Text

Application for Amendment to Facility Operating License NPF-30, Adoption of Industry Traveler TSTF-490
	ML061430309
Person / Time
Site:	Callaway
Issue date:	05/09/2006
From:	Keith Young; AmerenUE, Union Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	ULNRC-05281
	Download: ML061430309 (53)
	v • d • e

4, Union Electric PO Box 620 Callaway Plant Fulton, MO 65251 May 9,2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 ULNRC-05281 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNION ELECTRIC CO.

APPLICATION FOR AMENDMENT TO WAMems FACILITY OPERATING LICENSE NPF-30 YE ADOPTION OF INDUSTRY TRAVELER TSTF-490 AmerenUE herewith transmits an application for amendment to Facility Operating License Number NPF-30 for the Callaway Plant. The proposed license amendment request (LAR) would revise Technical Specification (TS) 1.1, "Definitions," and TS 3.4.16, "RCS Specific Activity."

The LAR proposes to replace the current 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 (corresponding to the Xenon-133 isotope) that would replace the current E - AVERAGE DISINTEGRATION ENERGY definition.

In addition, the current DOSE EQUIVALENT I-131 definition (corresponding to the Iodine-131 isotope) would be revised to allow the use of alternate, NRC-approved thyroid dose conversion factors.

This change is being proposed in order to implement an RCS specific activity Limiting Condition for Operation (LCO) that reflects the whole body radiological consequence analysis assumptions. Those assumptions are sensitive to the noble gas activity in the primary coolant, but not to the other, non-gaseous activity currently captured in the E definition. The E definition includes radioisotopes that decay by the emission of both gamma and beta radiation. Current Condition B of LCO 3.4.16 would rarely, if ever, be entered for exceeding 1OOIE since that value is very high (the denominator is very low) if beta emitters such as H-3 (tritium) and Fluorine- 18 (F- 18) are included in that value, as required by the £ definition. In addition, SR 3.4.16.1 requires the measurement of the degassed gamma activities and the gaseous gamma activities in the sample taken for the surveillance, resulting in a questionable AoD(

a subsidiary ofAmeren Corporation

ULNRC-05281 May 9,2006 Page 2 determination of operability when the result is compared to I 00/E with its beta-emitting isotopes. This has led to confusion over what to do with the beta-emitters when performing SR 3.4.16.1 and deciding whether Condition B entry is required.

Satisfying LCO 3.4.16 should be incumbent upon satisfying the radiological consequence analysis assumptions, something that is not attained with the current construct of the LCO.

AmerenUE is submitting this LAR in conjunction with an industry consortium of six plants as a result of a mutual agreement known as Strategic Teaming and Resource Sharing (STARS). The STARS group consists of the six plants operated by AmerenUE, Wolf Creek Nuclear Operating Corporation, TXU Power, Pacific Gas and Electric Company (PG&E), STP Nuclear Operating Company, and Arizona Public Service Company.

PG&E's Diablo Canyon Power Plant is the lead STARS plant for this amendment request and submitted an amendment application on January 25, 2006.

Wolf Creek Nuclear Operating Corporation (on October 27, 2005) and TXU Power (on February 21, 2006) have also submitted license amendment requests similar to this one. These license amendment requests contain plant-specific information presented within brackets (i.e., within [ ]) in Attachment 1 (other than TS LCO numbers which vary between the Standard TS of NUREG-0452, NUREG-1431, and NUREG-1432). All other Attachments are plant- specific in nature.

The TS developed for the Westinghouse AP600 and AP1000 advanced reactor designs utilized an LCO for RCS DOSE EQUIVALENT XE-133 specific activity in place of the LCO on gross specific activity based on E. This approach was approved by the NRC for the AP600 in NUREG- 1512, "Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003," dated August 1998 and for the API000 in the NRC letter to Westinghouse Electric Company dated September 13, 2004. This license amendment request is based on TSTF-490 which is currently under NRC staff review.

Attachments 1 through 6 provide the Evaluation, Markup of Technical Specifications, Retyped Technical Specifications, Proposed Technical Specification Bases Changes, Draft FSAR Changes, and Summary of Regulatory Commitments, respectively, in support of this amendment request. Attachments 4 and 5 are provided for information only. Final TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specifications Bases Control Program," at the time the amendment is implemented. Final FSAR changes will be implemented after this amendment is approved, subject to the updating requirements of 10 CFR 50.71(e). A revision to the Fuel Clad Degradation Emergency Action Levels that reflects the new TS 3.4.16 limits will be made at the time this amendment is implemented.

4-ULNRC-05281 May 9, 2006 Page 3 It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The Callaway Onsite Review Committee and a subcommittee of the Nuclear Safety Review Board have reviewed and approved the attached licensing evaluations and have approved the submittal of this amendment application.

The changes in this LAR are not required to address an immediate safety concern. AmerenUE requests approval of this LAR prior to February 1, 2007.

AmerenUE further requests that the license amendment be made effective upon NRC issuance, to be implemented within 90 days from the date of issuance. In accordance with 10 CFR 50.91, a copy of this amendment application is being provided to the designated Missouri State official.

If you have any questions on this amendment application, please contact me at (573) 676-8659, or Mr. Dave Shafer at (314) 554-3104.

I declare under penalty of perjury that the foregoing is true and correct.

Very truly yours, Executed on:7 9, ?4(?v Keith D. Young Manager, Regulatory Affairs Attachments 1 - Evaluation 2 - Markup of Technical Specifications 3 - Retyped Technical Specifications 4 - Proposed Technical Specification Bases Changes 5 - Draft FSAR Changes 6 - Summary of Regulatory Commitments

ULNRC-05281 May 9, 2006 Page 4 cc:

U.S. Nuclear Regulatory Commission (Original and 1 copy)

Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 7E1 Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street P.O. Box 360 Jefferson City, MO 65102-0360 Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Page 1 of 19 EVALUATION

1. DESCRIPTION Page 2

2. PROPOSED CHANGES Page 2

3. BACKGROUND Page 5

4. TECHNICAL ANALYSIS Page 8

5. REGULATORY SAFETY ANALYSIS Page 15 5.1 NO SIGNIFICANT HAZARDS CONSIDERATION Page 15 5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA Page 17

6. ENVIRONMENTAL CONSIDERATION Page 18

7. REFERENCES Page 18

Attachment 1 Page 2 of 19 EVALUATION

1.0 DESCRIPTION

The proposed amendment would revise Technical Specification (TS) 1.1, "Definitions,"

and TS 3.4.16, "RCS Specific Activity." The proposed changes would replace the current 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 (DEX) definition that would replace the current E - AVERAGE DISINTEGRATION ENERGY definition. In addition, the current DOSE EQUIVALENT I-131 (DEI) definition would be revised to allow alternate, NRC-approved thyroid dose conversion factors.

2.0 PROPOSED CHANGE

S The TS Section 1.1 definition for DEI would be revised from:

"DOSE EQUIVALENT I-13 1 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, I-132, 1-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites" or those derived from the data provided in International Commission on Radiological Protection Publication 30, "Limits for Intakes of Radionuclides by Workers," 1979.]

to "DOSE EQUIVALENT I-13 1 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, 1-133, 1-134, and I-135 actually present.

The determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from:

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

2) Table E-7 of Regulatory Guide 1.109, Revision 1,NRC, 1977, or

3) International Commission on Radiological Protection (ICRP) Publication 30, "Limits for Intakes of Radionuclides by Workers," Supplement to Part 1, pages 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," 1979, or

4) Table 2.1 of EPA Federal Guidance Report No. 11, EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.]"

Attachment 1 Page 3 of 19 The TS Section 1.1 definition for E - AVERAGE DISINTEGRATION ENERGY would be deleted and replaced with a new definition for DEX which states:

"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-87, Kr-88, 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 III.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil",

1993.]

TS Limiting Condition for Operation (LCO) 3.4.16, "RCS Specific Activity," would be revised from:

"The specific activity of the reactor coolant shall be within limits."

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

The current TS Figure 3.4.16-1, "Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER" would be deleted.

The Applicability of TS 3.4.16 would be revised from:

"MODES 1 and 2, MODE 3 with RCS average temperature (Tayg) > 500'F."

to "MODES 1, 2,3, and 4."

TS 3.4.16 Condition A would be revised from:

"DOSE EQUIVALENT I-131 > 1.0 piCi/gm."

to "DOSE EQUIVALENT 1-131 not within limit."

TS 3.4.16 Required Action A.1 would be revised from:

"Verify DOSE EQUIVALENT I-131 within the acceptable region of Figure 3.4.16-1."

to "Verify DOSE EQUIVALENT 1-131 < 60 IiCi/gm."

[I]

2 Attachment 1 Page 4 of 19 TS 3.4.16 Condition B would be revised from:

"Gross specific activity of the reactor coolant > 100/ E £Ci/gm.

to "DOSE EQUIVALENT XE-133 not within limit."

TS 3.4.16 Required Action B.1 would be revised from:

"Be in MODE 3 with Tayg < 500'F."

to

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

LCO 3.0.4.c is applicable.

Restore DOSE EQUIVALENT XE-133 to within limit."

TS 3.4.16 Required Action B.1 Completion Time would be revised 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 ."

TS 3.4.16 Condition C would be revised from:

"Required Action and associated Completion Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1."

to "Required Action and associated Completion Time of Condition A or B not met.

OR DOSE EQUIVALENT I-131 > 60 gtCi/gm."

TS 3.4.16 Required Action(s) for Condition C would be revised from:

"C.1 Be in MODE 3 with Tavg < 500'F."

to "C.1 BeinMODE3."

AND C.2 Be in MODE 5."

TS 3.4.16 Condition C would be revised to add a Completion Time for new Required Action C.2 of "36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ." The Completion Time for Required Action C. 1 would remain 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

-2 Attachment 1 Page 5 of 19 Surveillance Requirement (SR) 3.4.16.1 would be revised from:

"Verify reactor coolant gross specific activity <100/ EjpCi/gm."

to t--------------------------------------NOTE---------------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity < [225]

PCi/gm."

Current SR 3.4.16.3 would be deleted.

In summary, the proposed changes will revise the definition of DEI, delete the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition for DEX, revise TS 3.4.16 to specify an LCO limit on DEI, add a new LCO 3.4.16 limit for DEX, increase the Completion Time for Required Action B. 1, delete TS Figure 3.4.16-1, and revise the Conditions and Required Actions accordingly. Also, the Applicability of LCO 3.4.16 is extended to reflect the MODES during which pertinent accidents (SGTR or MSLB) could be postulated to occur, SR 3.4.16.1 is revised to verify DEX prior to MODE 1 entry, and SR 3.4.16.3 is deleted.

The TS Bases for LCO 3.4.16 would be revised to expand upon the proposed changes.

The TS Bases changes are included for information only.

[Attachments 2 and 3 provide the TS markups reflecting the above changes and the retyped TS. Attachment 4 provides an information-only copy of the associated TS Bases changes. Attachment 5 provides an information-only copy of related FSAR changes.

Attachment 6 lists the regulatory commitments associated with this amendment application.]

3.0 BACKGROUND

3.1 Radiological Consequence Analyses Radiological consequence analyses are performed for the Steam Generator Tube Rupture (SGTR) accident and for the Main Steam Line Break (MSLB) accident since these events involve the release of primary coolant activity. For events that also result in fuel damage (such as locked rotor, rod ejection, and loss-of-coolant accident) as a result of the accident, the dose contribution from the initial activity in the RCS is insignificant. The maximum dose to the whole body and the thyroid 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 100.11. The limits on RCS specific activity ensure that the offsite doses are appropriately limited as required by NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review

-I Attachment 1 Page 6 of 19 Plan," Section 15.1.5, "Steam System Piping Failures Inside and Outside of Containment (PWR)," Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment," Revision 2, for MSLB accidents and NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan," Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure (PWR)," Revision 2, for SGTR accidents.

The maximum dose to the whole body, or its equivalent to any part of the body, that an individual can receive in the plant control room for the duration of an accident is specified in General Design Criterion 19 (GDC 19) contained in Appendix A to 10 CFR

50. The limits on RCS specific activity ensure that the doses are less than the GDC 19 limits during analyzed transients and accidents, as required by NUREG-0800, "U.S.

Nuclear Regulatory Commission Standard Review Plan," Section 6.4, "Control Room Habitability System," Revision 2, and Regulatory Position C.4.5 of NRC Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors."

The SGTR and MSLB radiological consequence analyses establish the acceptance limits for the TS 3.4.16 RCS specific activity. These analyses consider two cases of RCS iodine specific activity. Case 1 assumes that an accident-initiated iodine spike occurs which results in an increase in the rate of iodine release from the fuel rods containing cladding defects to the primary coolant immediately after an MSLB or SGTR. Case 2 assumes that a pre-accident iodine spike occurs due to a transient prior to the MSLB or SGTR. The results of the SGTR radiological consequence analyses are described in

[FSAR Section 15.6.3]. The results of the MSLB radiological consequence analyses are described in [FSAR Section 15.1.5].

[The Case 1 radiological consequence analyses for SGTR and MSLB assume that the initial reactor coolant iodine specific activity corresponds to an isotope mixture that bounds the SR 3.4.16.2 limit for both tight and open fuel defects. The isotopic mix is based on the initial RCS concentrations from FSAR Table 1SA-5. This table provides conservative values for the iodine isotopic spectrum that bound the RCS concentrations which could be expected with either tight or open fuel defects. Since the assumed iodine spectrum represents bounding values for different types of fuel defects, the initial radioiodine inventory exceeds (by approximately 6% in the conservative direction) the SR 3.4.16.2 limit of 1.0 VlCi/gm.]

This analysis assumption provides the basis for the iodine specific activity limit of 1.0 pCi/gm contained in current TS 3.4.16 Condition A and SR 3.4.16.2. Thyroid dose conversion factors based on [Table E-7 of NRC Regulatory Guide 1.109, Revision 1, 1977 or International Commission on Radiological Protection (ICRP) Publication 30, 1979, have been used in radiological consequence analyses performed to date; however, only ICRP 30 was used in support of the current MSLB and SGTR radiological consequence analyses which form the technical basis for the iodine specific activity limit in LCO 3.4.16.] Any of the NRC-approved thyroid dose conversion factor references

Attachment 1 Page 7 of 19 cited in the revised definition of DOSE EQUIVALENT I-13 1 may be used in future analyses after this amendment is approved.

Case 1 also assumes an accident-initiated iodine spike that increases the rate of iodine release from the fuel rods containing cladding defects to the primary coolant immediately after an MSLB or SGTR. The iodine spiking factor is assumed to be [500 for the Case 1 radiological consequence evaluation for MSLB and 335 for the Case 1 radiological consequence evaluation for both SGTR radiological consequence analyses].

[The Case 2 radiological consequence analyses for SGTR and MSLB assume the initial reactor coolant iodine specific activity is a factor of 60 higher than Case I due to a pre-accident iodine spike caused by a transient prior the accident.] This [bounds] the allowable RCS specific activity value of 60 pCi/gm contained in current TS Figure 3.4.16-1 for RATED THERMAL POWER (RTP) between 80% and 100%. [Current] TS Figure 3.4.16-1 [which is being deleted by this amendment] provides DEI concentration limits during short periods in which iodine spiking may occur due to a power transient.

In both Case 1 and Case 2 radiological consequence evaluations for SGTR and MSLB, the noble gas specific activity in the reactor coolant is assumed to be [approximately 3%

higher than the proposed 225] pCi/gm DEX limit. The dose analysis assumptions are discussed further in [FSAR Tables 15.1-3 and 15.6-4]. The initial DEX concentrations were calculated assuming [1% fuel defects] and using [whole body dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93 -081, "External Exposure to Radionuclides in Air, Water, and Soil",

1993].

3.2 RCS Specific Activity The RCS specific activity level is used in design basis accident analyses to determine the thyroid and whole body radiological consequences of accidents that involve the release of RCS activity. For events that also include fuel damage, the dose contribution from the initial activity in the RCS is insignificant.

The current definition for DEI is based on thyroid dose conversion factors and reflects a licensing model in which the radiological consequences of iodine releases for accidents are reported as thyroid and whole body doses. [Two] additional NRC-approved source[s]

of thyroid dose conversion factors [are] being added to the revised definition.

LCO 3.4.16 specifies the limit for RCS gross specific activity as 100/E pCi/gm. "E is defined as:

"E shall be the average (weighted in proportion to the 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, with half lives > [15] minutes, making up at least 95% of the total non-iodine activity in the coolant."

Attachment 1 Page 8 of 19 In performing accident dose analyses in which primary coolant is released, the concentration of noble gas activity in the coolant is assumed to be that level associated with [1% fuel defects], which closely approximates the TS 3.4.16 limit of 100/E pCi/gm under accident conditions.

LCO 3.4.16 specifies a limit for RCS iodine concentration during equilibrium operation.

In recognition of the potential for exceeding the equilibrium iodine concentration due to iodine spiking following power transients, the LCO also permits the equilibrium value to be exceeded for a period of less than or equal to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . As currently presented, the value for the maximum allowable iodine concentration during the 48-hour period of elevated activity is a function of power level as provided in TS Figure 3.4.16-1. In accordance with the figure, as power is reduced below 80% RTP, the allowable RCS iodine concentration increases from 60 pCi/gm DEI to as high as [280] ACi/gm DEI at

[25%] RTP. Below [25%] RTP, no further increase is defined.

The curve contained in TS Figure 3.4.16-1 was initiated by the Atomic Energy Commission (AEC) in a June 12, 1974 letter from the AEC on the subject, "Proposed Standard Technical Specifications for Primary Coolant Activity." This letter does not provide any technical basis for the curve.

3.3 Purpose for Proposed Amendments The addition of the new DEX limit and TS 3.4.16 changes are being proposed in order to implement an RCS specific activity LCO that better reflects the whole body radiological consequence analyses which are sensitive to the noble gas activity in the primary coolant but not to the other, non-gaseous activity currently captured in the E definition. The E definition includes radioisotopes that decay by the emission of both gamma and beta radiation. Current Condition B of LCO 3.4.16 would rarely, if ever, be entered for exceeding 100/E since that value is very high (the denominator is very low) if beta emitters such as H-3 (tritium) and Fluorine- 18 (F- 18) are included in that value, as required by the F definition. [In addition, SR 3.4.16.1 requires the measurement of the degassed gamma activities and the gaseous gamma activities in the sample taken for the surveillance, resulting in a questionable determination of operability when the result is compared to 100/E with its beta-emitting isotopes. This has led to confusion over what to do with the beta-emitters when performing SR 3.4.16.1 and deciding whether Condition B entry is required. Satisfying LCO 3.4.16 should be incumbent upon satisfying the radiological consequence analysis assumptions, something that is not attained with the current construct of the LCO.]

4.0 TECHNICAL ANALYSIS

4.1 TS Changes

Attachment 1 Page 9 of 19 Revision to Definition of DEI The current TS 1.1 definition for DEI is revised to add new reference[s] for acceptable thyroid dose conversion factors. Also, the word "thyroid" is deleted from the first sentence.

New thyroid dose conversion factor reference[s are] added to the definition. The new reference[s are] "Table 2.1 of EPA Federal Guidance Report No. 11, EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988 [and Table E-7 of Regulatory Guide 1.109, Revision 1,NRC, 1977]. EPA Federal Guidance Report No. 11 is referenced in Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003, Section C, "Regulatory Position," Subsection 4, "Dose Calculational Methodology," Subsection 4.1, "Offsite Dose Consequences," assumption 4.1.2 as acceptable for determining thyroid dose from inhalation. [The thyroid dose conversion factor values contained in Table 2.1 of EPA Federal Guidance Report No. 11 are provided to three significant digits. The thyroid dose conversion factor values contained in ICRP Publication 30, Supplement to Part 1, pages 192-212 are the same as those listed in Table 2.1 of EPA Federal Guidance Report No. 11 when the EPA Federal Guidance Report No. 11 values are rounded to two significant digits.]

The deletion of the word "thyroid" from the first sentence is an editorial change only.

Deletion of Definition for E - AVERAGE DISINTEGRATION ENERGY and Addition of New Definition for DEX The current TS 1.1 Definition for E - AVERAGE DISINTEGRATION ENERGY is deleted and replaced with a new Definition for DEX.

When E is determined using a design basis approach in which it is assumed that 1% of the power is generated by fuel rods having cladding defects and there is no removal of fission gases from the RCS letdown flow, the value of E is dominated by the Xe-133 isotope. The other nuclides have relatively small contributions. However, during normal plant operation there is typically only a small amount of fuel defects and the radioactive nuclide inventory can become dominated by tritium and corrosion and/or activation products, resulting in the determination of a value of E that is very different than that which would be calculated using the design basis approach. Therefore, the radiological consequence analyses for accidents become disconnected from normal plant operation and the current TS 3.4.16 limit on gross specific activity is not relevant. The use of E also results in a TS limit that can vary during operation as different values for E are determined, resulting in different values for the gross specific activity limit (1 00/ pCi/gr).

Attachment 1 Page 10 of 19 Additionally, since the concern associated with the RCS noble gas activity is the acute whole body dose that the operators and the general public might receive in the event of a postulated accident, the manner in which E is calculated gives undue importance to nuclides that are primarily beta radiation emitters. Beta radiation will contribute to a skin dose, but not to the whole body dose. Dose limits for the general population do not include consideration of the beta skin dose.

Therefore the deletion of the current TS 1.1 Definition for E - AVERAGE DISINTEGRATION ENERGY and addition of a new definition for DEX will result in TS 3.4.16 requirements for RCS specific activity which are consistent with the assumptions contained in the radiological consequence analyses.

The new definition for DEX is similar to the definition for DEL. The determination of DEX will be performed in a similar manner to that currently used in determining DEI, except that the calculation of DEX is based on the acute dose to the whole body and considers the noble gases [Kr-85m, Kr-87, Kr-88, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138] which are significant in terms of contribution to whole body dose. [Some noble gas isotopes are not included due to low initial RCS concentrations, short half-lives, or small dose conversion factors. The excluded isotopes [Kr-83m, Kr-85, Kr-89, Xe-131m, and Xe-137] contribute less than 2.3% of the whole body dose from noble gases in the MSLB and SGTR radiological consequence analyses that form the technical basis for the noble gas specific activity limit in LCO 3.4.16. The DEX limit has been lowered to accommodate the exclusion of these five isotopes.] If a specific noble gas nuclide is not detected, the new definition states that it should be assumed the nuclide is present at the minimum detectable activity. This will result in a conservative calculation of DEX.

The new definition of DEX states that the determination of DEX shall be performed using the effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, EPA402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil," 1993. [ ] These dose conversion factors are applicable for determination of DEX. The use of the dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12 is endorsed by Regulatory Guide 1.195, Subsection 4.1, assumption 4.1.4 as acceptable for determining whole body doses because of the uniform body exposure associated with semi-infinite cloud dose modeling.

TS 3.4.16 LCO Revision The TS 3.4.16 LCO is modified to specify that the iodine specific activity in terms of DEI and noble gas activity in terms of DEX shall be within limits.

Currently TS 3.4.16 states that the specific activity of the reactor coolant shall be within limits. The limits are currently not explicitly identified in the LCO, but are instead defined in current Condition B and SR 3.4.16.1 for gross specific activity and in current

f Attachment 1 Page 11 of 19 Condition A and SR 3.4.16.2 for iodine specific activity.

The proposed change states "RCS DOSE EQUIVALENT I-13 1 and DOSE EQUIVALENT XE-133 specific activity shall be within limits." The DEI limit of < 1.0 jtCi/gm, contained in current Condition A and SR 3.4.16.2, will now be listed only in SR 3.4.16.2. In addition, the limit of 1.0 piCi/gm is [conservative with respect to] the current SGTR and MSLB radiological consequence analyses discussed in Section 3.1 above.

The DEX limit of [225] piCi/gm contained in revised SR 3.4.16.1 is [conservative with respect to the current SGTR and MSLB radiological consequences discussed in Section 3.1 above and has been lowered to accommodate the exclusion of five noble gas isotopes due to low initial RCS concentrations, short half-lives, or small dose conversion factors].

The primary purpose of the TS 3.4.16 LCO on RCS specific activity is to support the dose analyses for design basis accidents. Whole body doses are primarily dependent on the noble gas concentrations, not the non-gaseous activity currently captured in the E definition. It is appropriate to have the TS 3.4.16 LCO apply to the noble gas specific activity in the RCS. Thus, it is acceptable that the current TS 3.4.16 limit on gross specific activity can be replaced by an LCO limit based on RCS noble gas specific activity in the form of DEX. The limit on the amount of noble gas activity in the RCS remains consistent with the design basis accident radiological consequence analyses and would not fluctuate with variations in the calculated value of E during normal operation as is currently the case.

TS 3.4.16 Avplicability Revision The TS 3.4.16 Applicability is modified to include all of MODE 3 and MODE 4. It is necessary for the LCO to apply during all of MODES 1 through 4 to limit the potential radiological consequences of an SGTR or MSLB that may occur during these MODES.

In MODES 5 and 6, the steam generators are not 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 during MODES 5 and 6 is not required.

TS 3.4.16 Condition A Revision TS 3.4.16 Condition A is revised by replacing the limit "> 1.0 j+/-Ci/gmr" with the words "not within limit" to be consistent with the Revised TS 3.4.16 LCO format. The DEI limit of < 1.0 JlCi/gm is contained in SR 3.4.16.2. [1 TS 3.4.16 Required Action[1 A.1 [ 1 Revision TS 3.4.16 Required Action A.1 is modified to remove the reference to Figure 3.4.16-1 and insert a limit of less than or equal to 60 ttCi/gm for DEI.

Attachment 1 Page 12 of 19 The curve contained in Figure 3.4.16-1 was initiated by the AEC in a June 12, 1974 letter from the AEC on the subject, "Proposed Standard Technical Specifications for Primary Coolant Activity." However, this letter does not provide any technical basis for the curve.

The Case 2 radiological consequence analyses for SGTR and MSLB accidents that take into account the pre-accident iodine spike do not consider the elevated RCS iodine specific activities permitted by current TS Figure 3.4.16-1 for operation at power levels below 80% RTP (i.e. DEI of 60 pCi/gm at 80% RTP increasing linearly to [225] PCi/gm at [25%] RTP). Instead, the Case 2 analyses assume a DEI concentration 60 times higher than the corresponding accident's Case 1 analysis assumption [,which corresponds to the 60 pCi/gm specific activity limit associated with 100% RTP operation as discussed in Section 3.1 above]. Therefore, TS 3.4.16 Required Action A.1 should be based on a limit of 60 pCi/gm to be consistent with the assumptions contained in the radiological consequence analyses. It is not expected that plant operation at reduced power levels would result in iodine specific activity levels that exceed the 60 pCi/gm upper limit defined for full power operation.

[I TS 3.4.16 Condition B Revision to Include Action for DEX Limit Current TS 3.4.16 Condition B is replaced with a new Condition B for DEX not within limit. This change is made to be consistent with the change to the TS 3.4.16 LCO which requires the DEX specific activity to be within limit as discussed above. The DEX limit of [225] piCi/gm is contained in revised SR 3.4.16.1 and is [conservative with respect to the current SGTR and MSLB radiological consequences discussed in Section 3.1 above.]

The primary purpose of the TS 3.4.16 LCO on RCS specific activity and its associated Conditions is to support the dose analyses for design basis accidents. The whole body dose is primarily dependent on the noble gas activity, not the non-gaseous activity currently captured in the E definition and limited by current TS 3.4.16 Condition B.

The Completion Time for revised TS 3.4.16 Required Action B.1 will require restoration of DEX to within limit in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This is consistent with the Completion Time for current Required Action A.2 for DEL. [Since the radiological consequences reported for SGTR and MSLB in [FSAR Tables 15.6-5, 15.6-SA, and 15.1-4 at Callaway]

demonstrate that thyroid doses are a greater percentage of the applicable SRP acceptance criteria than whole body doses, it then follows that the Completion Time for noble gas activity being out of specification in revised Required Action B. 1 should be at least as great as the Completion Time for iodine specific activity being out of specification in current Required Action A.2.] The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for revised Required Action B.1 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 an MSLB or SGTR occurring during this time period.

Attachment 1 Page 13 of 19 A Note is added which states that LCO 3.0.4.c is applicable. This is consistent with the Note applicable to current Required Actions A. 1 and A.2 for DEL. This Note permits entry into the applicable MODE(s), relying on Required Action B.1 while the DEX LCO limit is not met. This MODE change allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event that 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.

TS 3.4.16 Condition C Revision TS 3.4.16 Condition C is revised to include Condition B if the Required Action and associated Completion Time of Condition B is not met. This is consistent with the changes made to Condition B which will no longer specify a shutdown track. Condition C is also revised to replace the limit on DEI from Figure 3.4.16-1 with a value of > 60 jLCi/gm. This change makes Condition C consistent with the changes made to TS 3.4.16 Required Action A. 1.

TS 3.4.16 Required Action C.1 is changed to require the plant to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and a new Required Action C.2 is added which requires the plant to be in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . These changes are consistent with the changes made to the TS 3.4.16 Applicability. The revised LCO is applicable throughout all of MODES 1 through 4 to limit the potential radiological consequences of an SGTR or MSLB that may occur during these MODES. Therefore, Condition C needs to default to a MODE 5 end state for TS 3.4.16 to no longer be applicable.

A new TS 3.4.16 Required Action C.2 Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is added for the plant to reach MODE 5. This Completion Time is reasonable, based on operating experience, to reach MODE 5 from full power conditions in an orderly manner and without challenging plant systems. The Completion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is consistent with other TS which specify a Completion Time to reach MODE 5.

SR 3.4.16.1 Revision to Include Surveillance for DEX The current SR 3.4.16.1 surveillance for RCS gross specific activity is deleted and replaced with a surveillance to verify that the reactor coolant DEX specific activity is

< [225] jiCi/gm. This change provides a surveillance for the new LCO limit added to TS 3.4.16 for DEX.

The revised SR 3.4.16.1 surveillance 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 activities in the sample taken. The surveillance provides an indication of any increase in the noble gas specific activity.

S T Attachment 1 Page 14 of 19 The results of the surveillance on DEX allow proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The 7 day Frequency considers the unlikelihood of a gross fuel failure during this time.

If a specific noble gas nuclide listed in the new definition for DEX in Specification 1.1 is not detected, it should be assumed to be present at the minimum detectable activity. This is consistent with the new TS 1.1 Definition for DEX and will ensure a conservative calculation of DEX when noble gas nuclides are not detected.

The SR is modified by a NOTE which allows entry into MODE 4, MODE 3, and MODE 2 prior to performing the surveillance. This allows the surveillance to be performed in any of those MODES, prior to entering MODE 1, similar to the current surveillance SR 3.4.16.2 for DEL.

SR 3.4.16.3 Deletion Current SR 3.4.16.3 is deleted. The TS 3.4.16 LCO on RCS specific activity supports the dose analyses for design basis accidents, in which the whole body dose is primarily dependent on the noble gas concentration, not the non-gaseous activity currently captured in the E definition. Therefore, with the elimination of the limit for RCS gross specific activity and the addition of the new LCO limit for noble gas specific activity, this SR to determine E is no longer required.

4.2 Impact on Radiological Consequence Analyses The proposed changes do not impact the radiological consequences of any design basis accident. Replacing the limit on E with a limit on DEX based on the value used in the current radiological consequence analyses will limit the RCS noble gas concentrations to values which are consistent with the radiological consequence analyses for those noble gases which are significant in terms of contribution to dose. These changes will also limit any potential RCS iodine specific activity excursion to the value currently associated with full power operation (i.e. 60 pCi/gm DEI). This concentration is more restrictive on plant operation than the current LCO which allows operation up to [225]

igCi/gm DEI as indicated in Figure 3.4.16-1. The proposed changes eliminate the potential for radiological consequences of a postulated accident to exceed those previously calculated.

4.3 SmmryM In summary, the proposed changes will revise the definition of DOSE EQUIVALENT I-13 1, delete the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition for DOSE EQUIVALENT XE-133, revise TS 3.4.16 to specify an LCO limit on DOSE EQUIVALENT I-13 1, add a new LCO limit to TS 3.4.16 for DOSE EQUIVALENT XE-133, increase the Completion Time for Required Action B.1, delete TS Figure 3.4.16-1, and revise the TS 3.4.16 Conditions and Required Actions

Attachment 1 Page 15 of 19 accordingly. Also, the Applicability of LCO 3.4.16 is extended to reflect the MODES during which pertinent accidents (SGTR and MSLB) could be postulated to occur, SR 3.4.16.1 is revised to verify DOSE EQUIVALENT XE-133 is within the prescribed limit, and SR 3.4.16.3 is deleted.

The revised definition of DOSE EQUIVALENT I- 131 allows the use of thyroid dose conversion factors which are acceptable for determining thyroid dose. The above changes will result in TS 3.4.16 requirements for RCS specific activity which are consistent with the assumptions contained in the radiological consequence analyses. The primary purpose of the TS 3.4.16 LCO on RCS specific activity is to support the dose analyses for design basis accidents, in which the whole body dose is primarily dependent on the noble gas specific activity, not the non-gaseous activity currently captured in the E definition. The TS 3.4.16 Conditions, Required Actions, and Surveillance Requirements are revised accordingly to support the deletion of the requirements for gross specific activity based on E and the addition of the new LCO limit for DOSE EQUIVALENT XE-133. The proposed changes do not impact the radiological consequences of any design basis accident.

5.0 REGULATORY SAFETY ANALYSIS This section addresses the standards of 10 CFR 50.92 as well as the applicable regulatory requirements and acceptance criteria.

The proposed amendment would revise the definition of DOSE EQUIVALENT I-131, delete the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition for DOSE EQUIVALENT XE-133, revise TS 3.4.16 to specify an LCO limit on DOSE EQUIVALENT I-131, add a new LCO limit to TS 3.4.16 for DOSE EQUIVALENT XE- 133, increase the Completion Time for Required Action B. 1, delete TS Figure 3.4.16-1, and revise the TS 3.4.16 Conditions and Required Actions accordingly. In addition, the Applicability of LCO 3.4.16 is extended to reflect the MODES during which pertinent accidents (SGTR and MSLB) could be postulated to occur, SR 3.4.16.1 is revised to verify DOSE EQUIVALENT XE-133 is within the prescribed limit, and SR 3.4.16.3 is deleted.

5.1 No Significant Hazards Consideration (NSHC)

[AmerenUE] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," Part 50.92(c), as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No

Attachment 1 Page 16 of 19 The proposed changes would add new thyroid dose conversion factor reference[s] to the definition of DOSE EQUIVALENT I-13 1, eliminate the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition of DOSE EQUIVALENT XE-133, replace the Technical Specification (TS) 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a limit on noble gas specific activity in the form of a Limiting Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, increase the Completion Time for Required Action B. 1, replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT I-13 1, extend the Applicability of LCO 3.4.16, and make corresponding changes to TS 3.4.16 to reflect all of the above. The proposed changes are not accident initiators and have no impact on the probability of occurrence of any design basis accidents.

The proposed changes will have no impact on the consequences of a design basis accident because they will limit the RCS noble gas specific activity to be consistent with the values assumed in the radiological consequence analyses. The changes will also limit the potential RCS iodine concentration excursion to the value currently associated with full power operation, which is more restrictive on plant operation than the existing allowable RCS iodine specific activity at lower power levels.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed changes do not alter any physical part of the plant nor do they affect any plant operating parameters besides the allowable specific activity in the RCS. The changes which impact the allowable specific activity in the RCS are consistent with the assumptions assumed in the current radiological consequence analyses.

Therefore, the proposed changes do not create the possibility of a new or different accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The acceptance criteria related to the proposed changes involve the allowable control room and offsite radiological consequences following a design basis accident. The proposed changes will have no impact on the radiological consequences of a design basis accident because they will limit the RCS noble gas specific activity to be consistent with the values assumed in the radiological consequence analyses. The changes will also limit the potential RCS iodine specific activity excursion to the value currently associated with

Attachment 1 Page 17 of 19 full power operation, which is more restrictive on plant operation than the existing allowable RCS iodine specific activity at lower power levels.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Conclusion:

Based on the above evaluation, [AmerenUE] concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements / Criteria The regulatory guidance documents associated with this amendment application include:

NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan,"

Section 15.1.5, "Steam System Piping Failures Inside and Outside of Containment (PWR)," Appendix A, "Radiological Consequence of Main Steam Line Failures Outside Containment," Revision 2, identifies the thyroid and whole body offsite radiological consequence acceptance criteria for main steam line break accidents.

NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan,"

Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure (PWR)," Revision 2, identifies the thyroid and whole body offsite radiological consequence acceptance criteria for steam generator tube rupture accidents.

NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan,"

Section 6.4, "Control Room Habitability System," Revision 2, identifies the thyroid, whole body, and beta skin radiological consequence acceptance criteria for control room occupants.

Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," provides acceptable dose conversion factors, radiological consequence acceptance criteria, and other dose analysis methodology parameters.

There are no changes being proposed in this amendment application such that commitments to the regulatory guidance documents above would come into question.

The evaluations documented above confirm that [Callaway Plant] will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Page 18 of 19 Commission's regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

[AmerenUE] has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 References

1. Environmental Protection Agency (EPA) Federal Guidance Report No. 11, EPA-520/1-88-020,"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," September 1988.

2. Environmental Protection Agency (EPA) Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil,"

1993.

3. International Commission on Radiological Protection (ICRP) Publication 30, "Limits for Intakes of Radionuclides by Workers," ICRP, 1979.

4. Atomic Energy Commission (AEC) letter "Proposed Standard Technical Specifications for Primary Coolant Activity," dated June 12, 1974.

5. Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003.

6. Regulatory Guide 1.109, Revision 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Octoberl977.

7. Atomic Energy Commission (AEC) Report TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 1962.

Attachment 1 Page 19 of 19

8. NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan,"

Section 15.1.5, "Steam System Piping Failures Inside and Outside of Containment (PWR)," Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment," