ML043480166

Addendum to License Amendment Request Dated March 1, 2004

ML043480166
Person / Time
Site:	Ginna
Issue date:	12/06/2004
From:	Widay J Constellation Energy Group
To:	Clark R Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML043480166 (43)

Text

i.A Joseph A. Widay 1503 Lake Road Plant Manager Ontario, NewYork 14519-9364 585.771.3000 Constellation Energy R.E. Ginna Nuclear Power Plant, LLC December 6, 2004 Mr. Robert L. Clark Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

Subject:

Addendum to License Amendment Request dated March 1, 2004 R.E. Ginna Nuclear Power Plant Docket No. 50-244

References:

1. Letter from Robert C. Mecredy (RG&E) to Robert L. Clark (NRC) dated May 21, 2003, License Amendment Request Regarding Revision of Ginna Technical Specification Sections 1.1, 3.3.6, 3.4.16, 3.6.6, 3.7.9, 5.5.10, 5.5.16, and 5.6.7 Resulting From Modification of the Control Room Emergency Air Treatment System and Change in Dose Calculation Methodology to Alternate Source Term.

2. Letter from Robert C. Mecredy (RG&E) to.Robert L. Clark (NRC) dated September 30, 2003, Summary of Public Meeting Between RG&E and NRC Staff Held on August 19, 2003.

3. Letter from Robert C. Mecredy (RG&E) to Robert L. Clark (NRC) dated March 1, 2004, Addendum to License Amendment Request submitted May 21, 2003.

Dear Mr. Clark:

On May 21, 2003 Ginna submitted Reference 1 related to the Control Room Emergency Air Treatment System (CREATS) modification and conversion to Alternate Source Term (AST).

Included in that submittal were proposed Technical Specification (TS) amendments for the new CREATS and related issues. Subsequent to that submittal, Ginna committed to resubmit the appropriate TS sections (Reference 2) upon approval of Tech Spec Task Force (TSTF)-448.

After a significant period of time, it became apparent that TSTF-448 approval was not forthcoming. Per phone conversations with your staff, Ginna resubmitted the appropriate TS sections (Reference 3) per the guidance in Regulatory Guide (RG) 1.196, Appendix B.

Subsequent to that submittal, Ginna and the NRC Staff discussed the submittal of a License Amendment which did not reference RG 1.196. Attachment 1 to this letter contains a revised submittal reflecting those discussions, and should be considered an addendum to Reference 3.

YdO 03

Very truly yours seph AW ay STATE OF NEW YORK  :

TO WIT:

COUNTY OF WAYNE I, Joseph A. Widay, being duly sworn, state that I am Acting Vice President - R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this response on behalf of Ginna LLC. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me, a Diptary Publi in and for the State of New York and County of 17oJ1,oe , this _ day of eceb , 2004.

WITNESS my Hand and Notarial Seal: t 4? fig Notary Public My Commission Expires:

MICHALENE A BUNTS Y) I // aof tNotary Public, State of NewYork U Dat6 Registration No. 01BU6018576 Monroe County Commission Expires Jan o2 7 6i Attachments:

1. Revised Submittal -Tech Spec Section 3.7.9, 5.5.10, and Removal of Sections 5.5.16 and 5.6.7 Previously Proposed in Reference 3.

Cc: Mr. Robert L. Clark (Mail Stop 0-8-C2)

Project Directorate I Division of Licensing Project Management Office of Nuclear Regulatory Regulation U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S. NRC Ginna Senior Resident Inspector Mr. Peter R. Smith New York State Energy, Research, and Development Authority 17 Columbia Circle Albany, NY 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire State Plaza, 10 th Floor Albany, NY 12223 James M. Petro Jr.

Counsel, Generation - Nuclear Constellation Energy 750 East Pratt Street, 17th Floor Baltimore, MD 21202

Attachment I Revised Submittal Tech Spec Section 3.7.9, 5.5.10, and Removal of Sections 5.5.16 and 5.6.7 Previously Proposed in Reference 3.

Marked-up Sections i

CREATS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Control Room Emergency Air Treatment System (CREATS)

LCO 3.7.9 -+hie CREATS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS CONDITION / REQUIRED ACTION g COMPLETION TIME A. CREATS figtratin 4 rain A.1 Restore CREATS fiftratim I 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

(? inoperable. train to OPERABLE status.

\ - NOTE -

Mcontrol roo yfy be unisoed fo1 hour every 24r while in this co~ndition

/;lac isltion I dam~

CREATS Mode F.

IB.1 Restore isolation damper to 17 days I OPERABLE status.

- NOT -

Separate Conditione allowed for each damper.

I One CREATS isolation damper in one outsid wpaths er able.

R.E. Ginna Nuclear Power Plant 3.7.9-1 Amendment-&D

CREATS 3.7.9 CONDITION REQUIRED ACTION COMPLETION TIME

-e. Required Action and '1Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A 4AND not met in MODE 1, 2,3, B or 4. ,'2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

'o2 Required Action and P-associated Completion inamper(G) in CREATS Time of Condition Axode F.

not met during movement of irradiated fuel or during CORE ALTERATIONS.

