ESF atmosphere filtration system," is leakage of 1500 gallons of water contained in engineered safeguards piping outside of containment must be assumed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident.

This amounts to leakage into the Auxiliary Building (AB) of highly contaminated water from the reactor building sump. An ESF atmosphere fiiiration system is like the CR-3 ABVS with HEPA and carbon filters, however it is required that an ESF system be powered from an onsite power source. Since the ABVS is not powered from an onsite power source, CR-3 control room dose calculations include the required leakage term.

This term is responsible for approximately 8 REM of the projected individual control room operator dose of 26.5 REM during the MHA with LOOP scenario.

The CR-3 ABVS has redundant fans and filters which are operated continuously during normal operation. Therefore, the system must be maintained in good ' operating condition.

The filters are tested and maintained in accordance with approved procedures that -implement regulatory guidance on emergency filter systems. As such, filter efficiency is routinely verified by carboa filter media testing. Evaluations of the FPC power grid performed for station blackout concerns have demonstrated a 6

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. i high degree of reliability. Studies have shown that power should be 3

restored to the_ site in under eight hours following a loss of offsite power due to dirtuitances on _the grid. Therefore,- there is' a high degree of j

assurance that,the ABVS _will be available and capable of--performing j

event; l

effectively; to' reduce radioactivity released from the 'AB following' an Calwlations 'show a near linear pelationship between AB source t'erm and operator dose. Therefore, using a conservative ABVS filtration efficiency of 75%1would result in a dose reduction of approximately 6 REM (8 REM contribution X 0.75 = 6-REM reduction) from the 26.5. REM projected dose for the MHA with LOOP scenario. This reduction in dose is applicable only to the MHA with LOOP analysis.175% efficiency for ABVS cebon filters has been previously accepted for CR-3 as an interim measure curing resolution of reactor building flood level issues, and is already credited in the analysis for the MHA without LOOP.

n 9

Added (NOTE: The isolation dampers were tested in-place and the test boundary exhibited insignificant leakage.)..."

13 Added "...in addition to that-measured by the.

tracer gas--leak-test..."

13 Reworded sentence to read "...Since other DBAs might not actuate this signal, a review of events has been performed which rely on automatic radiation detection or operator action for isolation..."'

. 21 Summary / Conclusions-item 9) required DBA control Room Operator dose analysis to be,-

reviewed and--verified prior to NGRC-approval of.

JCO..'This action was. complete so it was deleted from list.

21 Summary / Conclusions item 7) required administrative controls to fill loop seals-prior--

to Mode 4. This was complete so it was deleted from'the-list.

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4 REASON:

DATE:

Incorporate changes I

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M97-0109 Rev 1 M97-Ollo Rev $1 W

.M97-Olll Rev 13 p M97-0137 Rev 2 and miscellaneous editorial comments VERIFIER:

SUPERVISOR:

ORIGINATOR:(}f ffa[

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b, M/M U (2- %-n CHANGES Section Description All Added section numbers 3.1 Inserted sentence "...The source term applied in the MHA dose calculations is derived in accordance with the U.S. Atcmic Energy Commission Technical Information Document (TID) 14844..."

Deleted: "This 26.5 REM thyroid limit is currently taken as the acceptance limit at CR #3" To be consistert with Ref. 7 3.2 Changed; "The Fuel Handling Accident.is identified as limiting in MODES 5 &

6."

To; "The Fuel' Handling Accident is identified as the limiting radiological event in MODES 5 & 6."

3.3 Changed

"However, there is no requirement for the measurement or evaluation of inleakage in accordance with specific requirements."

To: "However, thers is no required specific method for the measurement or evaluation of inleakage."

3.4 Third bullet; corrected typo Changed "... 0.6 air changes...'"

To 0.06_ volume changes..."

Reworded last paragraph To be consistent with Ref. 7 4.0 First paragraph; Changed "... 26.5 REM limit..."

To 26.5 REM exposure..."

To be consistent with Ref. 7 8

c.:

5.0 First paragraph; Changed "..

26.5 REM'11mit..."

To

... 26.5 REM exposure..."

Fourth paragraph; Changed "... obtain the 26.5 REM limit.. "

To

"... calculate the 26.5 REM exposure..."

To be consistent with Ref. 7 6.0 Second paragraph; Inserted reference to TID

...-Reg Guide 1.<4 { TID-14844}

To be consistent'with Ref. 7 7.2.1 Fist paragraph:

Changed "...Since FPC is committed to SRP 6.4 only as a " guidance" document,..."

To: "...Since FPC follows SRP 6.4 only as a

" guidance" document..."

7.2.2 Second paragraph; Changed: ".. 462 cfm in emergency recircualtion..."

To: ".. 462 cfm in 'High Radiation' recirculation.. "

7.2.3 Revised section to be consistent with References 8 and 12.

7.2.3.1.2 Revised section to be consistent with Refs 8 & 11 7.2.4 Second paragraph:

Changed "...model considers dilution into the large Turbine Building..."

To: "...model accounts for the time dependent rate of change of activity in the Turbine Building..."

Last paragraph:

Changed:

...even though test data substantiates that very little inleakage is introduced into the Control Room from the floors below it without..."

To: "...sven though the location of the~

penetrations and the design of CREVS ensures that very little inleakage is introduced into the Control Room without..."

7.2.6 Revised to be consistenet with References 7, 8,

9, and 10.

7.2.7 Changed "SO2 concentration in the control complex of 17.9_ ppm" to "507 concentration in the control complex of 24.9 ppm" Changed:

"The allowable 4 ton leak bcunds this condition so no automatic-detection or isolation is needed at-CR3 for this source."

To Read:

"With no automatic detection or isolation, the calculated' control complex concentration is 7.5 ppm, well below the 15 ppm toxicity limit for 9

1 this source."

All changes per Rev 1 to SL-9929-M-0008 (Ref 1}l_;

i 7.3.1 Second Paragraph; Changed "...First, normal access..."

To "...First, CCHE access... "

7.3.2 Revised to be consistent with Ref. 8 and Edded a reference to NUREG-1032.-

7.3.3 Added additional information.

8.0-First paragraph; corrected typo changed "...to support-this elevation prior are listed..."

To "...to support this elevation are listed..."

Updated Action Item List to reflect above changes and items completed to date.

9.0 Updated References I

10

~

s CHANGE:

S REASON:

DATE:

Change JCO 12/31/97 Applicability to Modes 3, 4,

5,

&6 ONLY ORIGINATO}:

VERIFIER:

SUPERVIS We C Y y f M' f44]t 7,n M. Clary V. Esquitlo rMMf 5 hML"S SRO: & Sbt/

PRC Chairman:

/

I. Wilson G. Halnon Ihk

/U L CHANGES

_ Section Description 1.0 Expanded section to refer to reference 18 and Appendix A.

9.0 Updated References App. A Added Appendix A to address findings of Ref. 19 t

11

l REVISION 5 l

TABLE OF CONTENTS i

JUSTIFICATION FOR CONTINUED OPERATION FOR THE

-l CONTROL ROOM EMERGENCY VENTILATION SYSTEM AND THE CONTROL COMPLEX HABITABILITY ENVELOPE (For MODES 3, 8, 5, and 6 ONLY) 1.0 Description and Purpose 2.0 Safety Classification 3.0 Licensing Bssis 3.1 Licensing Background 3.2 ITS Requirements 3.3 CR-3 Operating License 3.4 FSAR Discussion 4.0 Impact Analysis and Reliability Concerns 5.0 Description ofIdentified Concerns 6.0 Operability Evaluation 7.0 Justification for Continued Operation 7.1 CREVS Operability 7.2 CCIIE Integrity 1

7.2.2 Test Conditions 7.2.3 Analysis of MilA w/ LOOP 7.2.3.1 Leakage Created by Secondary Effects 7.2.3.1.1 Stack Effect

a. Winter Conditions

b. Summer Conditions 7.2.3.1.2 Leakage Induced by CREVS Ops 7.2.4 Analysis of MHA w/o LOOP 7.2.5 Results of MIIA Analysis 7.2.6 Analysis of Other DBAs 7.2.6.1 Fuel Handling Accident 7.2.6.2 Steam Generator Tube Rupture 7.2.7 Analysis of Toxic Gas Events 7.3 Additional Protective Features l

7.3.1 Vestibules 7.3.2 Auxiliary Building Filtration

- 7.3.3 Dose Management 8.0 Summary and Conclusions 9.0 References Appendices:

A. Isotopes Currently in CR-3 Core I

2

l REVISION $

l JUSTIFICATION FOR CONTINUED OPERATION FOR THE CONTROL ROOM EMERGENCY VENTILATION SYSTEM AND THE CONTROL COMPLEX HABITABILITY ENVELOPE (For MODES 3,4, 5, & 6 ONLY)

1.0 DESCRIPTION

AND PURPOSE System Readiness Reviews conducted in 1997 during the CR-3 Design Basis Outage identified several issua which potentially impacted contrcl room habitability. A number of actions have been undertaken to address these concerns and sipificantly improve the level of protection provi&d for the control room operator.

These include (1) modifications to reduce CCHE inleakage by improving the integrity of boundary elements

- (Ref.1), (2) CREVS design changes to provide alternate means of mechanical equipment room ventilation and to improve system reliability (Ref.1), and (3) programmatic changes to ensure that the assigned efliciency of the Control Complex charcoal filters is consistent with regulatory guidance (Ref 2).

The modifications and design changes discussed above 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 found in regulatory guidance and utilized in prwious calculations. The determination of CCHE inleakage and the application ofinleakage in dose cdculation differs significantly from previous methodology. These differences have been determined to constitute a USQ as stated in an informational report to the NRC on the subject of Control Room Habitability dated 11/10/97 (ref. Docket Letter 3Fil97-09). This report stated that a License Amendment Request would be forthcoming from FPC to address this issue. However, as the time required for review and approval of the LAR is not expected to support the unit restart schedule, a Justification for Continued Operation is being pn: pared 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, Rev.1. The specific issues addressed in this JCO are (1) the operability of the Control Room Emergency Ventilation System, and (2) the integrity of the Control Complex Habitability Envelope.

In response to a verbal Request for Additional Information, Revision 3 of this JCO, dated December 6,1997 was transmitted to the U.S. Nuclear Regulatory Commission (USNRC) by reference 20. The USNRC's initial assessment of the JCO were transmitted to FPC by reference 18. Finding number 5 of reference 18 statesin part:

"...the staff has concluded that you have not adequately demonstrated that the radiation doses to control room operators will be less than the criteria of GDC-19 during a DBA. Since CR-3 has been shutdown for an extended period, the radioactive inventory has decsyed such that CCHE doses would not likely exceed GDC-19 criteria for any design basis accident that would occur prior to initial criticality following restart. However, operation at mode 2 or higher does not appear to bejustified on the basis ofinformation provided to date..."

