3F0798-15, Application for Amend to License DPR-72,changing Improved TS for CREVS & Ventilation Filter Test Program.Control Room Habitability Rept,Ts Pages & Commitments Encl

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Application for Amend to License DPR-72,changing Improved TS for CREVS & Ventilation Filter Test Program.Control Room Habitability Rept,Ts Pages & Commitments Encl
ML20236V874
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 07/30/1998
From: Cowan J
FLORIDA POWER CORP.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236V876 List:
References
3F0798-15, 3F798-15, TAC-M91823, NUDOCS 9808040260
Download: ML20236V874 (18)
v • d • e


Text

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, orn.72 July 30,1998 3F0798-15 U.S. Nuclear Regulatory Commission I Attn: Document Control Desk  !

Washington, DC 20555-0001 i l

Subject:

License Amendment Request #222, Revision 1 Control Room Emergency Ventilation System and Ventilation Filter Test '

Program (TAC No, M91823)

References:

1. FPC to NRC letter, 3F0687-16, dated June 30,1987, " Control Room Habitability Evaluation Report" l
2. FPC to NRC letter, 3F0588-10, dated May 23,1988, " Control Room Habitability, Request for Additional Information (TAC No ti4805)"
3. NRC to FPC letter, 3N0589-25, dated May 25,1989, " Crystal River Unit 3 - Control Room Habitability Evaluation (NUREG-0737 Item lli.D.3.4) (TAC No. 64805)"
4. FPC to NRC letter, 3F0997-14, dated September 9,1997, " License Amendment Request 218; Revision of the Makeup System Letdown Line Failure Accident Analysis" j
5. FPC to NRC letter, 3F1297-19, dated December 5,1997, " License l Amendment Request #222, Revision 0, Control Room Emergency l Ventilation ar.r! Emergency Filters"
6. NRC to FPC letter, 3N1297-07, dated December 9,1997, " Crystal River Unit 3 (CR3)- Request for Additional Information - License Amendment Related to Revised Analysis of Makeup System Letdown Line Failure Accident (TAC No. M99571)"

\

Dear Sir:

Florida Power Corporation (FPC) hereby submits License Amendment Request (LAR) #222, l Revision 1, regarding proposed amendments to Operating License No. DPR-72 for Crystal 3' River Unit 3 (CR-3). The attached LAR proposes changes to the Improved Technical V ; '

Specifications (ITS) for the Control Room Emergency Ventilation System (CREVS) and to the D Ventilation Filter Test Program (VFTP). This submittal supersedes LAR #222, Revision 0 (Reference 5), in its entirety. NRC approval of this LAR is requested by July 19,1999.

9000040260 980730 '

PDR ADOCK 05000302 P PDR CRYSTAL RIVER ENERGY COMPtNX: 15760 W. Power une Street

  • Crys'.al River, Florida 344284708 (362)796 4486 )

A FlorMa Progress Company j

U.S. Nuclear Regulatory Commission 3F0798-15 Page 2 of 3 This LAR includes the changes to the ITS previously submitted in LAR #222, Revision 0, and proposes additional changes. The description of the changes is provided in Attachment A. A revised Control Room Habitability Report, provided in Attachment B, supersedes the control room habitability information submitted in References I and 2. It is anticipated that the NRC will issue a new Safety Evaluation on Control Room liaiAtability for CR-3, based on this submittal, that will replace the current Safety Evaluation issued via Reference 3. The proposed changes to the ITS pages in redline / strikeout f wmat are provided in Attachment C.

The proposed changes in revision bar format are provide 1 in Attachment D.

In Reference 4 FPC submitted LAR #218 regarding revision of the makeup system letdown line failure accident analysis. In Reference 6, the NRC staff presented the results of a confirmatory analysis for post-accident control room aperator dose for the change proposed in LAR #218. The staff analysis indicated that doses co? I exceed 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19 requirements. The NRC requested that FPC submit an analysis demonstrating that CR-3 satisfies GDC 19 requirements. FPC's analysis is presented in Attachment B. The analysis demonstrates that CR-3 meets the GDC 19 requirements for this accident, and that doses from the letdown line failure are bounded by the postulated ma:imum hypothetical accident (MHA). With the submittal of the results of the letdown line failure accident dose analysis, FPC anticipates that LAR #218 can be approved.

Commitments made in this submittal are identified in Attachment E. If you have any questions concerning the information provided in this submittal, please contact Ms. Sherry Bernhoft, Manager, Nuclear Licensing at (352) 563-4566.

Sincerely, M C-- 3 John Paul Cowan l i

Vice President Nuclear Operations JPC/scp l

xc: Regional Administrator, Region 11 j NRR Project Manager i Senior Resident inspector Attachments:

A. License Amendment Request (LAR) #222, Revision 1, Control Room Emergency Ventilation System and Ventilation Filter Test Program B. Control Room Habitability Report C. Improved Technical Specificat%n Pages in Redline / Strikeout Formac D. Improved Technical Specification Pages in Revision Bar Format E. Commitments l

[. . - _ _ _ _ - _ _ _ _ _ _ - _ . - - _ A

l U.S. Nuclear Regulatory Commission  !

3F0798-15 i Page 3 of 3 l

l L

l STATE OF FLORIDA i

COUNTY OF CITRUS i

)

John Paui Cowan states that he is the Vice President, Nuclear Operations for Florida Power Corporation; that he is authorized on the part of said con.pany to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements J

made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.

I As John Paul Cowan Vice President Nuclear Operations L

Sworn to and subscribed before me thisM day of Oude ,1998, by John Paul Cowan.

0 /5G d b h e / $ / h ( 1 Signature of Notary Public

. State of Florida w - _ -_ _ _-_ ;_--- _ w _

(Print, type, or stamp Commissioned Name of Notary Public)

Personally Produced Known -OR- Identification l

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACilMENT A DESCRIPTION OF CIIANGES LICENSE AMENDMENT REQUEST (LAR) #222, REVISION I CONTROL ROOM EMERGENCY VENTILATION SYSTEM i AND VENTILATION FILTER TEST PROGRAM l

I 1

U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 1 of 13 DESCRIPTION OF CIIANGES l

LICENSE AMENDMENT REQUEST (LAR) #222, REVISION 1 CONTROL ROOM EMERGENCY VENTILATION SYSTEM AND VENTILATION FILTER TEST PROGRAM i LICENSING INFORMATION This submittal describes Florida Power Corporation's (FPC) actions to increase the level of protection provided to control room personnel, and to demonstrate the effectiveness of the j protective features through radiological dose and toxic gas exposure calculations. FPC has revised the Crystal River Unit 3 (CR-3) Safety Analysis Report (SAR) to incorporate the physical modiDeations to the Control Room Emergency Ventilation System (CREVS).

However, the SAR descriptions of the licensing basis for control room habitability and the accident dose analyses have not been revised to incorporate the analytical results presented in this submittal. This is considered appropriate until the unreviewed safety questions regarding  ;

the licensing basis for control room leakage and post-accident dose analysis are resolved and this LAR is approved. Accordingly, the NRC staff should be aware of the differences l between the information in the SAR and in this submittal during its review.

Attachment B describes control room dose calculations for a steam generator tube rupture (SGTR) accident and a letdown line failure accident. Both analyses use assumptions that are outside the CR-3 licensing basis. The predicted control room doses from these accident analyses are bounded by the maximum hypothetical accident analysis (MHA) results. The SGTR and letdown line failure analyses are provided for information so that the NRC staff can evaluate the protection provided by the CR-3 control room design in comparison to other designs. FPC is not planning to change its licensing basis to adopt the assumptions used in these two analyses.

The SGTR analysis was performed in accordance with the requirements of Standard Review Plan (SRP) 15.6.3, " Radiological Consequences of Steam Generator Tube Failure (PWR),"

with minor deviations that are explained in Attachment B. CR-3 is not an SRP plant and has not r.; de a commitment to incorporate SRP 15.6.3 into its licensing basis. This analysis is provioed for information only and does not represent a change to the CR-3 licensing basis for SGTR accident analyses. However, as noted in Section 1 of Attachment B, FPC will maintain this SGTR analysis current with changes that may affect the control room dose results. {

Both the letdown line failure and steam generator tube rupture (SGTR) analyses have been performed using an assumed pre-existing iodine spike of 60 pCi/gm in the Reactor Coolant System (RCS). Additionally, sensitivity studies determined that the pre-existing spike scenarios bounded dose calculations for an iodine spike concurrent with the accident. The assumption of a pre-existing or concurrent iodine spike in accordance with SRPs is not a part of the CR-3 licensing tasis. CR-3 Improved Technical SpeciDeations (ITS) Limiting Condition for Operation (LCO) 3.4.15, "RCS Specific Activity," establishes the dose equivalent 1-131 limit for operation as I pCi/gm. 60 Ci/gm is the Action Statement limit L____----_--______----_------____---__

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U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 2 of 13

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l when the LCO is exceeded and opera 4 i may continue for only forty-eight hours above the l LCO limit.

l The ITS Bases 3.4.15 " Applicable Safety Analyses," states that RCS specific activity satisfies Criterion 2 of the NRC Policy Statement. The NRC Policy Statement on Technical SpeciGcation Improvements defines the appropriate contents of ITS LCOs meeting Criterion 2 as follows:

"A process variable, design feature, or operating restriction that is an initial condition of a Design Basis Accident or Transient analysis that either assumes the failure of or represents a challenge to the integrity of a fission product barrier."

The LCO limit of 1 pCi/gm is the process variable or operating restriction that is an initial condition for accident analyses in accordancc with the Policy Statement. However, the Bases,

" Applicable Safety Analyses" also states that:

"The LCO limits on the specific activity of the reactor coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the site boundary w.ill not exceed a small fraction of the 10 CFR 100 dose guideline limits following an SGTR accident. These values represent a reasonable operating capability rather than a specific analytical result. The CR-3 specific SGTR safety analysis assumes the specific activity of the reactor coolant is representative of 1% defective fuel and a primary to secondary leak rate of I gpm through the steam generators."

The reactor coolant specific activity representative of 1% defective fuel (approximately 7 i Ci/gm) is conservatively assumed in accident analyses in the CR-3 SAR. Reactor coolant specific activity representative of 1% defective fuel is the licensing basis for CR-3 accident analyses. CR-3 is not changing the licensing basis for accident or transient analyses with the submittal of the Control Room Habitability Report, Attachment B, which describes control room dose analyses that assume pre-existing RCS dose equivalent iodine concentrations of 60 pCi/gm. These analyses are presented for information and demonstrate that even with the additional conservatism of SRP assumptions and iodine spiking the results are bounded by the MHA analysis.

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1 U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 3 of 13 LICENSE DOCUMENT INVOLVED: Technical Specifications PORTIONS: Specification 3.7.12 Specification B3.7.12 i

Specification 5.6.2.12

SUMMARY

OF CHANGES:

Specifications 3.7.12 and B3.7.12, Control Room Emergency Ventilation System (CREVS)

This amendment request proposes the following changes:

1. A new Required Action for the existence of breaches in the Control Complex Habitability i Envelope (CCHE) that are in excess of allowances in the approved control room dose l calculations. A single breach or multiple breaches, that in combination, are less than or l_ equal to one square foot in excess of normal breach allowances would be permitted for up I to seven days.
2. The Note that modifies Required Action D.1 is being moved to be applicable to Required  ;

Actions D.1 and D.2. This is an editorial change to correct a format error in the existing j specification. Appropriate bases changes are made to conform with the revised Note l structure.

3. A new Surveillance Requirement for the performance of a periodic integrated leak test of the CCHE boundary on a twenty-four month frequency. l l
4. Bases changes which describe the treatment of CCHE breaches that are within the limits of l l the approved control room dose calculations. Proposed Bases changes establish that the i CCHE is inoperable if there are breaches in the CCHE in excess of those allowed within J the approved calculations. The proposed change further establishes that CCHE integrity is required for CREVS operability.
5. Bases changes which revise the CREVS description to include the modifications made recently, and to correct an error, in that a CREVS return fan is necessary for system  ;

operability. Other editorial changes are made for clarity. l l

6. Discussions of the proposed new Required Action and new Surveillance Requirement are also added to the Bases. The Bases discussion of the proposed breach allowance describes application of the allowance to leakage measured during a CCHE leak test. If measured leakage is less than the leakage through the maximum allowed breaches at CCHE test conditions, then operation may continue for up to seven days.

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- l U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 4 of 13 Specification 5.6.2.12, Ventilation Filter Test Program (VFTP)

This amendment request proposes to change the VFTP as follows: i 1

1

.l. Adopt current standards for laboratory carbon adsorber testing and change the test j

conditions to beter match the operating conditions inside the Control Complex (CC) following the postulated design basis radiological accident. FPC proposes to perform laboratory tests of carbon adsorber in accordance with ASTM D 3803-1989 at a temperature of 30"C and relative humidity (Rll) of c5%. The laboratory test acceptance i criterion is also being changed to less than 2.5 % me.hyl iodide penetration. I

2. Change the acceptable values for CREVS flow rate and filter differential pressure. A new lower value of Dow rate is being established which matches the control room dose I

! calculation, and a lower allowed filter differential pressure that corresponds to the new Dow rate.

l 3. Editorial changes are being made for correction and clariGcation. i 1 \

l 4. The Auxiliary Building Ventilation Exhaust Filters (ABVEFs) are being added to the 1 l VFTP. Laboratory tests of carbon adsorber will be in accordance with ASTM D 3803-i 1989 at a temperature of 30"C and relative humidity (RH) of 95%. The laboratory test acceptance criterion is proposed to be methyl iodide penetration of less than 12.5%. Other

tests and acceptance criteria are unchanged from requirements for the ABVEFs that existed previously in CR-3 Standard Technical Specifications prior to conversion to improved Technical Specifications in 1993. l l

REASON FOR REQUEST:

l Specifications 3.7.12 and B3.7.12, Control Room Emergency Ventilation System l (CREVS)

The CREVS and the CC11E act together to provide an enclosed environment from which the L J. ant can be operated following an uncontrolled release of radioactivity or toxic gas. The design of the CCHE for radiation protection is based on the radioactivity release associated with a design basis loss of coolant accident (LOCA). The source term used in the design basis

! LOCA analysis is consistent with Regulatory Guide (RG) 1.4, Revision 2 June 1974. The l limiting event for the CCliE due to a toxic gas release is a catastrophic rupture of the sulfur dioxide or chlorine storage tanks at the adjacent coal fired power generation units.

I The current CREVS Technical Specification (TS) addresses the components of the ventilation system loop including fans, dampers, Glters, and associated ductwork. The TS is si!cnt on the components that make up the CCilE boundary including the walls, doors, roof, floors, and floor drains of the CC, and does not explicitly address the isolation dampers. The Bases l change revises the system condguration and operating description, and establishes a clear connection between CCilE integrity and CREVS operability.

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  • l U.S. Nuclear Regubtory Commission Attachnient A 3F0798-15 Page 5 of 13 Calculations have determined the dose to control room operators in the thirty day period following an MHA due to radioactivity leaking into the habitability envelope (HE) will be below General Design Criterion (GDC) 19 limits. Calculations have also been performed to determine how much additional leakage can be tolerated while assuring doses will not exceed the regulatory limit. The additional leakage that can be tolerated is convened into an equivalent breach size that can be allowed in the HE to accommodate wear and tear on doors I and maintenancehnodification activities on HE boundary components. Based on the most l recent leak test, 35.5 square inches of breaches are allowed in the HE boundary. This l allowable breach size may increase or decrease when a leak test of the HE is performed and  ;

finds less or more leakage than the last test. Breaches are authorized and tracked via CR-3 '

administrative control procedures to assure the breach limit is not exceeded and operability is maintained in day-to-day operation. l The proposed Condition B and Required Action B.! sanction additional breaches in the CCHE in excess of those discussed above. Condition B will be entered when breaches totaling greater than 35.5 square inches but less than 179.5 square inches exist. Operation mr continue for up to seven days in this condition. This allowance would permit additio !!

breaches to facilitate maintenance / modification activities, or to accommodate door or seal failures thn may occur. Bases additions describe the application of the one square foot breach  ;

margin to CCHE leak test results. Action B.1 would apply if a CCHE leak test measured total leakage that is greater than leakage through 35.5 square inches of breach but less than or equal to leakage through 179.5 square inches of breach at HE test conditions. This would permit testing of the HE to be performed on-line, prior to an outage, without the threat of immediate shutdown. Testing experience has shown that we can conduct three integrated leak tests of the HE in one week. The proposed provision would permit us to test, evaluate results, inspect / seal, and test again within 'he a.llowed seven day period while operating. These provisions are important to FPC to facilitate work management and to meet goals for minimum duration outages.

The proposed Surveillance Requirement to perform periodic leak tests will determine the leakage across the habitability boundary, and will allow FPC personnel to verify that the integrity of the CCHE is being maintained within the design basis.