7 /

D.^; Suspend CORE Immediately ALTERATIONS.

AND Suspend movement of Immediately C 3irradiated fuel assemblies.

Two CREATSmiseoeeion Y'.1 Enter LCO 3.0.3. Immediately dampor, for ono or moro P

outeide air flow paths inoperable in MODE 1, 2, 3, or 4. {i X Two CREATS iOAA1RFd Suspend CORE Immediately dampeFc for one BF MGrc ALTERATIONS.

eutside air flew peths inoperable during AND movement of irradiated r fuel assemblies or during Y'2 Suspend movement of Immediately CORE ALTERATIONS. irradiated fuel assemblies.

R.E. Ginna Nuclear Power Plant 3.7.9-2 .Amendment CREATS 3.7.9 SI IRPX/Irl I AMCrf PREQUIIRFMEMTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate!4 CREATS filtration train 2 15 minutes. 31 days

.131 SR 3.7.9.2 Perform required CREATS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). VFTP

+

SR 3.7.9.3 VerifC actuation signal. N'> I CREATSkctuates on an actual or simulated 24 months R.E. Ginna Nuclear Power Plant 3.7.9-3 Amendment-8

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs and manuals shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain:

a. The methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and does not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after review and acceptance by the onsite review function and the approval of the plant manager; and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

R.E. Ginna Nuclear Power Plant 5.5-1 Amendment&I

i Programs and Manuals 5.5 5.5.2 Primary Coolant Sources Outside Containment Program This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident. The systems include Containment Spray, Safety Injection, and Residual Heat Removal in the recirculation configuration. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

5.5.3 Deleted 5.5.4 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the plant to unrestricted areas, conforming to 10 CFR 50, Appendix I and 40 CFR 141;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days; R.E. Ginna Nuclear Power Plant 5.5-2 Amendment8-

Programs and Manuals 5.5

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix l;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table 2, Column 1;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the plant to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-I 31, iodine-I 33, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from the plant to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit Pro-ram This program provides controls to track the reactor coolant system cyclic and transient occurrences specified in UFSAR Table 5.1-4 to ensure that components are maintained within the design limits.

5.5.6 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 2.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

R.E. Ginna Nuclear Power Plant 5.5-3 Amendmen~t8t

Programs and Manuals 5.5 5.5.7 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Testing frequencies specified in Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and Required Frequencies for applicable Addenda terminologv for Inservice performing Inservice testing testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

5.5.8 Steam Generator (SG) Tube Surveillance Proaram Each SG shall be demonstrated OPERABLE by performance of an inservice inspection program in accordance with the Nuclear Policy Manual. This inspection program shall define the specific requirements of the edition and Addenda of the ASME Boiler and Pressure Code, Section Xl, as required by 10 CFR 50.55a(g). The program shall include the following:

a. The inspection intervals for SG tubes shall be specified in the Inservice Inspection Program.

R.E. Ginna Nuclear Power Plant 5.5-4 Amendment-&4t

I Programs and Manuals 5.5

b. SG tubes that have imperfections > 40% through wall, as indicated by eddy current, shall be repaired by plugging or sleeving.

c. SG sleeves that have imperfections > 30%O through wall, as indicated by eddy current, shall be repaired by plugging.

5.5.9 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. This program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.10 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems and the Spent Fuel Pool (SFP) Charcoal Adsorber System. The test frequencies will be in accordance with Regulatory Guide 1.52, Revision 2. except that in lieu of 18 month test intervals, a 24 month interval will be implemented.

The test methods will be in accordance with Regulatory Guide 1.52, Revision 2, except as modified below.

nment Post-Accident Charcoal System

1. Demons 4 pressure drop across the charcoal adsoqtw

2. Demonstrate that Fre the charcoal adsob ows a penetration and s ass

%, when tested under ambient conditions.

R.E. Ginna Nuclear Power Plant 5.5-5 Amendment94

Programs and Manuals 5.5 Guide 1.52, Revaisas O less than 14.5% W hy D380 - st temperature of h humidity of 95%.

cLAtr Containment Recirculation Fan Cooler System

1. Demonstrate the pressure drop across the high efficiency particulate air (HEPA) fitter bank is < 3 inches of water at a design flow rate (+/- 10%).

2. Demonstrate that an in-place dioctylphthalate (DOP) test of the HEPA filter bank shows a penetration and system bypass

< 1.0%.

bX Control Room Emergency Air Treatment System (CREATS)

Demonstrate the pressure drop across the i ISA filterbnk is ACV-%0 /4 EPA .C6Icr., 1. Y-11 f water at a design flow rate (+/- 10%).