W 1

Appendix A, tc f this JCO has been prepared to demonstrate that the radioactive inventory has Revsion 5 c decayed such tle m% doses would not likely exceed GDC-19 criteria for any design ba.is accident that 3

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vl REVISION 5 -

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would occur prior to initial criticalsty following restart. Thisihanges the intent of the JCO to only apply to MODES 3,4,5, and 6,1 j

2.0 SAFETY CLASSIFICATION 3

= CREVS is; credited with providing environmental control for personnel comfort 'and equipment-operation,"aslwell as with the protection of control room personnel during radiological and toxic gas -

events.: It is considered to be a safety related system. CCHE integrity is needed to support CREVS in the role of control room personnel protectionLbut-is not considered safety related. The Control Complex s'ructure itselfis seismically qualified and considered safety related, but many of the elements of the habitability boundary (i.e., doors, penetration seals) are not, s

3.0 LICENSING BASIS 3.1 Licensing Background 3 nfl981, the NRC issued an order to FPC confirming the commitments for TMI related requirements I

applicable to CR'#3, NUREG-0737, Item III.D.3.4, Control Room Habitability, was included in this order. 'In response to requirements pertaining to this item, FPC performed a comprehensive habitability evaluation of the CR-3 control room and eventually submitted its findings in the form of the revised CR-

3. Control Room Habitability Evaluation" report, dated 6/30/87. This report concluded that the Maximum Hypothetical Accident was the limiting event with regards to Control Room Habitability and that thyroid dose was the most challenging criteria. The source term applied in the MHA dose calculations is derived in accordance with the U.S. Atomic Energy Commission Technical Information Document (TID) 14844 (Ref 7). Based on methodology consistent with SRP 6.4 guidance, the habitability evaluation found that the MHA would result in a thyroid dose of 26.5 REM. Subsequently, the NRC issued an SER on the findings of the habitability report, stating that the 26.5 REM thyroid.

dose was less than the 30 REM regulatoiy limit, and was therefore acceptable (Ref. 4).

3.2 ITS Requirements Section 3.7.12 of the ITS addresses requirements pertaining to the Control Room Emergency Ventilation System, and requires that two CREVS trains shall be operable during MODES 1,2,3 & 4, as well as during the movement ofirradiated fuel. The LCO associated with one CREVS train inoperable provides for restora:!on of the out of service train within 7 days, or be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . If movement ofirradiated fuel assemblies is in process, '. hen CREVS is to be placed into emergency recirculation mode'immediately or movement of irradiated fuel suspended immediately. With both trains of CREVS inoperable the plant must enter LCO 3.0.3.

iThe ITS bases for Section 3.7.12 discusses operation of RM-A5 in support of CREVS, as well as isolation during toxic gas events. It states that the Maximum Hypothetical Accident is the limiting event with regard to Control Room Habitability in MODES 1,2,3 & 4, and provides a general reference to the 1

. " Control Room Habitability Evaluation" report dated June 30,1987 in this regard. The Fuel Handling EAccident is identified as the limiting radiological event in MODES 5 & 6. The ITS bases also state that

- CREVS ensures that the control room will remain habitable following all postulated design basis events,

?

maintaining exposures to control room operators within the limits of GDC 19 of 10CFR50, Appendix A. ) Notably, no reference or commitment is provided in the ITS or its Bases to SRP 6.4, " Standard 4

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4 7

l-REVISION S l

Review Plan for Control Room Habitability". There are no specific req 2irements regarding CCHE inleakage included in either the ITS or its Bases. Beyond the aforementioned general reference to the

" Control Room Habitability Evaluation" report, no discussion.or reference is given either on the-quantity ofinleakage allowable or the manner in which inleakage is determined and applied. It can be considered an implicit requirement that the integrity of the CCHE be demonstrated as adequate to support CREVS in maintaining operator exposure within regulatory limits.

ITS Section 5.6.2.12 defines requirements pertaining to the Ventilation Filter Testing Program at CR

3, and requirer that the CREVS filtration units meet minimum requirements 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" wg when operating in this range.

3.3 CR-3 Operating License The CR #3 Operating License contains a requirement to maintain Control Room habitability as specified

' in the post-Tl.il requirements of NUREG-0737. However, there is no required specific method for the r

measurement or evaluation ofinleakage.

3.4 FSAR discusson Section 7.4.5 of the FSAR (Rev. 23) provides information on the elements of control room habitability at CR-3. Specific items discussed include:

licensing background, including reference to the NUREG-0737, item Ill.D.3.4 (Control Room Habitability Requirements), the June 30,1987 CR-3 Control Room Habitability Evaluation Report, and the May 25,1989 SER on the same subject discussion on allowable CCHE breach area discussion on CCHE inleakage, including specific reference to comparisen with 0.06 volume change per hour cri*eria In addition to the above, evaluation of the radiological dose consequences of the MHA and other DBAs are found in Chapter 14 of the FSAR. However, Control Room dose consequences are only assessed for the MHA. The consequences of the MHA are presented in section 14.2.2.5,10 and Table 14-54.

This section also includes specific discussion of inputs into control room operator dose calculations, incluaing CCHE inleakage, assumptions for ECCS leakage, ABVS filtration efficiency (0%), and post-accident meteorology.

4.0 IMPACT ANALYSIS AND RELIABILITY CONCERNS CCHE integrity and CREVS performance are primary inputs into the evaluation of control room habitability. The 26.5 REM exposure found in the habitability report and the NRC's SER on this

- subject arc based on assumptions regarding these parameters. No specific value for CCHE inleakage is found in the ITS or CR #3 Operating License, but it is implicit that CCHE integrity must be such that dose limits are not exceeded ITS Basis 3.7.12 does provide a reference to the habitability evaluation report only in identifying that the MHA is the limiting DBA from the standpoint of control room _ dose

l REVISION 5 l

consequence, but does not list or reference any specific inputs or assumptions contained in the report.

Therefore, revised analyses could change inputs into the Control Room Habitability Evaluation report such as the determination and application ofinleakage without affecting the ITS or its Bases, so long as it did not invalidate the conclusion that the MHA is the limiting DB A of record.

ITS Section 5.6.2.12 defmes requirements pertaining to the Ventilation Filter Testing Program at CR

3. 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 pressurc drop across the filtration unit be less than 6" wg when operating in this range. Requirements for the Ventilation Filter Testing Program as implemented by CP-148 require that in place testing be performed at flow rates within 43,500 cfm +/- 5%, with the tighter limits being conservatively imposed for testing inaccuracies.

Revised dose calculations (Ref 8 & 9) have been performed considering a flow rate of 37,800 cfm, (corresponding.to a filter dP of 4" wg1 which is well below the 41,325 cfm corresponding to the lower end of this range. The impact of this lower flow rate on CIGVS operability requirements and control room dose consequences must be evaluated. Also, the apparent disparity between the design flow rate in the ITS at d CREVS performance with current plant configuration must be reconciled.

5.0 DESCR!PTION OF IDENTIFIED CONCERN System Readiness Reviews conducted in 1997 during the CR-3 Design Basis Outege identified several issues which potentially impacted control room habitability. The predominant issue pertained to the validity of assumptions for CCHE inleakage. CCHE integrity and CREVS psformance are p'imary inputs into the evaluation of control room habitability. The 26.5 REM exposure found in the habiiability report and the NRC's SER is based on specific assumptions regarding these parameters.

The June 30, 1987 habitability report evaluated CCHE integrity based upon calculation methods found in SRP 6.4, which determined that up to 355 cfm of in!:akage could be tolerated without exceeding the licensing dose limit. Of this,285 cfm was assumed to be through unfiltered pathways, and 70 cfm was assumed to be filtered inleakage through inlet dampers AHD-1 and AHD-lD. This value ofinleakage was applied as a constant in control room operator dose calculations. Testing identified that, contrary to the assumptions found in the Control Room Habitability Evaluation report, differential pressures across the CCHE inlet dampers was such that leakage was occuning outward, which would correspond-to a like amount of unfiltered inleakage at other boundary locations. In addition, differential pressures at all boundary locations differed from that assumed to the extent that leakage less the 355 cfm lindt could not be conclusively demonstrated. This condition was reported to the NRC in LER 97-022.

Piior to the current analysis, inleakage was assessed on the basis of guidance found in SRP 6.4, to which FPC is committed "for guidance" caly. Using this method, inleakage was assessed by summing the calculated leakage past CCHE boundary elements (i c., doors, penetrations) at a differential pressure of 1/8" wg, then dividing the result by two. Additional " enhancements" were made for inleakage past ventilation system boundary dampers which might be operating at a higher differential pressure and for access / egress. This inleakage value was then applied non-mechanistically into dose calculations to determine operator exposure.

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.. If a non mechanistic baseline inleakage value were obtained by correcting the tracer gas test results to 1/8" wg, and this value applied as a constant value in dose calculations (as was done previously),

the acceptance limit for dose would be exceeded. To calculate dose based on test results, FPC developed a model which predicts inleakage under various differential pressure conditions and so can assess 6

n 3

4

- REVISION 5 '

l

'i

_ mechanistically induced inleakage under postulated post accident conditions._ The' measurement ofinleakage

> and its mechanistic application'in dose calculations is a significant change from previous' methodology and is i

expected _to produce a result which is more realistic, but might_be_less conservative. This issue has been l

. detennined to represent an unresolved safety question, and is the primary focus of this JCO.

. LER 97-022 also' identified that past modifications were implemented which added resistance' to Control -

Complex Ventilation System ductwork without fully assessing the impact on recalculation' flow ratei i

- (Ref. 6). As a result, the system flow rate with clean Alters is now somewhat lower than the 43,500 cfm

- nominal design flow rate referenced in the habitability evaluation and used as an input to calculate the '

245 REM exposure. Subsequent dose calculations have addressed flow rates _as low as 39,150 cfm

- (43,500 '- 10%), but evaluation of current conditions predict flow rates on the order of 37,800 cfm

= assuming 4" wg drop across fouled filters. It is noted that the 4" filter fouling value is less than the f currently reflected in the'ITS and taken as the combined (HEPA and charcoal) filter fouling limit.

Current procedures constrain operation within 43,500 cfm t/. 5% and ensure that the 37,800_ cfm

, minimum flow rate requirement for dose calculations is met. However, the use of a filter fouling limit which is less than ITS requirements must also be assessed.