The CREVS return fan is being added to the LCO section of the Bases to clearly establish CREVS operability requirements.

Specification 5.6.2.12, Ventilation Filter Test Program (VFTP) l The change in carbon adsorber laboratory test standard will provida a more conservative test than that to which FPC is currently committed. The new test condit!ons for the CC and Auxiliary Building (AB) filters are more representative of the operating conditions to which the carbon adsorber material will be exposed when in service. This change is conservative since testing at a higher temperature may overestimate adsorption capability at the expected filter media operating temperature.

l L_______

1 U.S. Nuclear Regulatory Commission Attachment A j 3F0798-15 Page 6 of 13 FPC has made major changea to improve control room habitability and CREVS performance. j New control room radiological dose calculations have been performed. Lower, more i conservative CREVS How rates were used in the dose calculations. Flow rates were chosen which reHect the reduced flow associated with Elter fouling. The values of allowed Glter differential pressure and lower flow rate limit are being changed for the CREVS Glter tests to be consistent with the calculations.

I Control room habitability dose calculations include scenarios that postulate both a loss of.

offsite power (LOOP) and no LOOP in combination with radiological release events. For scenarios where there is no LOOP, the Auxiliary Building Ventilation Exhaust-System and Filters will remain functional and are credited in control room dose analyses. Therefore, the .

ABVEFs are being added to the VFTP.

l JUSTIFICATION FOR REQUEST:

l Specifications 3.7.12 and B3.7.12, Control Room Emergency Ventilation System I (CREVS)

Required Action for CCHE Breaches CR-3 has administrative controls established in Compliance Procedure, CP-147, " Control Complex Habitability Envelope'(CCHE) Breaches" for authorizing and tracking breaches.

This procedure requires that any planned or discovered breach be evaluated agamst the remaining breach margin available to determine continued integrity and operability of the HE.

FPC docketed a commitment in FPC to NRC letter, 3F0597-11, dated May 15, 1997,

" Control Complex Habitability Envelope," that if breaches are discovered in the HE which exceed the margin provided in approved dose calculations, that both trains of CREVS would l be declared inoperable and Required Actions D or E of ITS 3.7.12 will be entered as applicable.

CP-147 has been in effect at CR-3 since November 20, 1995, and has been effective in l controlling HE breaches. The breach margin that has been available has been between 22 and )

43 square inches. This represents an extremely tight restriction on the types of maintenance i and modification activities that can be accomplished during applicable plant modes. l Therefore, the proposal for an allowance of less than or equal to one additional . square foot of I breach margin, or its equivalent in leakage, for up to seven days would provide significantly l increased flexibility in planning and scheduling work activities. l This provision is similar to one approved for the Waterford-3 Technical Specifications by License Amendment No.115, issued October 4,1995.

l The safety considerations for a breach allowance as proposed are:

1. An effective control process exists that identifies the location and size of open breaches. In the event of an accident that poses a threat to control room habitability, breaches can be

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U.S. Nuclear Regulatory Commission Attachment A r 3F0798-15 Page 7 of 13 l

readily closed. The relatively small size of any single breach would facilitate closure using materials that are readily available. This is significant since operator dose is calculated for j' the thirty day period following an accident. The time required to close breaches, if l determined necessary by observation of actual post-accident conditions, would be a small

fraction of that time.

l I 2. Conservatism in dose calculations which postulate massive core damage, source release that assumes continuous high containment pressures, and unfavorable atmospheric transport provide significant margin above realistically anticipated exposures. Core reload i analyses that demonstrate 10 CFR 50.4(> compliance validate the effectiveness of Emergency Core Cooling Systems (ECCS) and the ability of the reactor core to withstand

i. accident conditions without extensive damage. Realistically anticipated core releases from an event are the equivalent of 1% failed fuel versus a RG 1.4 core release of 25% of the equilibrium iodine and 100% of the noble gases contained in the core.
3. Recent NRC sponsored research on release of radioactive material from a damaged core, NUREG-1465, indicates much lower releases than are currently approved for use in dose calculations (RG 1.4). Lower releases would translate directly into lower calculated doses and lower actual doses, thereby minimizing the threat from additional breaches or leakage into the HE.
4. No threat to operators from toxic gas is created by the proposed breach allowance due to the reduction in the amount of toxic gas on site, new analyses of the remaining toxic gas sources, and the insensitivity of toxic gas concentrations in the control room to breaches in the HE. Limiting toxic gas exposures come from " puff" releases that result in a highly concentrated cloud of gas entering the CC through the normal ventilation intake. Since this type of release is quickly transported beyond the CCHE, the existence of breaches would not contribute significantly to toxic gas concentrations in the CCllE.