Dem rate that an in-place DOP test of the HEPA filter s CAJCd 1!1.c Posy .f/rr bank shows a penetration and system bypass < 1;5

32. nomoncttu; th; progglirg droc across 3ho nharco3! ;acor.bor l hnkni is -E Incnes D; waie~~e; adeion 1nzW ralef[1 01U1

__I ,

,. Demonstrate that an in-place Freon test of the charcoal tJofve 'eo A-e~r~cio adsorber bank shows a penetration and system bypass c when tested under ambient conditions.

O ey-'C jO C Demonstrate at a laboratory test of a sample of the fcOZ% -~Ail_ charcoal adsorber, when obtained as described in Regulatory 5sj(&_f~F ~porProv.kA'I Guide 1.52, Revision.2. shows a methyl iodide penetration of 4

less than 1.84 when tested in accordance with ASTM YD , ,/l 'p-c eof v Its

- D3803-1989 at a test temperature of 300C (860 F) and a relative humidity of 95%.

C~ct SFP Charcoal Adsorber System

1. Demonstrate that the total air flow rate from the charcoal adsorbers shows at least 75% of that measured with a complete set of new adsorbers.

2. Demonstrate that an in-place Freon test of the charcoal adsorbers bank shows a penetration and system bypass

< 1.0%, when tested.under ambient conditions.

R.E. Ginna Nuclear Power Plant 5.5-6 Amendment-6+

Programs and Manuals 5.5

3. Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows a methyl iodide penetration of less than 14.5% when tested in accordance with ASTM D3803-1989 at a test temperature of 30 0C (86 0F) and a relative humidity of 95%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP frequencies.

5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas decay tanks and the quantity of radioactivity contained in waste gas decay tanks. The gaseous radioactivity quantities shall be determined following the methodology in NUREG-0133.

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the waste gas decay tanks and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in each waste gas decay tank is less than the amount that would result in a whole body exposure of 2 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.12 Diesel Fuel Oil Testing Proaram A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits, R.E. Ginna Nuclear Power Plant 5.5-7 Amendment-M4

Programs and Manuals 5.5

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. a clear and bright appearance with proper color; and

b. Within 31 days following addition of the new fuel to the storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 2D fuel oil.

5.5.13 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license; or

2. A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 5.5.13.b.1 or Specification 5.5.1 3.b.2 shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71e.

5.5.14 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected; R.E. Ginna Nuclear Power Plant 5.5-8 Amendment-&

t Programs and Manuals 5.5

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the supported system(s) is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the inoperable support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.15 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pat is 60 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.2% of containment air weight per day.

R.E. Ginna Nuclear Power Plant 5.5-9 Amendment 81-

Programs and Manuals 5.5 Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is

• 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. For each air lock, overall leakage rate is

• 0.05 La when tested at 2 P.a and

2. For each door, leakage rate is < 0.01 La when tested at 2 Pa.

c. Mini-purge valve acceptance criteria is < 0.05 La when tested at 2 Pa.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

R.E. Ginna Nuclear Power Plant 5.5-1 0 Amendment64

Delete new sections 5.5.16 and 5.6.7 proposed in Reference 3 Typed Sections Note: The typed page 5.5-6 will be resubmitted when the value associated with 5.5.1O.b.1 is determined during system operational testing.

CREATS 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Control Room Emergency Air Treatment System (CREATS)

I LCO 3.7.E Two CREATS Trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4, During movement of irradiated fuel assemblies, During CORE ALTERATIONS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CREATS train A.1 Restore CREATS train to 7 days inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, 3, or4.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. Required Action and C.1 Suspend CORE Immediately associated Completion ALTERATIONS.

Time of Condition A not met during movement of AND irradiated fuel or during CORE ALTERATIONS. C.2 Suspend movement of Immediately irradiated fuel assemblies.

D. Two CREATS trains D.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or4.

E. Two CREATS trains E.1 Suspend CORE Immediately inoperable during ALTERATIONS.

movement of irradiated fuel assemblies or during AND CORE ALTERATIONS.

E.2 Suspend movement of Immediately irradiated fuel assemblies.

R.E. Ginna Nuclear Power Plant 3.7.9-1 Amendment

CREATS 3.7.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Operate each CREATS filtration train 215 minutes. 31 days SR 3.7.9.2 Perform required CREATS filter testing in accordance In accordance with l with the Ventilation Filter Testing Program (VFTP). VFTP SR 3.7.9.3 Verify each CREATS train actuates on an actual or 24 months simulated actuation signal.