' Finally, numerous changes to assumptions / inputs associated with Control Room Habhability have been implemented under 10CFR 50.59 since the habitability report was issued in 1987. Generally, these changes have been assessed individally at the time of their implementation, but without updates to the habitability

_ report to mainta!n this licensing document consistent with plant design on a real time basis. _ This JCO includes a matrix which lists the former and current values of habitability evaluation inputs, identifies the iterations associated with each parameter up to the present point in time, and provides a brief discussion for the basis and acceptability ofeach change.

6.0 OPERABILITY EVALU6TlQE The activities associated with Restart Issue R-12 have improved the performance of CREVS and the CCHE in assuring control room habitability. Redundant bubble-tight dampers have been installed which assure positive closure of ventilation flow paths Vestibules have been installed on all CCHE boundary doors which provide an extra measure ofleak resistance and provide defense in depth for normal operation and wear. An extensive penetration sealing program was conducted to improve boundary leak tightness. Leakage of the habitability envelope has been accurately measured under actual operating cond:tions using tracer gas test methods;

. The' modifications and efforts driven by 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

-thess revised calculations is consistent with that found in regulatory guidance and utilized. in previous y

alml=* ions (i.e., Reg. Guide 1.4 { TID-14844} source terms, RB modeling per SRP 6.5.2, meteorology per Murphy-Campe). However, many of the inputs found in the Control Room Habitability Evaluation report Lhave been' superseded on the basis of actions associated by R-12, and still others have been revised under 10CFR50.59 since its submittal _in 1987. _ A detailed comparison of significant inputs into current Control

~

Room Habitability analyses vs. those found in the 1987 habitability evr.luation report is provided in Table A to thia JCO. Notably, the' methodology for determination of CCHE inleakage and the manner in which L

inleakage values are applied in dose calculations is a significant change from previous methodology, including -

that described in the Control Room Habitability Evaluation report.

1.

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REVISION $

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Previous dose calmiminns were based on a model defined in NRC regulatory guidance'which determines intenkage at a Control Complex pressure of 1/8" wg., then applies this inleakage as a constant value in done ;

calculations.LThis model is.a good correlation for more common conhol room designs. which rely on taking contaminated air ftom' 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_ is used to prevent inleakage, and is generally accepted as being high'enough to overcome the effects of wind, thermal l

4 effects, and operation of ventilation systems in adjacent structures.1 This model does not correlate well to the.,

- CR 3 habitability envelope, which is a filtered recalculation system with isolation (no makeup). In this design,

_ the bulk pressure in the envelope is neutral,- and the 1/8" differential' pressure in the NRC's model is used to

drive inleakage across the envelope boundaryc The quantity ofinleakage induced in this manner is used to '

develop a baselinefmlenkage value for use in control room operator dose calculations.

FPC has developed ~its own model for detuiisning CCHE inleakage for the purpose ofinput into control

. room operator dose calculations This model is based on more realistic, but still conservative, considerations 4 of the actual motive forces which would exist for driving inleakage under postulated post-accident conditions.

cThe use of this model is a significant departure from previous methodology and is not described in regulatory

^ guidance, and is considered to be an unreviewed safety question on that basis. The existence of an=

unreviewed safety question does not necessarily mean that an activity or system condition is unsafe Rather, _

l L

= it is accessary for FPC to evaluate the safety of the condition and to assess its implications on unit restart as provi: led in Generic Letter 9118, Rev.1. As detailed within this JCO, FPC and its contractors bnve reviewed the treatment of CCHE inleakage in control room habitability calculations and concluded that the change in methodology is safe and that CREVS operability and CCHE integrity are not compromised.

Ultimate resolution of the unreviewed safety question requires either approval by the NRC or placing the system in a condition which has been resiewed.

7.0 JUSTIFICATION FOR CONTINUED OPERATION Relative to the issue of control room habitability, justification for continued operation in all operating MODES is based on establishing the operability of the CREVS and the adequacy of CCHE integrity.

Additionally, the existence of other mitigation features, such as vestibules at CCHE boundary doors, source term reduction by the Auxiliary Building filters, and Radiological Emergency Response Plan dose management procedures provide additional assurance of control room personnel protection. - A discussion of the basis for operability of the CCHE follows:

I 7.1 CREVS Onerability Operability of CREVS requires that the specific requirements of ITS Sections 3.7.12= and 5.6.2.12 are satisfied and that the bases for these technical specifications are not invalidated. Surveillance Activities required to demonstrate CREVS operability include (1) operating each CREVS train for at least 15 minutes each month, (2) satisfying the ventilation filter testing program, and (3) verifying that each CREVS train actuates to the emergency recalculation _ mode on an actual or simulated actuation signal

every' 24 months., Of these criteria, the first and last will be included in post-modification testing

~

. associated with damper replacement; The second criteria refers to the requirements of the ventilation' filter test program defined irilTS Section 5.6.2.12; The ventilation filter test program requires that the CREVS filtration units meet minimum requirements regarding penetration bypass and adsorption Section 5.6.2.12 of the ITS also requires that the c

pressure drop across the combined HEPA~ and charcoal filters be demonstrated as no more than 6" wg Twhen tested at a system flow rate of 43,500 cfm. LER 97-022 identified that past modifications were-implemented which added resistance to Control Complex Ventilation System ductwork without fully y

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REVISION $

l anaa=Ag thr[ impact on recalculation slow rate. 2 As a result, the system flow rate with clean filters is =

now somewhat lower than the 43,500 cfm_ nominal design flow rate. This is evident in a recer.t flow -

evaluation of system performance which a calculated recalculation flow rade of 37,800 cfm with 4" wg

across fouled Alters.-

The use of-4" wg filter fouling limit is less than the 6"wg value currently reflected in ITS section 5.6.2.12 and taken as the comleined (HEPA and charccal) filter fouling limit Since a 4" wg differential' pressure across the filtration unit is expected to correspond to 37,800 cfm, it follows that the flow rate associated with the 6" fouling limit would be lower still. However, current surveillance procedures

constrain operation to flow rates at or greater than 41,325 cfm (43,500 - 5%).1 Operation within this constraint has the effect oflimiting filter fouling to a much tighter range than even 4" wg, and ensures both that the level of filter fouling is within the 6". wg ITS limit and that the 37,800 cfm minimum flow

rate postulated for dose calculations is met. License Amendment Request, LAR #222 addresses this

lasue in the ITS, such that CREVS flow rate and allowable filter pressure drop are reconciled.-

Reduced' flow rate also has the potential to degrade CREVS equipment cooling capability. Issues pertaining to CREVS cooling capability are being addressed _ separately by Technical Specification

- Change Request Notice (TSCRN) #210, and are not included in this JCO.

This LER also identified that charcoal testing has been performed in the pt c asing non-conservative temperature and humidity conditions.

Charcoal adsorption capability incesses with increasing temperature and decreasing humidity. Prior to this nutage, laboratory charcoal testing has been

. performed at a 80 C and 70% RH. However, CREVS system readiness reviews determined that this criteria was unconservative with regard to possible post accident conditions. To address this concern the ventilation filter test program is being revised to require laboratory charcoal testing at 30 C and 95%

RH. Prior to restart, laboratory testing will be performed to both the old and new test conditions.

7.2 CCHE Intenrity Not a system per se, the CCHE is the physical boundary which is credited with protecting the control room operator from the effects ofpostulated DBAs or toxic gas events. The CR-3 control room ventilation system

~ design is categorized according to SRP 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 leak tightness and fihration capability to provide the neeaary level of protection. Operability of the CCHE is predicated on demonstrating a level of integrity such that, in conjunction with the operation of CREVS, sufficient protection is provided for the control room operator to ensure that exposure limits associated with DBAs and toxic gas events are not exceeded 7.2,l Inleakage Detennination:

Previously, CCHE inleakage has been determined on the basis of methodology described in SRP 6.4, which' provides a standard means by which to derive a " baseline" inleakage. The basic SRP 6.4 methodology for

determining base inleakage in a recalculation 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 subjerf to test pressure and provides a conservative (but relatively arbitrary) baseline inleakage value. The 1/8" value is not associated with ar. particular post-accident conditions, but;is large enough to minimize the impact of test inaccuracies and local pressure effects. The wind 9

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I il-REVISION S

=l 1

speed necessary to generate this differential pressure is on the order of 14 - 16 mph; much larger than lthe low wind speeds associated with the 5% worst x/Q value used in SRP 6.4 to minimize source -

j J dispersion, SRPE 6.4 ' test i methods would include additional enhancements for 'inleakage through-l boundary dampers.which may be subject to unusually high differential pressures, and a 10 cfm!

l allowance for personnel access /. egress during an sccident. S'mce FPC folloves SRP 6.4 only as a q

'" guidance" document, there is no commitment to adhere to this methodology, nor is there any requirement to assess building inleakage at a differential pressure of 1/8"_ wg for. the purposes of

j evaluating Control Room dose consequences.

3 Although not prescribed by regulatory guidance, application of tracer gas technology is~ recognimi as a means to accurately and directly measure building inleakage and is being increasingly utilized in the nuclear industry for this purpose. Using tracer gas test methods, it is posmtie to set up.a test to measure intenkage

~ under conditions which are representative of a specific postulated scenario. Determination of inleakage based on testing under simulated post-accident conditions is a departure from the exim*mg licensing basis for -

-CCHE inleakage, but is justifiable given that this methodology is expected to provide a more accurate i prediction ofdose consequences and that CR 3 is not committed to SRP 6.4 except as a guidance document.

Tracer gas testing under post accident conditions would have the Control Complex in its emergency

- L,.alculation mode, and treat the entire CCHE as a lumped volume. No additional penalty would bc

)

required at boundary damper locations because these dampers would be subjea to the same pressures during testing as would be expected during post-accident operation. (Note that the latter effect is inconsequential for CR 3 given that bubble-tight dampers will be installed at all boundary isolation locations.)~ The use of a 10 cfm allowance for access / egress is still applicable,- and would be incorporated into the final inleakage result.

j D:veloping a test replicating post-accident conditions requires that the scenario be postulated which provides both realistic and challengir.g conditions from the standpoint of exposure to source term and 4

maximizing the differential pressure which drives inleakage. The postulated conditions may result in a variance of differential pressure across individual CCHE elements, and are not intended to necessarily test all penetrations to a certain min mum differential pressure. Rather, the objective is to develop a i

1

- test which realistically and conservatively gauges the overall consequence associat-d with the limiting-1 post-accident scenario. The motive forces ~ which might induce a significant differentN pressure across the CCHE are taken as wind pressures (assuming a loss of offsite power) and ventildion systems in g

adjoining structures (no loss of offsite power). Significantly higher differential _ pressee would be expected assuming no loss of offsite power, but the source term would be lower given that this would g

. necessarily require the Auxiliary Building Filtration System to be in operation.