Periodic Leak Test Surveillance Requirement This Surveillance Requirement is similar to that included in Babcock & Wilcox Standard Technical Specifications, however, a change to the standard requirement is necessary. The standard surveillimce test acceptance crit'eria is expressed as a specific amount of leakage (or makeup air) when the control room is pressurized at a nominal 1/8 inch wg. The CR-3 CREVS is not designed to pressurize the CCHE, therefore an alternate test is required.

\. l The CCHE upgrade accomplished during the recent design improvement outage included a leak test of the CCHE. The test results have been evaluated in relation to control room  ;

habitabinty, and found to be acceptable. The proposed Surveillance Requirement description l in the ITS Bases provides the basic requirements for the leak test. The Bases references the CR-3 Control Room Habitability Report, dated July 1998, which is being docketed with this

, submittal, for a detailed description of the leak test conditions and acceptance criteria. The addition of a requirement to perform a periodic leak t9st is conservative. The adoption of this i

surveillance requirement will provide assurance of the continued integrity of the CCHE boundary, i

U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 8 of 13 l

l Bases Changes l

The Babcock & Wilcox Standard Technical Specifications Bases include a statement in the LCO section that requires the integrity of the control room habitability boundary be l maintained within the assumptions of the design analysis. FPC is proposing to include the l same requirement in the CR-3 ITS Bases LCO section. Additional discussion of the treatment l of CCllE breaches within the limits of the approved design calculations, and Bases descriptions of the proposed breach allowance are provided for clarification and to assure l consistent interpretation of this LCO. The addition of a return fan to the list of CREVS l equipment necessary for operability corrects an error in ITS. Other Bases changes are l provided to update the CREVS functional description reflecting improvements in the system design.

l These changes are justified in that they provide better information to the users of the ITS on the basis for CREVS operability.

Specification 5.6.2.12, Ventilation Filter Test Program (VFTP)

Test Standard and Acceptance Criteria Changes The existing VFTP requires laboratory testing of carbon adsorber samples in accordance with Regulatory Guide (RG) 1.52, Revision 2,1978 and ASME N509-1976. These documents ultimately require that the test be performed to USAEC Division of Reactor Developrnent and Technology Standard RDT M-16-lT. Performance of laboratory tests in accordance with ASTM D 3803-1989 eliminates technical problems that have been identified with the RDT M-16-lT test method.

The RDT M-16-lT test standard requires the carbon to be equilibrated by sweeping air at 25"C and 70% Rii through the test carbon. The methyl iodide test medium is then instantaneously introduced at 80"C and 70% RH. Testing carbon with such thermal step changes is technically l incorrect because it causes condensation on the samples. Condensation on the carbon makes the test invalid. To correct this problem ASTM D 3803-1989 includes a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> pre-test thermal stabilization at 30"C and specifies a temperature of 30"C for all phases of the test.

Therefore, ASTM D 3803-1989 is a better test because it solves the problem of the formation i of condensation on the carbon sample.

The current test conditions specified in the VFTP are temperature of 80"C and 70% Ril. The proposed test conditions of 30"C and 95% RH are more representative of the conditions to which the CR-3 carbon adsorber would be exposed following a radiological accident. l Information Notice 86-76, ' Problems Noted in Control Room Emergency Ventilation Systems," indicated that laboratory testing of carbon at a temperature higher than that expected during the course of an accident could result in a significant over-prediction of the i i

f capability of the carbon to remove methyl iodide. Therefore, the proposed test conditions will l give a more accurate prediction of the performance of the carbon under post-accident conditions.

U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 9 of 13 l

Control Room Emergency Ventilation System Filters The ASThi D 3803-1989 standard is more stringent than the RDT hi-16-lT standard since it has closer tolerances on temperature, relative humidity, and time. These result in better reproducibility of the test results. Testing in accordance with ASThi D 3803-1989 at 30"C and 95% RH with a 2.5% penetration limit is more conservative than testing to RDT hi-16-lT at 80"C and 70% RH with a penetration limit of 1% due to the more stringent tolerances on temperature and humidity in ASThi D3803-1989. The combined effects of testing at more representative conditions and the use of the improved test standard compensates for the increased penetration limit.

The existing limits on ventilation flow rates are 43,500 cfm 10%. The proposed revision to the VFTP will not change the upper flow rate (43,500 + 4,350 = 47,850 cfm). Calculations have verified that filter performance will be maintained at this flow rate. The lower limit on flow in the proposed revision matches the assumptions in the control room dose calculations.

The assumption of lower flow rates results in increasing the dose calcul:ded for a given accident scenario. The new lower limit on flow of 37,800 cfm was chosen based on actual measured flow rates, and the calculated reduction in flow due to tilter fouling and increased differential pressure across the filters. A corresponding change to the maximum allowed filter differential pressure from six inches to four inches is consistent with the flow and dose calculations.

Auxiliary Building Ventilation Exhaust Filters l

The ABVEFs were included in the CR-3 Standard Technical Specificauons (STS), and were '

removed in the conversion to ITS in 1993. The basis for removal was that the filters were not credited as part of any accident analysis. The analyses presented in the Control Room Habitability Report included in this submittal take credit for the ABVEFs at 75% iodine removal efficiency when the accident scenario does not include a LOOP. Therefore, the ABVEFs are being added to the ITS VFTP. ,

l The acceptance criteria for high efficiency particulate air filter penetration and in-place carbon adsorber system bypass are consistent with the former STS requirements. The acceptance criterion for laboratory testing of carbon adsorber samples is being changed to 12.5%

penetration to be consistent with the filter efficiency credited in control room dose calculations. In accordance with guidance provided in draft Generic Letter 97-XX,

" Laboratory Testing of Nuclear-Grade Activated Charcoal," a factor of safety of 2 has been applied to the credited 75% removal efficiency to determine the laboratory sample acceptance criterion, No LCO is being proposed for the Auxiliary Building Ventilation Exhaust Fans. These fans are normally operating 100% of the time to maintain environmental conditions within the AB, l and to maintain a slight negative pressure in the AB to control the release of radioactivity to l the environment. Normal operating and maintenance practices assure one train of AB exhaust fans are available and in operation. Dose analyses of the hillA and letdown line failure

U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 10 of 13 l

accident were performed for both LOOP and non-LOOP scenarios. The effects of the unavailability of the AB exhaust fans, and therefore the filters, for the non-LOOP scenarios l are bounded by the existing calculations. In general, unavailability of the AB fans results in l lower leakage into the CCHE and results in lower calculated dose.

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U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 11 of 13 NO SIGNIFICANT HAZARDS CONSIDERATION: l 1

An evaluation of the proposed license amendment has been performed in accordance with 10 CFR 50.91(a)(1) regarding significant hazards consideration, using the standards on 10 CFR 50.92(c).

1. . Does not involve a sigmficant increase in the probability or consequences of an accident previously evaluated.

Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. The Control Room Emergency Ventilation System (CREVS) and the Contro! Complex Habitability Envelope (CCHE) are designed to limit the radiation dose to the control room operating staff following a design basis accident. Since these systems are only effective in limiting dose following an accident, the existence of limited breaches in the CCHE, the performance of periodic leak tests, and changes to the Ventilation Filter. Test Program (VFTP) would not increase the probability of occurrence of any evaluated event. The features of the CREVS and the Control Complex emergency Glters, or the CCHE have no direct function in mitigating the offsite consequences of any evaluated accident. The Auxiliary Building exhaust filters are not credited with reducing offsite doses, however, if available would filter releases from the Auxiliary Building. Adding them to the VFTP will not increase the consequences calculated for any evaluated accident.

The proposed changes are consistent with the revised control room operator dose calculations as presented in the Control Room Habitability Report dated July 1998.

Since all calculated doses are within 10 CFR Part 50, Appendix A GDC 19 limits there is no signiHeant increase in consequences.

It is conceivable that the existence of additional breaches in the CCHE could result in an increase in operator dose, however the low probability of a catastrophic reactor accident, the relatively short time allowed for breaches to be open in excess of approved dose calculation assumptions, and the ability to close breaches expeditiously makes the risk increase insigniHeant.