R.E. Ginna Nuclear Power Plant 3.7.9-2 Amendment

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs and manuals shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain:

a. The methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

1. sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s),

2. a determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and does not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after review and acceptance by the onsite review function and the approval of the plant manager; and

c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

R.E. Ginna Nuclear Power Plant 5.5-1 Amendment

Programs and Manuals 5.5 5.5.2 Primary Coolant Sources Outside Containment Program This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident. The systems include Containment Spray, Safety Injection, and Residual Heat Removal in the recirculation configuration. The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

5.5.3 Deleted 5.5.4 Radioactive Effluent Controls Proaram This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the plant to unrestricted areas, conforming to 10 CFR 50, Appendix I and 40 CFR 141;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days; R.E. Ginna Nuclear Power Plant 5.5-2 Amendment

Programs and Manuals 5.5

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix l;

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with 10 CFR 20, Appendix B, Table 2, Column 1;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from the plant to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from the plant to areas beyond the site boundary, conforming to 10 CFR 50, Appendix l; and

j. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

5.5.5 Component Cyclic or Transient Limit Proaram This program provides controls to track the reactor coolant system cyclic and transient occurrences specified in UFSAR Table 5.1-4 to ensure that components are maintained within the design limits.

5.5.6 Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 2.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

R.E. Ginna Nuclear Power Plant 5.5-3 Amendment

Programs and Manuals 5.5 5.5.7 Inservice Testinc Proaram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shall include the following:

a. Testing frequencies specified in Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and Reauired Freauencies for applicable Addenda terminology for Inservice nerformina Inservioe testing testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

5.5.8 Steam Generator (SG) Tube Surveillance Program Each SG shall be demonstrated OPERABLE by performance of an inservice inspection program in accordance with the Nuclear Policy Manual. This inspection program shall define the specific requirements of the edition and Addenda of the ASME Boiler and Pressure Code, Section Xl, as required by 10 CFR 50.55a(g). The program shall include the following:

a. The inspection intervals for SG tubes shall be specified in the Inservice Inspection Program.

R.E. Ginna Nuclear Power Plant 5.5-4 Amendment

Programs and Manuals 5.5

b. SG tubes that have imperfections > 40% through wall, as indicated by eddy current, shall be repaired by plugging or sleeving.

c. SG sleeves that have imperfections > 30% through wall, as indicated by eddy current, shall be repaired by plugging.

5.5.9 Secondary Water Chemistry Proaram This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. This program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;

c. Identification of process sampling points;

d. Procedures for the recording and management of data;

e. Procedures defining corrective actions for all off control point chemistry conditions; and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.10 Ventilation FilterTesting Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature filter ventilation systems and the Spent Fuel Pool (SFP) Charcoal Adsorber System. The test frequencies will be in accordance with Regulatory Guide 1.52, Revision 2, except that in lieu of 18 month test intervals, a 24 month interval will be implemented.

The test methods will be in accordance with Regulatory Guide 1.52, Revision 2, except as modified below.

a. Containment Recirculation Fan Cooler System

1. Demonstrate the pressure drop across the high efficiency particulate air (HEPA) filter bank is < 3 inches of water at a design flow rate (i 10%).

2. Demonstrate that an in-place dioctylphthalate (DOP) test of the HEPA filter bank shows a penetration and system bypass

< 1.0%.

R.E. Ginna Nuclear Power Plant 5.5-5 Amendment

Programs and Manuals 5.5

b. Control Room Emergency Air Treatment System (CREATS)

1. Demonstrate the pressure drop across the combined HEPA filters, the prefilters, the charcoal adsorbers and the post-filters is < [ inches of water at a design flow rate (+/- 10%).

2. Demonstrate that an in-place DOP test of the HE PA filter bank shows a penetration and system bypass < 0.05%.

3. Demonstrate that an in-place Freon test of the charcoal adsorber bank shows a penetration and system bypass

< 0.05%, when tested under ambient conditions.

4. Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows a methyl iodide penetration of less than 2.5% when tested in accorda nce with ASTM D3803-1989 at a test temperature of 30 0C (86 0F) and a relative humidity of 95%.

c. SFP Charcoal Adsorber System

1. Demonstrate that the total air flow rate from the charcoal adsorbers shows at least 75% of that m easured with a complete set of new adsorbers.

2. Demonstrate that an in-place Freon test of the charcoal adsorbers bank shows a penetration and system bypass

< 1.0%, when tested under ambient conditions.

3. Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows a methyl iodide penetration of less than 14.5% when tested in accord ance with ASTM D3803-1989 at a test temperature of 300 C (86 0 F) and a relative humidity of 95%.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP frequencies.

5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas decay tanks and the quantity of radioactivity contained in waste gas decay tanks. The gaseous radioactivity quantities shall be determined following the methodology in NU REG-0133.

The program shall include:

R.E. Ginna Nuclear Power Plant 5.5-6 Amendment

Programs and Manuals 5.5

a. The limits for concentrations of hydrogen and oxygen in the waste gas decay tanks and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in each waste gas decay tank is less than the amount that would result in a whole body exposure of > 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.12 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. an API gravity or an absolute specific gravity within limits,

2. a flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. a clear and bright appearance with proper color; and

b. Within 31 days following addition of the new fuel to the storage tanks, verify that the properties of the new fuel oil, other than those addressed in a. above, are within limits for ASTM 2D fuel oil.