Therefore, to fully

assess limiting post-accident conditions requires that both scenarios be evaluated. This is accomplished by measuring inleakage at a known differential pressure using tracer gas methods, then analytically adjusting this value to correspond with postulated conditions.

4

- 7.2.2 Test Conditions Tha tracer gas test model is established based on consideration of site layout, source terms and possible plant

= operating conditions.= In the event of a MHA w / LOOP, 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 / south wind directions would tend to maximize CCHE inleakage. Similarly, with no loss ofoffsite

) power, the: Auxiliarp Building supply fans would be ocured by radiation monitor RM-A2, causing the -

. Auxiliary Building to develop a negative pressure and inducing leakage through the CCHE in that direction.

^

lIn"either case, tk tracer gas test which-models these conditions would utilize the -Auxiliary Building 10 x

a a.

==

=

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l REVISION 5 l

1 LVentilation System to induce CCHE innenlage by creating a negative pressure in'the Auxiliary Building. The y

following conditions wera prescribed for the tracer gas testing which modeled CCHE 'mleakage:

The Control Room " Emergency Ventilation; System _ (CREVS) was placed in -

l

- emergency recalculation mode and operating normallyf Both " Toxic Gas" and the-l "High Radiation" recalculation lineups were tested. All CREVS boundary damper locations were sealed " bubble tight" to duplicate-post modification conditions.

(NOTE: The isolation dampers were tested in-place and the test boundary exhibited :

~

insignificant leakage.);

All fans-in the Turbine' BuildingL Ventilation System were secured. The turbme 3

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.

l The-Aux. Building Ventilation was operated to test pressure of approximately _

C 0.171" wg negative pressure vs. the Turbine Building. This value is large enough to -

minimize test inaccuracies and external effects and was sustained for the duration of the test,-

The test was conducted on backshift when personnel trafUc is minimized. Since a 10 cfm allowance fbr access / egress would be analytically applied, minimizing

. trafHe precludes additional penalization for this effect.

Testing-was conducted with vestibule doors blocked open. This conservatively 7

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 periodically filled during plant operation.

Tracer gas testing conducted under these conditions measured an inleakage of 462 cfm in "High Radiation" recalculation mode at a differential pressure of 0.171" wg. Using this information, the inleakage at other differential pressures can be predicted by the use of the formula Q = C AP", where Q = sir flow in cfm -

C = inleakage coefficient P = differential pressure, and -

n = flow exponent.

z

> According to ASHRAE, values of n are typically between 0.6 and 0.7. Values less than 0.171 are -

estimated by

!ng'n = 0.5, which gives conservative results for estimating inleakage at pressures

'less than the tot value (Ref 8)c For extrapolation to conditions above 0.171" wg, the use of n =

,0.5 is somewhat unconservativ6 and a more realistic value of n = 0.65 is chosen (Ref. 9).

i-11-

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REVISION 5 l

7.2.3 Analysis of MHA w / LOOP (Ref. 8)

Since wind pressures are assumed to be the primary motive force under MHA w / LOOP conditions, inleakage for this scenario is determined by examining ?neteon :ogical conditions associated with event analysis. SRP 6.4 methadology 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 tV radioactive plume as 'it is carried from the conta! ament building to the Control Complex.

The methodology then allows for three incremental increases in wind speed and direction over the duration -

of the accident due to the extreme improbability thst these initial wind conditions would be sustained '

over an extended period of time. Based on these considerations, inleakage values are derived for each of the four time intervals over which x/Q values vary by correcting inleakage at the test differential pressure to the differ-ntial pressure induced by the wind speed associated with that interval. These wind induced differemial pressures were consenatively calculated using ASHRAE methods. 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.

For the MHA w / LOOP, it is noted that the use oflow wind speeds prmide relatively small motive force for inducing leakage through the CCHE. However, a parametric study (Reference 8 Attachment G) shows 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.

7.2.3.1 Leakage Created by Secondary Effects 7.2.3.1.1 Stack Effect It is also noted that, at these relatively low wind speeds, the potential efTects of thermally induced inleakage becomes significant. Differential pressure across walls induced by differences in inside and outside temperatures (i.e., stack efTect) can be pronounced in tall structures, as 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. Reference 8 section 5.4.3, " Thermal Induced Inleakage", discusses how this efTect is quantified and combined with the wind driven infiltration. The following highlights the conservatisms used ia calculating the " stack effect".

a) Winter Conditions The following conditions were used to conservatively evaluate the winter stack effect. The control room is at its hot weather design temperature,75 F and the adjacent buildings a: st 31*F for the entire 30 day accident. The 31*F is the ASHRAE published 99% Design Dry Buh DDD) Temperature for 1

Ocala, Fla. These are very conservative temperatures since neither the Turbine or Auxiliary Buildings could be at 31 F when the plant was in MODE 1. Also, the 99% DDB Temperature occurs for only 88 hours0.00102 days 0.0244 hours 1.455026e-4 weeks 3.3484e-5 months annually (1% of the time), whereas this scenario rnaintains it for 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months .

b) Summer Conditions The following conditions were used to conservatively evaluate the summer stack effect. The control room is at its design temperature,75 F and the adjacent buildings are at, or below,118 F for the entire 30 day accident. Per Reference 8 section 5.4.3, the Turbine Building summer transient temperature at the start of LOOP is 20 'F higher than the summer time outside design temperature of 95 F (i.e.

12

0 l

REVISION $

l 115'F) and. the Auxiliary building is 4*F above ambient (99'F). As a conservatistn, it is assumed that the 118'F is maintained for the entire 30 day accident. (Reference 8 does consider the temperature transients expected due to the LOOP's effect on the ventilation systems.)

An average stack effect wall pressure was calculated, and inleakage associated with this value determined by application of relationship between building dif5 ential pressure and inleakage derived t

from tracer gas test results. This value was then combina to wind induced inleakage using the ASilRAE formula Q... - (Q,,' + Q,')"

Where: Q and Q, are the wind and stack e.T+;t leakage terms.

7.2.3.1.2 Leakage Induced by CREVS Operation In addition to the CCIIE in leakage resulting from the wind load and stvk effects, the operation of the CREVS will introduce some leakage due to the difTerential pressures gc.ierated within the CCliE. This pressure dis:ribution resu4s in areas of slightly positive pressure due to the discharge from the ventilation supply ducts and slightly negative pressure due to flow into return ducts. In the areas of positive pressure, any leakage will be out of the CCllE, which will not contribute to the operator dose.

The inleakage that occurs due to flow into the return ducts will be circulated through the CREVS charcoal (;lter prior to reh rning to an occupied area.

i The specifl: distribution of pressure within the CCliE has not been quantified, liowever, its impact on inleakege was measured as part of the testing performed to quantify the overall CCIIE in leakage.

Duririg this test, the CREVS was operating so the effect of any CREVS induced leakage was mea >ured as part of the total 462 cfm of measured in leakage.

Total h. leakage = Differential Pressure inleakt.ge + CREVS Induced Inleakage The leakage rest.! ting from CREVS operation is independent of the other CCliE inleakages since it is a characteristic of the ventilation system. The msximum possible CREVS induced inteakage would result if all of the inleakage measured during the testing was the result of CREVS operation. (i.e. 462 cfm).

As modeled by the Murpy Campe guidance, the CREVS induced i leakage is analogous to return n

actwork inleakige and is filtered by the CREVS before it contrioutes to the control room dose.

'Therefore the maximum possible contribution, in terms of unfiltered in leakage, from the CREVS operation would be:

462 cfm x 0.05 = 23.1 cfm using the 95% filter effectiveness of the CREVS charcoal filten.

Some location specific mixing of the CREVS operation induced !nleakage with the CCIIE could occur prior to it being filtered. This would have an increased impact on the calcriated dose. Therefore to address this effect and produce a more conservative dose assessment, the value for the unfiltered inleakage due to CREVS operation was assumed to be 125 cfin, approximately 25% of the total inleakage measured by the testing.

13

l REVISION 5 j

i This value was directly added to the wind induced in leakage in the dose calculation (i c., the in leakage determined from the total in leakage measured during the test). In determining the wind induced inleakage, the total inleakage during the test was not reduced by the assumed CREVS induced f

inleakage This approach ensures that the CRBVS operation impact is bounded by the dose calculation.

[

(Refs 8 & 11 apply to above discussion.)

7.2.4 Analysis of MilA without LOOP (Ref. 9)

Given the occurrence of the MilA without a loss of offsite power, the ventilation systems in~

adjacent buildings are assumed to continue to operate during and after the accident. Increasing i

levels of raoiation in the Auxiliary Building, as sensed by radiation monitor RM A2, would result in a trip of the Auxiliary Building Ventilation System (ABVS) supply fans, resulting in a significantly larger negative pressure _in the Auxiliary Building.

The Turbine Building is considered to be est,entially at atmospheric pressure due to the numerous laige openings in that

. structure. Under these conditions, the post accident leakage into the Control Room could be significantly higher (especially during the early time steps) than that postulated on the basis of wind pressures (i.e., MilA w / LOOP).

The releawe patl for this 3cenario la based on the activity being released from the Containraent and subject to initist <!!spersion as it travels to the Turbine Building Ventilation System intakes and into *he Turbine Building. From that point it ultimaiely enters the Control Room as unfdtered inleakage by the difTerential pressure induced across the Control Complex. This release path model accounts for the time dependent rate of change of activity in the Turbine Building volume as well as minor decay and holdup while the activity is in the Turbine Dvilding.

The evaluation of MilA without LOOP has four distinct changes from the version of the event which assumes LOOP; (1) given that the AllVS must be in ope,ation tu induce the difTerential pressures of concern, then filtration by the ABVS charcoal fdters occurs and there is no requirement to assume an ECCS pump seal failure at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the accident with a leak rate c'50 gpm for 30

minutes, (2) the normal ECCS leakage which does occe is assumed to be filtered to 75% efliciency, 1

(3) the activity 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 (4)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 CCisd towards the Auxiliary Building.

Temperature effects in this. scenario are assumed to be insignificant given that continued operation of adjacent ventilation systems minimizes the temperature differentials between taese areas and the Control Complex. The 75% efliciency assumed for the Aux. Building charcoal l

filters is consistent with that recently allowed for these filters by the NRC in control room habitability analyses. (ref, discussion on the ABVS in ' fable A) Inleakage induced by CREVS 14

~

l REVISION 5 l

operation is also ignored on the basis that conditions at the time of tracer gas testing are similar to those postulated under these post accident conditions, such that this factor was present during in the tests. As with the MIIA w / LOOP analysis of this scenario assumes that inleakage is distributed evenly throughout the CCllE <olume, even though the locat on of the i

penetrations and the design of CREVS ensures that s ery little inleakage is introduced into the Control Room without being subject to filtration. Given this and other conservatisms in the analysis, the treatment of MilA without LOOP described above is considered to be a very conservative treatment of this scenario.