The changes to the ITS Bases improve information on the operation and function of  !

l CREVS, and establish that CREVS operability is dependent on maintaining CCHE l l integrity. The inclusion of this information reinforces the importance of maintaining l

! the CCHE boundary, and will help to ensure the CREVS is capable of performing its i intended safety function.

The Control Room Habitability Report, dated July 1998, provided with this LAR presents the methodology used and the results of the operator dose calculations for the Maximum Hypothetical Accident, toxic gas release, and other design basis accidents.

The report provides the information needed for NRC review of LAR 222, Revision 1 and the associated unreviewed safety question. This evaluation concludes that the

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  • i U.S. Nuclear Regulatory Commission Attachment A 3F0798-15 Page 12 of 13 l

current level of CCHE integrity provides adequate protection for the control room operator.

Based on the foregoing, the proposed amendment does not signincantly increase the probability or consequence of an accident previously evaluated.

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! 2. Does not create the possibility of a new or different kind of accidentfrom any accident previously evaluated.

l Neither performance of periodic CCHE leak tests nor changes to the existing VFTP can create the possibility of a new or different kind of accident. During the period of time when CCHE breaches are greater than the design calculation, there exists the possibility that control room dose from an analyzed accident may be greater than l specified in General Design Criterion 19. This condition will not however create the possibility of a new or different kind of accident. Since CREVS and the emergency filtration units function to provide protection following a radiological accident the changes proposed to improve their performance cannot create a new or different kind of accident. Changes to the Bases to provide better information on determining l CREVS and CCHE operability cannot create the possibility of a new or different kind of accident.

3. Does not involve a sigmficant reduction in a margin ofsafety.

The proposed amendment does not involve a significant reduction in a margin of safety. Neither performance of periodic CCHE leak tests nor changes to the existing VFTP can create a reduction in the margin of safety. The changes to both of these l programs will result in improved assurance that the CREVS and CCHE will perform as expected if required for operator protection. Changes to the Bases of the CREVS I

Technical Specification which clarify the conditions necessary for operability will improve understanding of the requirements for maintaining control room habitability, and will not create a reduction in the margin of safety. The existence of additional breaches in the CCHE for short periods of time does not significantly increase the risk

of control room operator exposure to airborne radioactivity or toxic gas. There is no l change in the risk to the public since the CCHE has no direct function in mitigating the l offsite consequences of any evaluated accident. Any event that could create these
exposures has an extremely low probability of occurrence, and while the potential for  ;

higher operator exposure exists if additional breaches are open, the short duration l allowed would not signincantly increase the risk of exposure. Therefore, for the reason stated above the existing margin of safety would not be reduced.

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j U.S. Nuclear Regulatory Commission Attachment A l 3F0798-15 Page 13 of 13 l ENVIRONMENTAL IMPACT EVALUATION 1

Radiological Evaluation l

While 10 CFR 51 requires an environmental assessment (EA) or environmental impact statement (EIS) for any " major Federal action significantly affecting the quality of the human environment," it does allow the NRC discretion in evaluating the extent to which EA's or EIS's are necessary. EA's or EIS's are not required for any action included in the list of

" categorical exclusions" set forth in 10 CFR 51.22(c). SpeciHeally,10 CFR 51.22(c)(9),

provides that an EA is not required for the issuance of an amendment provided that:

(i) the amendment involves no significant hazards consideration, (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

FPC considers that the provisions of 10 CFR 51.22(c)(9) are applicable to this request for these changes to the Improved Technical Specifications for the CREVS and CCHE, and for the Ventilation Filter Test Program. For the reasons described in this submittal, FPC believes that the three criteria of 10 CFR 51.22(c)(9) are satisfied. Therefore, this License Amendment should be considered under the " categorical exclusions" provisions of 10 CFR 51.22(c)(9). There will be no environmental impact from allowing breaches to exist in the j CCHE, froni performing periodic leak tests of the CCHE, or from changes to the VFTP. For the reasons given in this submittal that there will be no change in offsite consequences due to this action, its impact is bounded by the impacts assumed in the existing Final Environmental Statement (FES) for CR-3. Even if the NRC chooses to perform an EA, information provided l in the FES, together with this submittal should assist the NRC in making a " finding of no signiGcant impact" in accordance with 10 CFR 51.32.

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i FLORIDA POWER CORPORATION CRYSTAL RIVER UNTI' 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ,

i ATTACIIMENT B CONTROL ROOM IIABITABILITY REPORT I

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