5.5.13 Technical Specifications (TS) Bases Control Proaram This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

R.E. Ginna Nuclear Power Plant 5.5-7 Amendment

Programs and Manuals 5.5

1. A change in the TS incorporated in the license; or

2. A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 5.5.13.b.1 or Specification 5.5.13.b.2 shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71e.

5.5.14 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions; This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion lime is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the supported system(s) is also inoperable; or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or R.E. Ginna Nuclear Power Plant 5.5-8 Amendment

Programs and Manuals 5.5

c. A required system redundant to the inoperable support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.15 Containment Leakaae Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa. is 60 psig.

The maximum allowable primary containment leakage rate, La, at Pat shall be 0.2% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is

• 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are

• 0.60 La for the Type B and Type C tests and

• 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1. For each air lock, overall leakage rate is

• 0.05 La when tested at 2 Pa, and

2. For each door, leakage rate is

• 0.01 La when tested at 2 Pa.

c. Mini-purge valve acceptance criteria is

• 0.05 La when tested at 2 Pa.

R.E. Ginna Nuclear Power Plant 5.5-9 Amendment

Programs and Manuals 5.5 The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

R.E. Ginna Nuclear Power Plant 5.5-1 0 Amendment

Marked-up Bases Sections Note: These bases pages are being provided for information only to show the changes that Ginna intends to make following approval of the LAR. The bases are under Ginna control for all changes in accordance with Specification 5.5.13. Ginna requests that the NRC document acceptance of these bases changes in the SER.

CREATS B 3.7.9 B 3.7 PLANT SYSTEMS B 3.7.9 Control Room Emergency Air Treatment System (CREATS)

BASES BACKGROUND According to Atomic Industry Forum (AIF) GDC 11 (Ref. 1), a control oom shall be provided which permits continuous occupancy under any

/o r I OC F 504* 7 .) redible postaccident condition without excessive radiation exposures of

/3- Z; Z' ersonnel. Exposure limits are provided in GDC 19 of 10 CFR 50, 6.

' ' Appendix A (Ref. 2) r iros that control room A ff effe sc~re Iquih fs.C. mstrictod to 5m r- body, o its for the duration of the

)' cac >e aecident. The CREATS provides a protected environment from whic

=e..'&c O t operators can control the plant following an u noorollo release of

/ , S rr.ct- z'

/sE t o radioactivity for 30 days without exceeding this 5 rem limit.

Dt -~ - JThc CREATS ic part of tho Control Building vontilation Gysctom.

too Z^, > The CREATS consists oIa high efficiency particulate air (HEPA) filter,

( ch/ * \ activated charcoal ads rs for removal of gaseous activity (principally e267 Ctrtvb andL+vh faiodines), room roturn air fan and omorgoncy rotum fanzcnAtrol

-eii ef)u(see Figure B 3.7.9-1). Ductwork, dampers, and instrumentation also form part of the system (Ref. 3).

• c ,< ,,, The CREATS is an emergency system, pazrt; of which msy oporFt redircLRt during norm palpmnt oportionc. Actuation of the CREATS p!6r the

, A , Rcytom in ono of four copsito ctstoc of tho omorgoncy modo of OAkpevs /kD2, loperation,depending on-the initiotien sige!l. The felwi-wng aro the Ao,__P. A lf% . I- - I - . J .. - ,- . -; ov^

/ °3,/ V -/"t,/'/ , normi sna omorgoncy r s onc otfi peratldn tor {

jkb3, AebD2, Zt Ab.23 / 6REATS ModeA rec',-77 e Mvc%/ The C ATS is in the standby mode with the excepti hat the sc~ .4( Acontrol rooturn air fan is in operation.

c/t4AI.. r" Ced '( CREATSModeB or e cc~(J/r~t \

e LrThis o mode was permane at under Reference 6.

_.CRFATS Mode C s . ^ . _ . . .__ _

This is the CREATS figuration follo an accident with a radiation release detected by radiation itor R-1. Upon receipt of an iuation signal, the control room ergency return air fan will tuate and system dampers align to re culate a nomin OOO cfm (approximately one fourth of the Co Building VepKation System design) through the CREATS charcoa d APA filters. CREATS automatic isolation dampers AKDOI a R.E. Ginna Nuclear Power Plant B 3.7.9-1 Revision 30

CREATS B 3.7.9 1kKDO4, and permanently closed isolation damper AKDO8, isolat th GREATS from outside air.

CRE TS Mode D This is tCe GREATS configuration following the detectio of smoke within the ontrol Building. Upon receipt of an actuati signal, the system con nues to draw outside air. However, the ontrol room emergency rum air fan will actuate and system mpers align to recirculate a n 2000 cfm through the CRE S and HEPA 1inal filters. This effe vely purges the control room/ir environment.

CREATS Mode E This is the same GRE TS configuration s Mode D with exception that all outside air is isol ed to the co rol room by one damper in each air supply flow path.

CREATS Mode F This is the CREATS configur i following the detection of a toxic gas as indicated by the chlone o ammonia detectors, or high radiation as detected by 5 or Ra. Upon receipt of an actuation signal, the system align itself consis nt with Mode C except that two dampers in each i supply path a isolated.

Normally open air pply isolation dampe are arranged in series so that the failure f one damper to close wi not result in a breach of isolation.

The air ente ng the control room is continuously onitored by radiation d toxic gas detectors. One detector o ut above the setpoint ill cause actuation of the emergency radi *on state or toxic g isolation state, as required. The actions oft e toxic gas and gh radiation state (Mode F) are more restrictive thn the em rgency radiation state (Mode C) and close an add iti al ation damper in the outside air flowpaths. Only the hig adiation state CREATS Mode F is addressed by this LCO.

APPLICABLE The location of components and CREATS related ducting within the SAFETY control room emergency zone envelope ensures an adequate supply of ANALYSES filtered air to all areas requiring access. The CREATS provides airborne radiological protection for the control room operators in MODES 1, 2, 3, and 4, as demonstrated by the control room accident dose analyses for Ac o'aOp tcIC tho moct limiting design baric logs of coolant accidont and ctcom gonoator tubo rupturo (Ref. 3). This analysis shows that with credit for the CREATS, orAwith crodit for instantannou; isolation ofthe control room R.E. Ginna Nuclear Power Plant B 3.7.9-2 Revision 30

CREATS B 3.7.9

-'Incidont ".ith the accident initiator Ind no CREATS filtration traiCn S (34iaeble, the dose rates to control room personnel remain withinG;G 1 limits.

ecurrent control room dose analysis assumes 45;8 cfm of in-leakage dur esign basis accidents (DBA), which correlates to a 5.25squ -

inch brea Therefore a breach of up to 5.25 square inches w require no co nsatory actions. A sensitivity analysis aping potassium iodide is ingested within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of th A Indicates an in-leakage of greatH t several thousand cfn be tolerated and remaine within thedo its.

... The current toxic chemical ana iates 500 cfm of in-leakage would be required before oxic limits Id be exceeded. This correlates to a 58.9 are inch breach. Sinc has no impact on toxic chemicals, co oom in-leakage is driven by theemical analysis, even con rIng KI compensatory actions. However, gas prote n is not a technical specification requirement and th fore LCO

, .3 does not need to be entered for exceeding this limit.

During movement of irradiated fuel assemblies or during CORE ALTERATIONS, the CREATS ensures control room habitability in the event of a fuel handling accident. I t has been demonstrated that the r 1 CL T ib/-AA CREATS is not required in the event of a waste gas decay tank rupture

( o7for d Aowo/<v1 l(Ref. 5, i?,/ The CREATS satisfies Criterion 3 of the NRC Policy Statement.

LCO -tration trainFand two independent and redundant isolation damper trains all of which are required to be OPERABLE. Total system failure could result in exceeding a dose of 5 V j re to the control room operators in the event of a large radioactive reease.

The CREATS is considered OPERABL the individual components necessary to permit CREATS p e tion are OPERABLE (see Figure B 3.7.9-1). .Fk&CREAT rain is OPERABLE when the A_ associated: 9 ot c-;Wm-1 A;dzo~9Q~ etu ai as 2i S OPERABLE and capable of providing forced flow;

b. HEPA filters and charcoal adsorbers for the omorgonc'; roturn air

-4M+are not excessively restricting flow, and re capable of performing their filtration functions;=7r R.E. Ginna Nuclear Power Plant B 3.7.9-3 Revision 30

.~-Ird*O~h)'C~'1Ts ,ijCREATS aeB. / B3.7.9 4 cap g 4B LaE. c. Ductwork and autematie dampers ArKD07 end A449 a vD nre orvperp S tA9 1) o ao, XDU7DE) OPERABLE, and the CREATS filter flow is a nominal fcfm' oua.c A D b 3aAd ar-e 'cx- , BsLo4Eit9 ASutA tAQe 4 heg; E ~ ahG utomatcioa ionapersar cnsi iere RABLE 1 D >, R 7 RD2 2 CK ia D hen the damper (AI(D01, AO014, AKD06, and /zAK9 ' ) can close on an actuation signal to isolate outside air or is closed with motive force are'~ CiocSi-o' x ~te removed. The CREATS isolation damper AKDO3 isOPERABLE ohie maintainod in a cloosed and deactivated tate (fRof. 6). Two damporc aFm pran io for o CREATEl otcdo air path.

In ~ - 4fadd e control room emergency zone boundary must be amaintained, including the integrity of the walls, floors, ceilings, ductwork, tAz e Jo't, .- j s and personnel access doors. Tho control room omorgoncsy zon boundary also includac the control room lavatory oxhauct dampfr n s6e /

7er 4

(AKDl2), which Glococ OnR RR actaticm sinRal, aRd ac-Acsocriateo; d dwtomrwk WA. it -rjerrte Opening of the personnel access doors for entry and exit does not violate t-e-w"'Oed 5u( S the control room emergency zone boundary. A personnel access door or I f ra ventilation system ductwork access door may be opened for extended I__ e periods provided a dedicated individual is stationed at the access door to t

iensure closure, if required (i.e., the individual Qprforms the isolation 4

0. oe j /as~ . ofunction ,the door is able to be closed within 4:seconds upon indian of the need to ;lse the door, end the CR.ATS filtration train Isc E PERAE3BE.frhe ventilation ductwork may also be opened for.extended periods provided that the REATS filtration train is declared inoperable, M a the fensarcoff,and the portion of ductwork that is open is isolated from

/Uop..Pr i d'tic t' S~ \the control room by a damper that is closed with motive force removed or t>;c M' i'e ( / Age/ a passive isolation device. Who control room air handling u'nit rced, /Avd ~ g Ae c t omporaturo control dampor. (AVD42, A.KD13, and AI'DI4) provlide for O jo, A iflowraFrd, let odamprori;+t!Qn.

1<aziX he Oar#>1 rowx3d-.

aSaK

' ~ot p s

/

,roo breach greater than-5.5 squareinches is discovered in the contw mergency zone boundary and cannot be immediately seal, the Z°followinaompensatory O r i actions must be implemented to ensign compliance th GDC 19, per Regulatory Guide 1.196. jiese actions cannot be pe ed, then LCO 3.0.3 must be enter

• The on duty per nel required by Sp i cation 5.2.2.a, b, c and f must take a potassi odide (KI) et within two hours of any valid automatic isolation CR S that occurs from the radiation monitors due to a radiolog ievent in progress, or as directed by the Emergency Resp Or ation.

• Provisions mu made to provide ( o the oncoming shifts as required.

The abo ompensatory measures are not intendeda substitute for tim action to seal control room boundary breaches, but roer to sure safe plant operation while maintenance activities proce R.E. Ginna Nuclear Power Plant B 3.7.9-4 Revision 30

CREATS B 3.7.9 r;6,:;0 rZ7" ,. 5 APPLICABILITY In MODES 1, 2,3, and 4,4he CREATS'nust be OPERABLE to control operator exposure during and following a DBA.

During movement of irradiated fuel assemblies or during CORE 4 ALTERATIONS CREATS ust be OPEROSE to cope with the release rom a uel handling accident.

ACTIONS 6-With e CREATS filtration train inoperable action must be taken to restore OPERABLE status within 4ePi-us eo iselete thee t4-reoeomA adED-5 Wrm ebutcide sir. 1H thiG 9onditio, tho icoICWJR d1aFmpBFG We zidezuate te perform the control room protection function but-no moanc oxict to fiQtor the rclcesc of redieoative gas within the control room. The Completion Time is based on the low probability of a DBA occurring during this time frame, and the ability of the CREATS deapers te isolate Cfet7- 0 itftJo~

OPP143 the eontrol 8rrom.

cP1bZtd 4o, oiAT-cc,

'T"JDC ZtCCI01 TRequd6fonA.2 is modified by a Note which allows the coze to be unisolate ' eery 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TI i i lmakeup to improvie the worklr hin the control room and ofu /oe-Wiego. ZA is acceptab elow probability of a Ddr ts up period.

ne CREATS isolation damper inoperable for one or more out air flow action must be taken to restore OPERABLE within 7 days. In this C oftdk. the remaining OPERABL TSiolation damper is adequate to p-& f the control Poetn uction.

However, the overall reliability is erbAu0 se a single failure of the OPERABLE CREATS -iso Ia mpe r ult in loss of the CREATS isolation fun g. The 7 day Co mpletio based on the low probabili )BA occurring during this time perioda ity of the reing isolation damper to provide the required isolation ability.

,h. and 92 A In MODE 1, 2, 3, or 4, if the Required Actions of Conditio nnot be completed within the required Completion Time, the plant must be*

placed in a MODE that minimizes accident risk. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full R.E. Ginna Nuclear Power Plant B 3.7.9-5 Revision 30

CREATS B 3.7.9 power conditions in an orderly manner and without challenging plant systems.

4c, / <>u"/ C. 2:

D., . D.2.L. SHIV D.2.2 During movement of irradiated fuel assemblies or dun g ALTERATIONS, if the Required Actions of Conditio A anot be completed within the required Completion Time, action must be taken to

_y ~fiermdiately place !he O)PERABLE- iselatien damnper(s) in CREA/TS Mevde

. TH5 _eie emt fs ! h_7 . t te \eann depie We GPGRARLS, h no failubroc pro'nting automatic actuation will occur, and that any

\ -Am alternative to Required Actiont D1ismmediately suspend activitie that could result in a release of radioactivity that might enter the control room. This requires the suspension of CORE ALTERATIONS and the suspension of movement of irradiated fuel assemblies. This places the plant in a condition that minimizes risk. This does not preclude the gvement of fuel or other components to a safe position.

In MODE 1,2, 3, or 4, if both CREATS isolation dampers for one or moro outside 21r flgA p2ahs are inoperable, the CREATS may not be capable of performing the intended function and the plant is in a condition outside the accident analyses. Failuro of tho initgrity of the control room omOrgoncy; zono boundary (i.W., wa'lls, floor, ceiling, duot WorlF, pcrco^ nnl occc6 doors, -or ontrol room l--torn' eoxh2--t dampeor A4D)92) O8Ib Fcults in a condition outcido the accident an aitsos.

Therefore, LCO 3.0.3 must be entered immediately.

X.1 andY2 During movement of irradiated fuel assemblies or during CORE ALTERATIONS with two CREATS isolation dampors for one ormoro outsido ir flow paths inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might enter the control room. This requires the suspension of CORE ALTERATIONS and the suspension of movement of irradiated fuel assemblies. This places the plant in a condition that minimizes accident risk. This does not preclude the movement of fuel or other components to a safe position.

SURVEILLANCE SR 3.7.9.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions R.E. Ginna Nuclear Power Plant B 3.7.9-6 Revision 30

CREATS B 3-7.9 on this system are not too severe, testing each CREATS filtration train once every 31 days for 2 15 minutes provides an adequate check of this system. The 31 day Frequency is based on the reliability of the b r 9 ;1- , -dcAl Y ) equipmenf ;

CoU C Ito, 6.Aty SR 3.7.9.2 This SR verifies that the required CREATS testin perfo in accordance with the Ventilation Filter Testing Pr gram (V TP). The CREATS filter tests are in general accordanc ith Re latory Guide 1.52 (Ref. 4). The VFTP includes testing th perfor ce of the HEPA K

(_tY filter, charcoal adsorber efficiency, of the activated charcoal. The uhrougsoCREATS filtration train isp flow rae, and the physical re d flowrate cubic feet per minute (+/-10%).

--ODCI Specific test Frequencies and additional information are discussed in detail in the VFTP. However, the maximum surveillance interval for refueling outage tests is based on 24 month refueling cycles and not 18 month cycles as defined by Regulatory Guide 1.52 (Ref. 4).

9fR 3.7.9.3 t /

This SR verifies that-4he CREATSI rain starts and operates and that each CREATS automatic damper actuates on an actual or simulated actuation signal. The Frequency of 24 months is based on Regulatory Guide 1.52 (Ref. 4).

REFERENCES 1. Atomic Industry Forum (AIF) GDC 11, Issued for comment July 10, 1967.

2. 10 CFR 50, Appendix A, GDC 19.

3. UFSAR, Section 6.4.

4. Regulatory Guide 1.52, Revision 2.

5. Letter from Robert C. Mecredy, RG&E, to Guy S. Vissing, NRC,

Subject:

Application for Amendment to Facility Operating License Control Room Emergency Air Treatment System (CREATS)

Applicability Change (LCO 3.3.6 and LCO 3.7.9), dated July 21, 2000.

6. PCR 2003-0019, "C REATS Modifications Inside of the Control Room Emergency Zone".

\- 7. /a C FR er-67, cc,-ts 5 - 7A, 91 D b A.C - Z 1S-;Z0 9 /ter 7-rvfacO e, o e 4toenoee CfF:51ec a~t-e sooth/5 4 ds ~-

R.E. Ginna Nuclear Power Plant B 3.7.9-7 Revision 30

The value of 2.5% methyl iodide penetration was chosen for the laboratory test sample because, even though the new system contains 4-inch charcoal beds, the design face velocity Ad of Ginna's system results in approximately one half of the standard 0.25 second per 2-inch 0 residence time. Therefore, since the resulting residence time is 0.25 seconds per 4-inch bed depth, the 2-inch bed depth section of Regulatory Guide 1.52, Revision 3, Table 1, was considered appropriate.

CREATS B 3.7.9 AIR HANDLING UNIT I .. _ .. _ . . _ . . _ .. _ . . _ .. _

AKL03 HEATNG I COOLING COIL AKD01

,

• z ROUGHING

~SUPPLY FAN 4' ';

4, I.

Note 3 I FILTER * - -

AK~D1 A02: a:AKD1:

AKCO7 AKF07 S

... CP..N.

EMNRGEI'M L REnURN FAN CONTROL ROOM ETR .N .

RETURN FAN EAST

, ALLOF CONTROL ROOM Legend:

- - CREATS Filbation Notes: I For illustmion only dampers Include AKD0O1,1 IAKDDO. N 2.The ontraln does not include the Air Handling Unit

3. AKD tUydeactivated In the closed position.

Figure B 3.7.9-1 CREATS R.E. Ginna Nuclear Power Plant B 3.7.9-8 Revision 30

CREATS Design AMC5 Jwm Ia

• = CONTROL ROOM EMERGENCY ZONE BOUNDARY

• = EXISTING CREATS & HEATING/COOLING

( = PROPOSED NEW SYSTEMS