Text

Proposed License Amendment Request Adoption of TSTF-490, Revision 0, Regarding Deletion of E Bar Definition and Revision to RCS Specific Activity Using the Consolidated Line Item Improvement Process
	ML083530982
Person / Time
Site:	North Anna
Issue date:	12/17/2008
From:	Price J; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	08-0729, TSTF-490, Rev 0
	Download: ML083530982 (29)
	v • d • e

10 CFR 50.90 VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 17, 2008 U.S. Nuclear Regulatory Commission Serial No. 08-0729 Attention: Document Control Desk NL&OS/ETS RO Washington, D.C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST ADOPTION OF TSTF-490, REVISION 0, REGARDING DELETION OF E BAR DEFINITION AND REVISION TO RCS SPECIFIC ACTIVITY USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS Pursuant to 10 CFR 50.90, Dominion requests amendments, in the form of changes to the Technical Specifications to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2. The proposed changes would replace the current Technical Specification (TS) 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent XE-133 definition that would replace the current E Bar average disintegration energy definition.

The changes are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec." The availability of this TS improvement was announced in the Federal Register on March 15,2007 (72 FR 12217) as part of the Consolidated Line Item Improvement Process (CLlIP). provides a description and assessment of the proposed changes, as well as confirmation of applicability. Attachment 2 provides the marked-up TS pages to show the proposed changes. Attachment 3 provides the proposed TS Pages. provides the TS Bases pages for information only.

The proposed changes have been reviewed by the Facility Safety Review Committee.

Dominion requests approval of the proposed license amendment by June 30, 2009 with the amendment being implemented within 60 days of approval.

Serial No. 08*0729 Docket Nos. 50-338/339 Res Specific Activity Page 2 of 3 In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the appropriate designated officials of Virginia.

If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Very truly yours, a Price President - Nuclear Engineering Attachments

1. Description and Assessment of Proposed Changes

2. Marked-Up Technical Specification Pages

3. Proposed Technical Specification Pages

4. Technical Specification Bases Changes (information only)

COMMONWEALTH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. Alan Price, who is Vice President - Nuclear Engineering, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this/771 day ot/tiOrrn!u.4) ,2008.

i/o )/1 ~

1/d" I~'~

- - - - Notary ublic VICKI L. HULL

Notary Public Commonwealth of VIrginia

1.0~2

My commlilion bpi,.. May 31, 2010

Serial No. 08-0729 Docket Nos. 50-338/339 Res Specific Activity Page 3 of 3 cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Suite 300 Glen Allen, Virginia 23060 State Health Commissioner Virginia Department of Health h

James Madison Building - i floor 109 Governor Street Suite 730 Richmond, Virginia 23219 NRC Senior Resident Inspector North Anna Power Station Ms. D. N. Wright NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 H4A 11555 Rockville Pike Rockville, Maryland 20852 Mr. J. F. Stang NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-G9A 11555 Rockville Pike Rockville, Maryland 20852

Serial No. 08-0729 Docket Nos. 50-338/339 Res Specific Activity Attachment 1 Description and Assessment of Proposed Changes North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Serial No. 08-0729 Docket Nos. 50-338/50-339 Page 1 of 4 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGES

1.0 DESCRIPTION

Virginia Electric and Power Company (Dominion) is requesting amendments to Operating Licenses NPF-4 and NPF-7 for North Anna Units 1 and 2, respectively. The proposed changes would replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity. The noble gas activity would be based on DOSE EQUIVALENT XE-133 and would take into account only the noble gas activity in the primary coolant. The changes were approved by the NRC staff Safety Evaluation (SE) dated September 27, 2006 (ADAMS ML062700612) (Reference 1).

Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" was announced for availability in the Federal Register on March 15, 2007 (72 FR 12217) as part of the consolidated line item improvement process (CLlIP).

2.0 PROPOSED CHANGE

S Consistent with NRC-approved TSTF-490, Revision 0, the proposed TS changes:

Revise the current definition of DOSE EQUIVALENT 1-131, but maintain the existing approved dose conversions factors.

Delete the definition of "E Bar, AVERAGE DISINTEGRATION ENERGY."

Add a new TS definition for "DOSE EQUIVALENT XE-133."

Revise LCO 3.4.16, "RCS Specific Activity" to delete references to gross specific activity; add limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133; and delete Figure 3.4.16-1, "Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit versus Percent of RATED THERMAL POWER."

Revise LCO 3.4.16 "Applicability" to specify the LCO is applicable in MODES 1, 2,3, and 4.