7.2.5 Results ofMilA Analym The results of this analysis shows that the bounding version of the MHA is that associated with the accident occurring with LOOP. Calculations show that a 26.5 REM exposure can be maintained in this scenario whilc allowing an additional CCllE breach area of up to 22.8 in' in addition to that measured by the tracer gas leak test (Ref 8, Tables 6.2, and 7.2). The 26.5 REM value corresponds with that in the Control Room IIabitability Evaluation report dated June 30,1987 (as referenced in.he ITS Bases) and the NRC's f.ER in reply dated May 29,1989, and is taken as a utility imposed goal at CR 3. It is concluded that, given that CCliE breach areas are maintained below the value of 22.8 in', the level of CCIIE Integrity is sufficient to meet operability requirements pertaining to radiological consequences of the MilA.

7.2.6 Analysis of Other Dil As The MilA accident analysis uses TID 14844 source terms and assumes that CREVS boundary dampers are isolated from the outset of the event by virtue of the 4 psi Reactor Building liigh Pressure ES signal. A review of other design basis accidents fer which CR #3 is licensed was perfurmed to verify that the MilA as analyzed above is the limiting event. This review was based on (1) a review of source terms, (2) a review of the means by which isolation of the CCllE is achieved, and (3) consideration of plant operating conditions (i.e., operating MODES) at the time of the event. The appropriate source term for evaluating the effects of other-DBAs is one that is specific to, or counds, the current operating cycle core. For CR-3 this source term is developed by Framatome, FPC's nuclear fuel supplier, using the BURPE code.

7.2.6.1 Fuelliandling Accident (FIIA) Ref.10 The Fila analysis uscs cycle 11 specific source terms (Ref 14.) and starts with CREVS in the Toxic Gas recirculation line-up. (This line-up is a pre-requisite to handling fuel at CR-3). The CRET 3 charcoal filters are placed inservice within 30 minutes following event initiation. The conservative case takes no credit fut AB filtration. The calculated operator dose is bounded by the MilA described above.

7.2.6.2 Steam Generator Tube Rupture (SGTR) Ref.10 Analysis of the Steam Generator Tube Rupture event demonstrates that isolation of the CREVS is not necessary to maintain operator exposures less than regulatory limits. CREVS can remain in its normal ventilation line-up for the entire event and is well bounded by the MilA.

15

l REVISION $

l 12.7 Analysis of Toxic Gas Evtalt (Ref.13)

CCiiE integrity is also required to provide protection to the control room operator in the event of a toxic gas accident. Reg Guide 1.78 provides information 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 2 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 Reg guide allows for detection to be accompli;hed by personnel (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, AP-

$13 provides the appropriate instructions for the operator in the event of a toxic gas accident, including the use of air packs.

Based on previous evaluations, the locations and quantities of toxic gas storage sites at CR #3 which pose a potential liability to control room habitability and must be specifically addressed are listed below:

Toxic Gas container Size and Location Toxic Gas liciper Cooling CRl/2 CR4/5 1

Towers

'l ton cylinders Chlorine 17 tons none SO 50 tons 45 tons 1 ton cylinders The most limiting source of toxic gas is a SO2 tank at CR1 which has been administratively limited to storage of 30 tons 187 0005 Rev. 2, " Control Room SO2 Concentration from CRl/CR2 SO2 40 Ton Tank Failures", is the curreni calculation of record that analyzes this event. Case I analyzes a 30 ton tank rupture, with S'/00 cfm make up flow and a 30 second delay between attaining the detector's set point and the dampers reaching the closed position. Inleakage only develops after damper closure. In this scenario, the event is detected and recalculation is automatically initiated. The control room concentration 2 minutes aller detection is 29 ppm, vs the 36 ppm limit previously established in the Control Room liabitability Evaluation report.

The following equation from NUREG-0570 converts gas concentrations at the air intake to concentrations in the control complex:

C,(tj) = C,(tj.i) + [Co(tj) C,(tj.i)](1 - EXP""")

where:

(tj) = Control Complex concentration at the j* timestep (ppm)

Co(tj) = Air intake concentration at the j* time step (ppm)

W = the ventilation mfiltration rate beforc/after isolation (cfm/60sec/ min) r deltaT = time steps (seconds) 8 V = control complex volume (11 )

When CREVs shifts to recalculation, W changes from the normal make-up value of 5700 cfm to an inleakage value which is around 10% of the initial make-up flow rate. By virtue of this effect alone, subsequent changes in control room concentration occur much more slowly given that the concentration 16

c I

RIwtsioN$

l of tode gas in outside air remains constant. For this reason the most challenging version of the toxic gas scenario is the non mechanistic complete tank ruptare (rather than longer duration releases oflesser rates). This scenario is modeled by a "pufl" for the h.rge quantity of gas which flashes at the tirne of the accident, followed by a longer term " plume", modeling the subsequent evaporation of the remaining spilled volume. The pufTcontains a much higher concentration than the plume, but is of such brief duration that most of this release has passed by the Control Complex at the two minute mark and concentration in the outside air is rapidly dropping. Using this model rnaximizes control room toxic gas concentrations by introducing the highest outside concentration of gas concurTent at the maximum possible flow rate. Control room toxic gas concentration changes after isolation occur very slowly due to the twofold esTect of much lower concentration of toxic gas in tie outside air, being introduced at a flow rate which is approximately an order of magnitude less. The net effect is that the toxic gas analyses are insensitive to small variations in CCilE inleakage, A great deal of conservatism is inherent in the simplistic rnodel described abo',e. To demonstrate this, Cermak Peterka Petersen, Inc., modeled the CR1, 2 and 3 site and performed wind tunnel tests simulating a 30 ton release at the tank, empirically measuring the toxic gas concentration at the control complex intake. This data was input into Sargent & Lundy calculation SL 9929 M 0008 Rev. I with the above formula to determine the Control Complex toxic gas concentration. This analysis determined a S0 concentration in the control complex of 24.9 ppm two minutes aller nasal detection without 2

CCilE isolation, well below the toxicity limit of 36 ppm. Comparison of this result with the previous analyses which predicted a concentration of 29 ppm with automatic isolation approximately 30 seconds into the event, it is readily apparent that a great deal of conservatism exists in the analytical model (FPC is in the process of removing the 30 ton tank from CRI and replacing it with a pelletired system that generates sulfur dioxide gas when needed. This activity will not be completed prior to restart, and the above analysis is valid in the interim period.)

The next most limiting toxic gas source is the liclper Cooling Towers. Currently, there is no SO or Cl2 stored at this location and this is ensured by a Crystal River Unit 1 " Red Tag" clearance No.1997-01543. Prior to releasing this clearance, administrative controls will be in place to hmit the 11elper Cooling Towers to 8 tons of Chlorine and 30 tons of SO. 189 0053 Rev. 3, " Control Room 2

Habitability liciper Cooling Tower Project", is the current calculation of record that analyzes ruptures of these tanks. The calculation analyzes the 17 ton chlorine and 50 ton SO ank ruptures and fouad that 2

automatic isolation was,cquired only to meet the two minute chlorine toxicity limits. FPC has used the new Sargent and Lundy calculation (SL 9929 M-0008 Rev.1) to evaluate the lower quantities: 8 tons of chlorine and 30 tons of 502. Thie analysis combined the wind tunnel results from the CR 1 SO2 tank model and traditional atmorph.c dispersion mathematical modeling techniques to conclude that CREVs could remain in its norma; alignment (i.e., no CCllE isolation required) without exceeding Control Room toxicity limits if up to 9 tons of Cl or 50 Tons of SO were released. Thus, CCilE 2

2 inleakage is of no consequence given the new limits of 8 tons and 30 tons for Cl2 and SO2 respectively.

Sargent and Lundy calculation SL-9929 M 0008 also analyzed the Cl and SO stored at CR-4/S, This 2

2 calculation allows for a 4 ton chlorine release at the CR-4 and 5 cooling towers located 3600 feet from the CR3 control complex intake. There are eight I ton tanks with four in service on a single header at a time. The assumed accident has one tank fail and the other three leak out though the common piping.

With no automatic detection or isolation, the calculated control complex concentration is 7.5 ppm, well below the 15 ppm toxicity limit for this source. The one ton sulfur dioxide tanks were not analyzed due to the sulfur dioxide at the liciper Cooling Towers being more limiting. The amount at CR 4 and 5 is 17

l REVISION $

}

less (50 tons versur, I ton), farther away (3400 feet versus 3600 feet), and has a larger building wake (2 versus 3). Since the calculation allows > 50 tons at the llelper Cooling Towers without automatic detection and isolation, then the CR4 and 5 sulfur dioxide will also not require the same.

Based on the 6bove discussion, it can be seen that the current level of CCilE integrity provides adequate protection foi the control roorn operator for postulated toxic gas events. It is also noted that the updr.ted analyses would support operation without crediting the existing toxic gas detectors.

llowever, it is not the intention of this JC0 to delete these monitors, and they will continue to be installed and surveilled as in the past as an additional conservatinn.

CCilE breach margin in modes 5 & 6 is based on potential consequences of an SO2 release at Units I /

2, for which previous calculations demonstrated that up to 1400 cfm ofinleakage could be tolerated with the CCllE initially isolated, Since revised analyses for this event demonstrate that automatic isolation is no longer required, the use of considerably larger inleakage rates (up to as much as the 5,700 cfm makeup rate) could be justified. No increase in the breach margin for modes 5 & 6 is being undertaken by this JCO, 7.3 Addil10DAlhQltEllYiEtalDLC1 7.3.1 Veilibulra CCllE boundary doors represent a significant g,arce ofleakage into the CC. There are three double doors and three single doors. Vestibules were added to the three CCIIE double doors in 1996, and have proven to be effective in reducing differential pressure across the existing doors. Reducing the difTerential pressure exerted on the boundary doors has two benefits. First, lower differential pressure reduces leakage around the doors, and second, lower difTerential pressure allows the door closers to perfbrm more reliably, During this outage, vestibules were added to the single CCllE boundary doors.

All of the vestibules have additionally been sealed at interfaces with the CCilE boundary making the enclosures more effective in reducing CCllE boundary door leakage.

Control Complex tracer gas testing was performed with the vestibule doors blocked open to assure that the test was conservative. Blocking open the vestibule doors increased conservatism in two ways.

First, normal access and egress was permitted during the tracer gas leakage testing which contributed to the measured leakage. Dose calculations performed in accordance with standard methods include a factor of 10 cfm of continuous leakage to account for access and egress during an accident. This factor was added to the measured inleakage in performing CR 3 dose calculations. Therefore, the effect of leakage during access and egress was applied twice. In a real event the vestibules would not be blocked open. Since they function similar to an airlock, they would be effective in reducing leakage during access and egress. The second manner in which maintaining the vestibules open during the test was conservative is that the existing boundary door was the only barrier to leakage during the test. In an actual event the vestibule doors would be in their normal closed positions, and would be effective in reducing infiltration into the CCilE through the doors. This feature would be particularly effective in reducing operator dose during the MilA without LOOP scenario where leakage is induced from the Turbine Building into the CCllE due to operation of the ABVS.

18

l REVISlON $

l 212. z Auxiliarvjuilding fihration Perforrning dose calculations in accordance with the requirements of NUREG-0737 Item Ill.D.3.4, Control Room liabitability, includes assuming source term; specified in Standard Review Plan 15.6.5.

One aspect of this for a plant that does not have an "ESF atmosphere filtration system,"is leakage of 1500 gallons of water contained in engineered safeguards piping outside of containment must be assumed 24 hotirs following the accident. This amounts to leakage into the Auxiliary Building (AD) of highly contaminated water from the reactor building sump. An ESF atmosphere filtration system is like the CR 3 AllVS with IIEPA and carbon filters, however it is required that an ESF system be powered from an onsite power sourco. Since the ADVS is not powered from an onsite power source, CR 3 control room dose calculations include the required leakage term.

The CR 3 ADVS has redundant fans and filters which are operated continuously during normal operation. Therefore, the system must be maintained in good operating condition. The filters are tested and maintained in accordance with approved procedures that implement regulatory guidance on emergency filter systems. As such, filter efliciency is routinely verified by carbon filter media testing.

Evaluations of the FPC power grid performed for station blackout concerns have demonstrated a high degree of reliability. Studies have shown that power should be restored to the site in under eight hours following a loss of offsite power due to disturbances on the grid. This conclusion is based on the specific features of the FPC transmission system, and power restoretion data from NUREG 1032,

" Evaluation of Station Illackout Accidents at Nuclear Power Plants". Therefore, there is a high degree of assurance that the AllVS will be available and capable of performing effectively to reduce radioactivity released from the All following an event.

Radioactivity is released into the All via the one time 1500 gallon leak (previously described) and a continuous leak from the LPI & Building Spray trains. The continuous leakage rate is specified to be twice the maximum allowed by the unit's procedures. Since the ABVS is not powered from an onsite power supply, neither of these source terms is considered to be filtered prior release to the environment.

Reference 8. Table 7.2 shows that the combined effects of these two sources is responsible for 12.14 REh! of the projected individual control room operator dose of 26.5 REh! during the hillA w/ LOOP scenario.

There is a near linear relationship between All source term and operator dose. Therefore, using a conservative ABVS filtration elliciency of 75% would result in a dose reduction of approximately 9 REhi (12 REh! contribution X 0.75 = 9 REh! reduction) from the 26.5 REhi projected dose for the MilA with LOOP scenario. This reduction in dose is applicable only to the MllA with LOOP analysis.

(It is already being credited in the analysis for the MHA without LOOP (Ref 9).) 75% efliciency for ABVS carbon filters has been previously accepted for CR 3 as an interim measure during resolution of reactor building flood level issues (See AB Filtration in Table A).

7.3.3 Dose hinnagement if a radiological accident were to occur which involved the release of radioactive material from the reactor or spent fuel storage area, the CR 3 Radiological Emergency Response Plan would be implemented. The plan provides for stalling the emergency response organization and establishing emergency response actions commensurate with the severity of the event. Actions required in the 19

r O

l REVISION $

l Emergency Plan Implementing Procedures include dispatching a llealth Physics Technician to the Control Complex to monitor radiological conditions, and to provide radiological and meteorological data to the Dor,e Assessment Coordinator The llealth Physics Technician will perform radiological surveys within the Control Complex, including surveys for airbome radioactivity. Dose Assessment personnel in the Technical Support Center use data collected from surveys to project expected personnel doses. Doses above 5 REM require approval of the Emergency Coordinator. Provisions exist in the Emergency Plan Implementing Procedures for considering administration of potassium iodide (KI) to personnel based on projected dose. A projected dose of 25 REM to an individual has been established as the threshold for considering administration of KI, but it may be administered at lower doses as deemed appropriate by Medical and Dose Assessment Personnel stafling the emergency response organization. (References 15 & 16 apply.)

Control room dose calculations contain very conservative assumptions regarding operator presence in the control room. For example, in accordance with the Murphy-Campe methodology it is assumed that the operator is present in the control room continuously for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then 60% of the time for the next 3 days, and then 40% of his time for the remaining 26 days. Similarly, atmospheric conditions are assumed which channel the release of radioactivity toward the control complex at conditions which max! mite the plume concentration, particularly in the first 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; of the scenario. Based on the conservatisms that exists throughout control room habitability dose calculations, no doses approaching GDC 19 limits are anticipated in a realistic accident. Ilowever, provisions exist in currently approved procedures to monitor actual and projected doses based on measured and observed conditions, and to control personnel exposure. Through the protective features of the control complex evaluated in the control room habitability calculations, and the established emergency response procedures, protection of the control room operators is assured.

20

l REVISION S I

8.0

SUMMARY

/ CONCLUSIONS Based on the above discussion, it is concluded 3. hat, upon completion of certain actions, CREVS performance and CCllE integrity is adequate to (1) protect the control room operator in toxic gas events and DBAs for which CR #3 is licensed, such that regulatory limits are not exceeded, (2) meet operability requirements dermed in the ITS, and (3) not invalidate the assumptiens and conclusions of the ITS bases. Meeting these requirements is deemed adequate and appropriate basis for plant operation in any mode with regard '.o control room habitability. Specific actions which must be implemented and the timing with which they must be completed to support this evaluation are listed below:

=4)

Cc pte: =cd!5=:!=: =d pc:: r=dif=:!= :: ! g r:=!:::d d:h ap!=!:; CREV8 be=h j I = pre =d !=p ovemente4eCC".E !, :g.!.y-Thie+e64en4nud k==pk::d pr'=

le+4GDE4-(Complete) 2)

Complete test requirements associated with the Ventilation Filter Testing program defined in ITS 5.6.2.12. In addition to satisfying current ITS test!ng requirements, charcoal testing is to be performed at conditions of 30 C and 95% Ril, consistent with LAR # 222. This action must be completed prior to MODE 4 3)

Complete surveillance requirements associated with ITS Section 3.7.12. This action must be completed prior to MODE 4 4)

&bm!: LAR42234oiequeW4icensing-changee+asociated-with-GREVS;-ine'ud!:; ch=ce:!4est requiremente,-+onveillance ::quiremento-on{4HE4ntegrity;-breechmergin-end4eec=!!!::ica of GREV841ow-rate +nd-Giter-dP-Thio-is-e44k>wup+etion-which+houkit: compktedie+oon+s f:=!bk, befor-which*o+peci6ed-completionalmeissiven: (Complete) 5)

Administrative measures must be in place to ensure that CREVS is in recalculation mode prior to fuel movement until such time as updated Fila analyses finds this is not necessary. At that time, this JCO can be revised as appropriate. This action must be completed prior to any movement of fuel.

6)

Administrative measures must be in place to ensure that quantities of toxic gases stored on site do not exceed the limits evaluated herein. CR 1/2 " red tag" clearances are currently in effect.

These clearances or equivalent measures are to remain in place.

7)

Ad6k:=:!v: measuree Fe::!?:;:c !cep :=b4num be upd:::d c :=! ::!n th: nw!y instelled+eelefull--Thie+etionanum bc =mpk::d r m :0 MODE-4 (Complete).

8)

Procedures controlling allowable CCIIE breach margin must be updated to te!)ect an breach margin limit of 22.8 in.'in MODES 1-4. Breach margin limits for MODES 5 and 6 will remain at their previous levels for the time being. This must be completed prior to MODE 4.

9)

A!!'c=d:r C:kuht!=: =d :c=== DBA-ecatro! rec:n cge::or dc= :=!y;;; :=-:he

-i=:d+nd verified-Thie+etion-mum 4e<cmpk::d prior 4o-NGRG-epproval-of-this-DR4 40 0: - (Complete)

Wr-All+endor4eleulationeesed4e+seese-Toxic g= accident centtol+oom-operator : pc:are

=:!yn; = : be r:v!=:d =d verified-This+etioneudh ec=pk:ed prior-to-MODE-4 (Complete) 21

t ;,

j

'i I

anvisx w s I

{

i

9.0 REFERENCES

1. FPC CR#3, MAR 97207-05 01 " Control Complex Emergency Ventilation"-

2c CP-148, Rev. 2, " Ventilation Filter Testing Program," FPC CR3, 8/30/96.

l

3. FPC SA/USQ For Sargent & Lundy Calculation SL-9929 M-009 R1 l

4. NRC letter 3N0589 25 to FPC dated 5/25/1989, _" Crystal River Unit 3 Control Room Habitability

{

Evaluation (NUREG-0737 Item lli.D.3.4) ( TAC No. 64805)"

5. ' License Amendment Request (LAR) #222 RO, " Control Room Emergency Ventilation System and 1

Control Room Habitability" i

-6.

FPC CR#3, MAR 77-0411," Control Complex Ventilation Cooling Coils AHHE" 7 FPC IOC Ser NL97-0424 did 12/20/97, Fm D. Kunsemiller to M. Rencheck 1

Re: MHA Source Term and Dose Limits for Control Room Habitability

- 8. FPC Calc No. M97 0110 R/2 (SL-9929 M 0009 R/2)

Control Room Operator Dose due to MHA w/ LOOP -

- 9 _ FPC Calc No, M97-0137 R/2 (FPC-CED M 01 R/2)

Control Room Operator Dose due to MHA w/o LOOP l

10._ FPC Calc No. M97-0111 R/3 (SL-9929-M 0010 R/3)

Control Room Operator Dose due to SGTR & FHA

?

11. MPR ltr did 12/10/97, Fm J. Russell (MPR) to M. Clary (FPC)

Re: Third Party comments on validation ofinputs to SL 9929-M-0009 and 125 cfm of CREVS induced Filtered Inleakage

12. NUS ltr Ser CED DAS 97178, did 12/15/97, Fm D. Studley (NUS) to M. Clary FPC Re: Third party assessment of thermal stack effect presented in SL-9929-M-0009.

13. FPC Calc No. M97 0109 R/l (SL 9929-M 0008 R/l)

Toxic Gas Analysis

14. FTI ltr Ser INS 97-4997 dtd 12/19/97, Fm L. Lesniak (FTI) To M. Clary (FPC)

Re: Cycle 11 Source Terms-

15. EM 102," Operation of the Technical Support Center"

16. EM 219 " Duties of the Dose Assessment Team"

17. FPC Calc No. M96-0043 Rev 1 (FTl Doc No. 861257374-02), " Letdown Line Break" l 8.- USNRC ltr Ser 3N1297 19, did -12/24/97, Fm L. Raghavan (NRC) to R. Anderson (FPC)

Re: Interim Assessment Results of Justification for Continued Operation

19. Scientech ltr Ser CED DAS 97 196. did 12/31/97, Fm D. Studley (NUS) To M. Clary(FPC)

Re: Isotopes Currently in Core; Justification for Operation in Modes 3 and 4.-

20. FPC ltr Ser 3F1297 39, dtd 12/15/97 Fm R.E. Grazio (FPC) To USNRC Doc. Control Desk p

4 i

22

.a

.:,:=...

l REVISION 5 l

Comparison ofinputs to Control Room liabitability Analyses Parameter Value in Value in Conunents 6/30/87 Current Submittal Analysis Reactor 7I seconds 124 seconds Revision 3 to Calculation 186-0003 (dated 7/6/93) used a two Building Spray minute RD spray delay time based on request from FPC. Since Actuation then 186-0003 has been revised several times and uses 124 Time seconds as a conservative RH actuation time. This value is obtained by using 120 seconds for RD spray actuation plus 4 seconds for RD pressure to go from 0 psig to 30 psig after a 1.OCA.

More realistic values for RD Spray initiaticn time are found in Calculation M94-0004 Rev. 0 (dated I/26/94), which determined the full RD spray actuation time from initiation, to diesel start, including block loading, pump starting, header fill time and time to reach full flow. Calculation shows RB spray A reaching full flow in 81.1 seconds and D train reaching full flow in 86.1 seconds. His calculation modeled the spray system completely and included all the maximum espected delay times.

Reactor 1500 gpm i112 gpm in the 6/30/87 liabitability Evaluation, RH spray flaw is Building Spray described as full flow (3000 gpm), half now (1500 gpm). No Flow Rate differentiation was made between initial injection and recalculation flow rates. Reviewing OP 405 Rev,31, RB Spray System. which was in effect in 1987, has recalculation spray flow set at i150 gpm to 1250 gpm.

Calculation 190-0022 Rev. O, 3/12/91, determined that with RD spray controller set at 1500 gpm (during initial injection),

the actual RB spray flow could be as low as 1397 gpm considering instrumentation error, in recalculation with RB spray controller set at 1200 gpm, tim spray flow could be as low as 1112 gpm.

Calculation 186-0002 Rev. 5,1/16/96, determined containment spray removal corutants using the new instrument error corrected flow values of 1397 gpm (injection phase) and 1112 gpm (recalculation phase). Spray constants associated with the lower value of I 112 gpm are used in revised dose calculations.

He instrument loop uncertainties for spray flow indication and control were being reviewed concurrent with performing the revised dose calculations. As a conungency, the revised dose calculation looked at a containment spray flow rate of 1000 gpm and found that it was essentially the same as the 23

l REVISION 5

}

1112 gpm case. He calculation concludes that containment

(

spray rate of 1000 gpm can be tolerated. (Ref 8, sect 7.3.4).

Balliet to Widc!! ltr ser NOE97 23 I I dtd i1/11/97, shows l

that when spray is being supplied from the RB Sump, the actual flow may be 121 gpm below the indicated now of 1200 rpm. Hus, the lowest value may be 1079gpm l

Reactor 490,182 gal 343,347 gal ne habitability submittal assumes the liquid sump volune Duilding Sump (65,532 ft3)

(45,902 ft3) was as 490,182 gallons (7.48 gal /A' or 65,532.353 fl') His Volume volume was Calculation 186-0003 Rev.1,5/2/91, referenced GCI calculation DC.5515 0841.ME, Rev. O, dated 3/26/90 that calculated new RB sump volumes based on climinating NaOH tanks and switching to TSP baskets (MAR 88 05 01 01). New volumes were based on cubic feet and were referenced to 130' F. New volume was determined to be 500,718.7 gal or 66,941 A'.

Calculation 186-0003 Rev 6, 3/30/95, then switched to 45,902 ft' or 343,347 gallons. His Ogure was the output from Calculation M95-0007. An important design reference for Calculation M95 0007 was Calculation M95-0005, Minimum BWST Level to Prevent Vortexing Rev. O. E0P 8 swaps from BWST to RB sump starting at 15'. An instrument error of 1.2' was used in BWST level calculations. E0P 8 requires swapping over when BWST is less than 15' and has to be complete by 7' to prevent BWST Vortexing. (5.5' from Calculation M95 0005)

Rese low level considerations reduced the amount of BWST water going into the RB sump significantly.

Reactor 8.5 7 7.6

'IN 1987 habitability evaluation report contained spray Building Sump solution pil Table 4.1 1, Results of Drawdown Analysis for a Additive / pil Minimum of 6,0 wt % Sodium Hydroxide in the Storage Tank. His table listed five RB spray cases with initial spray pil and time post LOCA for spray pli tp reach 8.5. De iodine removal constants were calculated using SRP 6.5.2 Rev.1.

BAW 20 44,"Elimiuti.m of Containment Spray Additive",

was a B & W study to actermine how to convert to from Na011 storage tank to TSP. With TSP, the initial RB spray pil will be around 4 5 because that is the pH of the BWST water. After the water mixes with the TSP in the RB flooded level and RB spray is swapped to recalculation, then the RB spray water pil increases to the range of 7 7.6. FPC installed the TSP baskets by MAR 88-05-01-01.

GCI revised Calculation 186-0002, Containment Spray Removal Constants (lodine Removal) to Rev. 2 and calculated the CR 3 specific iodine removal constants using SRP 6.5.2 Rev. 2 methodology in 1991.186-0002 Rev. 5,1/16/96, i

recalculated the total containment spray iodine removal 24

e l

REVISION 5 l

1 constants for 1397 (1500 gpm with largest maximum negative l

crror) and 1112 gpm (1200 gpm with largest maximum negative enor). Rese constants are considered to redect current plant d: sign and configu'tation, and are used in revised dose cales.

hillA Source Dased on TID Dased on TID ne higher power rating was incorporated based on recent Terms 14844 and a 14844 and a licensing activities regarding a CR #3 power uprate. His power level power level action has not been complettd, but the post accident source term associated with the hi,her power rating has been of2594 of2619 t

h1Wth htWth incorporated into dose calt.ulations. Since the source term is detennined based on a per megawatt basis per TID 14844, the use of the larger htWth rating results in a source term slightly higher than that which would be predicted with the lower power rating. His is r,learly a conservatism (not a USQ) given that the plant is still licensed to the lower value.

Auxiliary 0% efUcient 0% efUcient By letter dated September 13,1989 (3F0989-01), FPC Duilding in LOOP submitted a revised licensing basis for the CR 3 Loss of Filtration events,75%

Coolant Acciden'. (LOCA) and the hiakeup System Letdown efUcient in Line Failure Accident (LLFA) offsite radiological events for consequences to eliminate the credit for the Auxiliary Building which power Ventilation System (ADV) due to lack of safety grade power, is assumed to FPC re evaluated the offsite radiological consequences of a be LOCA using the same methodology for Dssion product release maintained.'

as that used to evaluate the CR 3 control room habitability in its June 30, 1987 habitability report ('iF058716), i.e. no credit for Auxiliary Huilding filters.

Dusing calculational verification efTorts relative to the Reactor Duilding (RD) Occiing issue, FPC identified that the control room habitability duie is adversely effected by the change in RD Deod volume. This affect was documented in FPC letter to the NRC dated June 4,1990 (3F0690-04). He habitability report postulates a gross failure of a passive component which causes a 50 gpm leak for 30 minutes at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . it was considered that since CR 3 does have a filtration system associated with the areas contaming the Engineered Sr.feguards (ESF) systems and passive failures such as that pastulated to cause the 50 gpm leak have not been considered

.s part of the

N3 licensing basis, the gross failure of a passive componcat would not be postulated in the CR-3 control room habitability dose analyses. Discussion with the NRC regarding the RD Cooding issue and the adverse effect on the control habitability dose resulted in the FPC analyses including the postulated gross failure of a passive component causing a 50 gpm leak for 30 minutes at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with the ADV system in senice with 75% cfucient charcoal filters for iodice removal (3F0690 06 and 3F069013).

The NRC 25

y f

' l REV1SiON $

l documented acceptance of this in its lesser to FPC dated June 21,1990 (3N0690 l$) as an interim measure until the RB Aoodmg issue was p,rly resolwd.

Subsequent to replacement of Sodium Hydroxide spray additive solution with Tf.P baskets, calculations were performed which i

==.:.d acceptable done sequences without the ABVS Alters and credit for their operation was discontinued.

In revised done analyses, the ABVS Akers as assumed to be operating for any event which assumes that the Auxahary Ihulduis is at a high negative pressure. Under those condnions, the ABVS supply fans are assumed to be tripped and the exhaust fans discharging thmugh the charcoal Altration systen and out the stack. Ddferential pressures across the CCHE on the order of 0.20" us would be expected, which would result in leakages considerably higher than that associated with MIMOOP.

Iknmer, given that the ABVS is assumed to be operating throughout the event, per SRP l$.6.5 no 50 gpm leak would be postulated at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months into the evat, and " normal" ECC3 leakge would be subject to Altration %us,this semano is bounded by the MHA/ LOOP with respect to Control Room Habnability, in the event of a MHA w /100P, no credit is taken for ABVS Altration for the duration of the 30 day accident period CREVS Flow 43,500 cfm 37,800 cfm 43,500 cfm is the original design flow rate of the CREVS, and Rate is the value used to determme IPF in the 6/30/87 habitability rcPort. Donc consequences were later evaluated at 43,$00 -

(recirc mode) 10% (39,150 cfm) corresponding to the allowable range of operation found in ITS Section $ 6.2.12. elative to the filter test program. During a system readiness resiew it was recognized that previous modifications had been made which reduced system flow rate without adequately assessing the effect on CRCVS. Revised done analyses incorporate a value of 37,800 cfm, based on consideration of current system capabilities under dirty filter conditions.

CREVS 95 %

95 %

Filtration efliciency hasn't changed, but filter testing has been Filtration upgraded to utilize more challenging criteria. Presious Efficiency charcoal testing was performed at 80 C at 30% RH, test program has been revised to evaluate charcoal at 30 C and 90% RH. Criteria for in place filter testing is penetration and system bypass of <0.05%.

h 26

' l a

f 5

l l

REVISION $

CCllh/CR 355,311 ft' /

364,922 ft' /

Original volumes were based on an internal muno from Volume 85,573 ft' 88,000 ff Gilbert. CCliE volume was estinated by calculating the volume of the entire envelope, then subtracting 10% for internal walls and contents. Updated volumes were calculated based on a room by room survey performed by S&L for use in Control Room heat up evaluations.

CREVS/

As described As modified The following modifications are being implemented to address

~

CCilE in the by R 12 the concerns associated with Restart issue R 12. Figures C.1 Modifications habitability modifications and C 2 provide a schematic of the pre and post modification report 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 logie is not changed.

Damrer AllD 99, which brings supply air to the Ventilation Equipment Room is being removed and a permanent blank installed. New supply and retum registers shll 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 ofinleakage.

Existing damper AllD 12, located in the supply duct to the CA, is to be removed end replaced with two new bubble tight dampers, Ai!D 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 AllDJE, shall be installed in series in the exhaust path. AllD 2C will be nornully closed.

The position of recalculation air damper AllD 3 will be established during the process of balancing 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 IC and AllD lE, are being installed in series on the inlet duct. Dampers AllD lC, AllD 2C and AllD 3 will retain positioners which provide a nunual 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 CCilE.

Mechanical Equipment Room Ventilation Air llandling Fans, AliF 21 A/B and associated dampers AllD 24. AllD 25, AllD 26, and AllD 27 are being spared in place and the associated CC11E penetration sealed.

This portion of the system originally exluusted air from the Mechanical Equipment Room, Elevator Equipment Room, lavatory, kitchen and 27

\\

fe. Y J

,e

'I u vision s 1-i toilet. His enumnases another potential source of f

intenkageinto the CCHE.

New supply and retum registers are being installed in j

e the doctwork in the Mechanical Equipment Room j

nis will provide ventilation to this portion of the l

CCHE during both normal and recalculation modes.

A skid mounted air handling unit consisting of a fan j

and a charcoal nitration unit will be installed to ventilate the' Elevator Equipment Room, lavatory, kitchen and toilet, nis system is non safety and non, j

scismic and will vent approximately 1,000 cfm by way 'of a Acid connection to a non safety related i

(NSR) duct, Small bore drain pipes penetrating the CCHE are j

e being Atted with-loop seals to prevent inlenkage j

though the lines. %ese will be added to a queued work request in the work controls system which nmintains CCHE drain line loop seals.

Vestibules have been installed over all CCHE l

boundary doors, and have bem sealed to provide maximum leak tightness. nose vestibules prmide a means to test individual CCHE boundary door leak tightness, as well as reducing inleakage associated l

p with CCHE access / egress.

In addition to the above modincations, an extensive effort was undertaken to survey CCHE penetrations and scal as required l

to minimize inleakage. As a result of thir work, it is l

concluded that conduit penetrations do not pose a significant liability to CCHE integnty, Penetrations associated with =

electrical cable banks were inspected and scaled to the extent j

feasibic with existing procedures and materials, but some leakage paths remain through the interstitial spaces between irvlividual cables. Additional work is being planned to improve the sealing of penetrations with the most significant

-l

leakage, j

CCHE Estimated on Measured by See detailed discussion pertaming to inleakage elsewhere in inloakage the basis of -

tracer gas the " JUSTIFICATION FOR CONTINUED OPERATION"

-l summation '

testing and section, this JCO.

leakage past analytically CCHE'.

corrected 19 boundary predict

{

elements per' inleakage

^

SRP 6,4 ~

under postulated post accident conditions -

I

~

28:

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ca. - --

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REVISION $

~

Does ICRP2 ICRP30 he NRC Safety Evaluation of FPC's control room habitability is based on the_*' Control Room Habitability l

Conversion Factors Evaluation Report" submitted to the NRC on June 30,1987.

t At that time, ICRP.2 methodology was used for internal dose calculations. Revised mahods for calculating organ done and relating organ doec to whole body done were published in ICRP.30, and endorsed for use in this country by the j

Environmental Protection Agency (EPA) in Federal Guidance Report #11. For the radionuclidos of concem, use ofICRP 30

/ Federal Guidance Report #11 dose conversion fadors results In the accident thyroid dose to be ~30% lower than previously calculated. CR.3 Improved Technical Speci6 cations (ITS) include specine activity limits f. primary and secondary l

coolant, which is measured and reported as DOSE i

EQUIVALENT l.131. ne ITS defmition of DOSE EQUIVALENT l.131 specines that the thyroid dose conversion factors used for this calculation shall be those from IChP 30.

SoRware Accident Analysis SoAware (POSTDBA)

Computer program POSTDBA is Sargent A: Lundy proprietary soAware which performs radiological dose calculations and related analyses for the LOCA in a PWR or a BWR. POSTDBA was originally developut to calculate PWR control room (CR) and offsite doses in accordance with l

requirements and recommendations of Regulatory Guide (RG) 1.4, Standard Review Plan -(SRP) Section 6.4, and SRP 6.5.2., and was revised and revalidated most recently in 1994.

i POSTDBA is constructed to allow the user to select the time steps and to control variable parameters for each time step.

De variables include containment spray iodine removal rates; post accident source release rates (iodide and noble gases) and i

any iodine filtration; x/Q changes; CR parameters (nukeup, inleakage, iodine removal, breathing rates, and occupancy factors); plus the fractions of clementai, 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 Grst time step can be used to model a g

delayed release his degree of user control allows other types of accidents to be analyzed.

+

Similar to POSTDBA, Computer program AXIDENT is NUS SCIENTECH proprietary soAware which performs

'dWogical dose calculations and related analyses.

e i

29

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. -.a

. a.

.a

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i l

ITVISION $

j APPENDIX A: ISOTOPES CURRENTLY !N CR 3 CORE JUSTIFICATION FOR CONTINUED OPERATION FOR Tile CONTROL ROOM EMERGENCY VENTILATION SYSTEM AND Tile CONTROL COMPLEX IIABITAHILITY ENVELOPE (For MODES 3,4,5, & 6 ONLY)

In response to a verbal Request for Additional Information, Revision 3 of this JCO, de.ted December 6,

1997 was transmitted to the U.S.

Nuclear Regulatory Commission (USNRC) by reference 20.

The USNRC's initial assessment of the JCo were transmitted to FPC by reference 18.

Finding number 5 of reference 18 states in part:

...the staff has concluded that you have not adoquately demonstrated that the radiation doses to control room operators will be less than the critoria of GDC-19 during a DBA. Since CR-3 has been shutdown for an extended period, the radioactivo inventory has decayed such that CCllE doses would not likely exceed GDC-19 criteria for any design basis accident that would occur prior to initial criticality following restart. However, operation at modo 2 or higher does not appear to be justified on the basis of information provided to date..."

Appendix A, to Revision 5 of this JC0 has been prepared to demonstrate that the radioactive inventory has decayed such that CCHE doses would not likely exceed GDC-19 criteria for any design basis accident that would occur prior to initial criticality following restart. Until NRC concerns are resolved, this JC0 will only apply to operation in MODES 3,

4, 5, and 6.

The following assessment is an excerpt from reference 19.

"The objective... is to assesc the reduction of the fission product isotopes in the vessel that has occurred while Crystal River J has been shutdown.

This assessment is being prepared to support operation of Crystal River 3 in modes 3 and 4 while the remaining habitability issues are being resolved (specifically, those issues identified in the NRC letter from Mr. L. Raghavan to Mr. R. Anderson dated December 24, 1997).

Sinca crysta) River 3 has been shutdown for an extended period (approximately 15 months), the radioactive isotope inventory that would be available for release following a loss-of-coolant accident with severe core damage while in modes 3 or 4 is significantly bolov those assumed in the design basis analyses.

Accordingly, the GDC-19 habitability criteria would not be challenged for any design basis accident while operating in thes Qsodes.

,The impact on the fraction of the isotopes remaining after a 15-month shutdown can be assessed as follows:

30

oc,

, V.*

l O

l e.

l REVISION 5 l

l i

i Fractkat = e*

Nuc114 Decay constant Decay Fraction Decays...to A

.f sec"1 liRR Remainina tem 1

I-131 9.97E-7 3.9E7 1.3E-17 Xe131 sec.

(stable)

I-132 8.37E-5 3.9E7

< I-131 Xe132 sec.

(stable)

I-133 9.17E-6 3.9E7-

< I-131 Xe133 i

sec.

(T1/2=5.2d)

I-134 2.22E-4

'3.9E7

< I-131 Xe134 i

sec.

(T1/2=53m) 1-135 2.87E-5 3,9E7

< I-131 Xe135 sec.

(T1/2=9.1h)

As can be seen from the above table, the fraction of the iodines currently in the core is negligible (as compared to the amount of the iodines that were present at the end of the most recent operating cycle).

The assessment also considered the precursor nuclides that will decay and generate additional iodines since power operation.

The precursor radioactive tellurium, available in the core when power generation was ceased,-will contribute to the generation of iodines after power generation.

However, the half-lives of the tellurium isotopes-are also fairly short and as such ample time is available for the additional lodines to decay to a negligible level.- Some of-the noble gases-with long half-lives, such as Kr-85, will still be available.

The noble gases.are not a concern to the habitability of the control Room since the whole body doses are well below the limits.

In summary, there are no control room habitability concerns

'with operation in modes 3 and 4 considering the extended outage duration..."

The same analysis would apply to the iodines in the spent fuel assemblies and in the reactor coolant. Thus, the Jc0 applies to Modes 3,

4, 5, and 6..

i t

p i

31t

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10 U.S. Nuclear Regulatory Commission i

310198 20 l

A'lTACllMl!NT 11 t

i i

CR 3 CCilli leakage data taken from calculation M97 0110 Rev 2 is shown below:

1 MIIA with LOOP f

Duration Wind Speed

'Unlittered I;iltered Total Inicakage inleakage inleakage (wind + stack + 10 (CRIIVS cfm)

Induced) j 0 to 8 lirs 2.7 mph 197 cfm 125 cfm 322 cfm 1 to 24 lirs 4.0 mph 212 cfm 125 cfm 337 cfm j

I to 4 Days 5.4 mph 233 cfm.

125 cfm 358 cfm t

4 to 30 Days 8.1 mph 284 cfm 1" cfm 409 cfm i

Thyroid 26.5 Ri!M lixposure w/22.8 sq. In.

j lireach Margin

M97-0137 Rev. 2. MIIA without LOOP uses a constant 523 cfm leakage.

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