Modify ACTIONS Table as follows:

o Condition A is modified to delete the reference to Figure 3.4.16-1, and define an upper limit that is applicable at all power levels.

o Condition B is modified to provide a Condition and Required Action for DOSE EQUIVALENT XE-133 instead of gross specific activity. The Completion Time is changed from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . A Note allowing the applicability of LCO 3.0.4.c is added, consistent with the Note to Required Action A.1.

o Condition C is modified based on the changes to Conditions A and B and to reflect the change in the LCO Applicability.

Serial No. 08-0729 Docket Nos. 50-338/50-339 Page 2 of 4 Revise SR 3.4.16.1 to verify the limit for DOSE EQUIVALENT XE-133. A Note is added, consistent with SR 3.4.16.2 to allow entry into MODES 2,3, and 4 prior to performance of the SR.

Delete SR 3.4.16.3.

Variance from the TSTF The reviewer's note (in TSTF-490) associated with the definition of 1-131 states: "The first set of thyroid dose conversion factors shall be used for plants licensed to 10 CFR 100.11. The following Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) conversion factors shall be used for plants licensed to 10 CFR 50.67."

Thyroid dose conversion factors from:

a. Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or

b. Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977, or

c. ICRP-30, 1979, Supplement to Part 1, Page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or

d. Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

Committed Dose Equivalent (CDE) or Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11.]

Although North Anna is licensed to 10 CFR 50.67, North Anna is maintaining the current conversion factors in the definition of 1-131. The current definition of dose equivalent 1-131 allows dose equivalent iodine to be calculated using either TID-14844 or RG 1"109 dose conversion factors.

Serial No. 08-0729 Docket Nos. 50-338/50-339 Page 3 of 4 Based upon the following arguments, which were presented in letters dated September 12, 2003 and May 7, 2004 (Serial No.03-464 and 03-4640), the definition and the application of dose conversion factors were approved in amendments 240/221, dated June 15, 2005 (TAC Nos. MC0776 and MC0777).

RG 1.183 requires that the pre-accident and concurrent iodine spikes used in the design basis analysis be based on the maximum value permitted by Technical Specifications.

The use of FGR-11 dose conversion factors to calculate dose is consistent with the Total Effective Dose Equivalent methodology described in RG 1.183. However, the use of either the RG 1.109 or TID-14844 dose conversion factors to perform the Technical Specification surveillance for dose equivalent 1-131 will restrict plant operations to a lower total allowable iodine inventory in the RCS than would be attainable using FGR-11 dose conversion factors. The 1 jJCi/gm dose equivalent 1-131 inventory calculated using RG 1.109 dose conversion factors was used to establish the design basis analysis source term for both the pre-accident and concurrent iodine spikes. The use of the RG 1.109 dose conversion factors to determine the design basis analysis source term bounds the use of TID-14844 dose conversion factors. Therefore, use of the RG 1.109 dose conversion factors in the design basis analysis is consistent with the proposed change in the Technical Specification definition of dose equivalent 1-131 and the requirement to use the maximum value permitted by Technical Specifications.

It is acceptable for the pre-accident and concurrent iodine spike source terms to be based on RG 1.109 dose conversion factors and the doses to be calculated using FGR-11 dose conversion factors because the source term bounds allowable plant operating parameters as defined in the Technical Specifications.

3.0 BACKGROUND

The background for this application is as stated in TSTF-490, Revision 0 (Reference 1),

the NRC Notice for Comment (Reference 2), and the model SE in NRC's Notice of Availability (Reference 3).

4.0 TECHNICAL ANALYSIS

Dominion has reviewed References 1, 2 and 3, and the model SE published as part of the CUIP Notice for Comment. Dominion has applied the methodology in Reference 1 to develop the proposed Technical Specification changes. Dominion has also concluded that the justifications presented in TSTF-490, Revision 0 and the model SE prepared by the NRC staff are applicable to North Anna Units 1 and 2 and justify this amendment for the incorporation of the changes to the North Anna Units 1 and 2 Technical Specifications.

5.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in TSTF-490, Revision 0 (Reference 1), the

Serial No. 08-0729 Docket Nos. 50-338/50-339 Page 4 of4 NRC Notice for Comment (Reference 2), and the model SE in NRC's Notice of Availability (Reference 3).

5.1 No Significant Hazards Consideration Dominion has reviewed the proposed no significant hazards consideration determination published in the Federal Register on March 15, 2007 (72 FR 12217), as part of the CUIP. Dominion has concluded that the proposed determination presented in the notice is applicable to North Anna Units 1 and 2 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

5.2 Environmental Evaluation Dominion has reviewed the environmental consideration included in the model SE published in the Federal Register on March 15,2007 (72 FR 12217) as part of the CUIP. Dominion has concluded that the staff's findings presented therein are applicable to North Anna Units 1 and 2 and the determination is hereby incorporated by reference for this application.

6.0 REFERENCES

1. NRC Safety Evaluation (SE) approving TSTF-490, Revision 0 dated September 27,2006.

2. Federal Notice for Comment published on November 20, 2006 (71 FR 67170).

3. Federal Notice of Availability published on March 15,2007 (72 FR 12217).

Serial No. 08-0729 Docket Nos. 50-338/339 Res Specific Activity Attachment 2 Marked-up Technical Specification Pages North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Definitions 1.1 1.1 Definitions CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal TEST (COT) into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle.

These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5.

Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 Shall be that oonoentration of I 131 (mioroourieslgram) that alone '/olould produoe the same thyroid dose as the quantity and isotopic mixture of I 131, I 132, I 133, I 134, and I 135 aotually present. DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using t+he thyroid dose conversion factors used for this oaloulation shall be those listed in Table III ofTID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977-:

North Anna Units 1 and 2 1.1-2 Amendments 240/221

Definitions 1.1 DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" or the average gamma disintegration energies as provided in ICRP Publication 38, "Radionuclide Transformations.:

E AVERAGE E shall be the average (weighted in proportion to the DISINTEGRATION ENERGY concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, 'Nith half lives> [15] minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (Le., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

North Anna Units 1 and 2 1.1-3 Amendments 248/228

Res Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPLICABILITY: MODES 1 and 2, 1, 2, 3, and 4.

MODE 3 '/lith RCS average temperature Fav!}) > 500°F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT --------------------NOTE-------------------

1-131 not within LCO 3.0.4.c is applicable.

limit> 1.0 IJCi/gm. ------------------------------------------------

A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1-131 within the acceptable region of Figure 3.4.16 1

< 60 !lCi/gm.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

S. Gross specific activity of --------------------NOTE-------------------

the reactor coolant not LCO 3.0.4.c is applicable.

l/lithin limit.DOSE ------------------------------------------------

EQUIVALENT XE-133 not within limit. B.1 Be in MODE 3 !\/ith M8 hours

+avg < 500°F.Restore DOSE EQUIVALENT XE-133 to within limit.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Be in MODE 5.

OR 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DOSE EQUIVALENT 1-131 > 60 J.l.Ci/gm.

North Anna Units 1 and 2 3.4.16-1 Amendments 231/212 I

Res Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity

< 100/6 IJCi/gm.

NOTE------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-133 7 days specific activity < 197 uCilgm.

SR 3.4.16.2 -------------------------------NOTE------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity s 1.0 IJCi/gm.

Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of ~ 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period SR 3.4.16.3 NOTE Not required to be pertormed until 31 days after a minimum of 2 effective full pm'Jer days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for> 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine 6 from a sample taken in MODE 1 after a 184 days minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor '<'vas last subcritical for> 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

North Anna Units 1 and 2 3.4.16-2 Amendments 231/212 I

Res Specific Activity 3.4.16 300 THIS FIGURE FOR ILLUSTRATION ONLY.

DO NOT USE FOR OPERATION.

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Reactor Coolant DOSE EQU IVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER North Anna Units 1 and 2 3.4.16-3 Amendments 2J-U.2~ I

Serial No. 08-0729 Docket Nos. 50-338/339 Res Specific Activity Attachment 3 Proposed Technical Specifications Pages North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Definitions 1.1 1.1 Definitions CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL TEST A COT shall be the injection of a simulated or (COT) actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using the thyroid dose conversion factors listed in Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

North Anna Units 1 and 2 1.1-2 Amendments

Definitions 1.1 1.1 Definitions DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" or the average gamma disintegration energies as provided in ICRP Publication 38, "Radionuclide Transformations."

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (continued)

North Anna Units 1 and 2 1.1-3 Amendments

Definitions 1.1 1.1 Definitions LEAKAGE 2. LEAKAGE into the containment atmosphere from (continued) sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to anyone inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

OPERABLE-OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, (continued)

North Anna Units 1 and 2 1.1-4 Amendments

Definitions 1.1 1.1 Definitions OPERABLE-OPERABILITY component, or device to perform its specified (continued) safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper RATIO (QPTR) excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2893 MWt.

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval (RTS) RESPONSE TIME from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

(continued)

North Anna Units 1 and 2 1.1-5 Amendments

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) a. All rod cluster control assemblies (RCCAs) are (continued) fully inserted except for the single RCCA ~f highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip OPERATIONAL TEST (TADOT) actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy.

The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

North Anna Units 1 and 2 1.1-6 Amendments

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT 1-131 ------------NOTE------------

not within limit. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1-131 ~ 60 ~Ci/gm.

AND A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT 1-131 to within limit.

B. DOSE EQUIVALENT XE-133 ------------NOTE------------

not within limit. LCO 3.0.4.c is applicable.

B.1 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT XE-133 to within limit.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B not met.

OR AND DOSE EQUIVALENT 1-131 C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

> 60 ~Ci/gm.

North Anna Units 1 and 2 3.4.16-1 Amendments

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 -------------------NOTE--------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE- 7 days 133 specific activity ~ 197 ~Ci/gm.

SR 3.4.16.2 -------------------NOTE--------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 14 days 1-131 specific activity ~ 1.0 ~Ci/gm.

AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of 2 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period North Anna Units 1 and 2 3.4.16-2 Amendments

Serial No. 08-0729 Docket Nos. 50-338/339 Res Specific Activity Attachment 4 Associated Bases Changes (for information only)

North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Res Specific Activity B 3.4.16 Bases changes are for information only B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant.

The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in the Standard Review Plan (Ref. 2).

APPLICABLE The LCO limits on the specific activity of the reactor SAFETY ANALYSES coolant ensure that the resulting offsite and control room doses meet the appropriate SRP acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 ~Ci/gm DOSE EQUIVALENT 1-131 from LCO 3.7.18, "Secondary Specific Activity."

The analyses for the SLB and SGTR accidents establish the acceptance limits for RCS specific activity. Reference to these analyses is used to assess changes to the unit that could affect RCS specific activity, as they related to the acceptance limits.

(continued)

North Anna Units 1 and 2 B 3.4.16-1 Revision

Res Specific Activity Bases changes are for information only B 3.4.16 BASES APPLICABLE The safety analyses consider two cases of reactor coolant SAFETY ANALYSES iodine specific activity. One cases assumes specific (continued) activity at 1.0 ~Ci/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively.

The second case assumes the initial reactor coolant iodine activity at 60.0 ~Ci/gm DOSE EQUIVALENT 1-131 due to an iodine spike cause by a reactor or an RCS transient prior to the accident. In both cases, the noble gas specific activity is assumed to be 197 ~Ci/gm DOSE EQUIVALENT XE-33 The SGTR analysis also assumes a loss of offsite power at the same time as the reactor trip. The SGTR cause a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature ~T signal.

The loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the Residual Heat Removal (RHR) system is placed in service.

The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. Reactor trip occurs after the generation of an SI signal on low steam line pressure. The affected SG blows downs completely and steam is vented directly to the atmosphere. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the RHR system is placed in service.

Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed 60.0 ~Ci/gm for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

North Anna Units 1 and 2 B 3.4.16-2 Revision

Res Specific Activity Bases changes are for information only B 3.4.16 BASES LCO The iodine specific activity in the reactor coolant is limited to [1.0] ~Ci/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 197 ~Ci/gm DOSE EQUIVALENT XE-133. The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).

The SLB and SGTR accident analyses (Refs. 3 and 4) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a SLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2).

APPLICABILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 is necessary to limit the potential consequences of a SLB or SGTR to within the SRP acceptance criteria (Ref. 2).

In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal.

Therefore, the monitoring of RCS specific activity is not required.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the specific activity is ~ 60.0 ~Ci/gm. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample.

Sampling is continued every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Actions A.l and A.2 while the DOSE EQUIVALENT 1-131 LCO limit is not met. This allowance is acceptable due (continued)

North Anna Units 1 and 2 B 3.4.16-3 Revision

Res Specific Activity Bases changes are for information only B 3.4.16 BASES ACTIONS to the significant conservatism incorporated into the (continued) specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

B.1 With the DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits that the use of the provisions of LCO 3.0.4.c.

This allowance permits entry into the applicable MODE(S),

relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

C.1 and C.2 If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is > 60.0 ~Ci/gm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner an without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measurement is the sum of the degassed gamma activities and the gaseous gamma (continued)

North Anna Units 1 and 2 B 3.4.16-4 Revision

Res Specific Activity Bases changes are for information only B 3.4.16 BASES SURVEILLANCE activities in the sample taken. This Surveillance provides REQUIREMENTS an indication of any increase in the noble gas specific (continued) activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The 7 day Frequency considers the low probability of a gross fuel failure during this time.

Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes within similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.16.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

A Note modifies the SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

SR 3.4.16.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days.

The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change

> 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the lodine levels peak during this time following the iodine spike initiation; samples at other times would provide accurate results.

The Note modifies this SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performing the SR.

This allows the Surveillance to be performed in those MODES.

prior to entering MODE 1.

North Anna Units 1 and 2 B 3.4.16-5 Revision

RCS Specific Activity B 3.4.16 Bases changes are for information only BASES REFERENCES 1. 10 CFR 50.67.

2. Standard Review Plan (SRP) Section 15.0.1 "Radiological Consequence Analyses Using Alternative Source Terms."

3. UFSAR, Section 15.4.2.

4. UFSAR, Section 15.4.3.

North Anna Units 1 and 2 B 3.4.16-6

ML083530982

Contents

Text

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 REFERENCES

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ML083530982

Text

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 REFERENCES

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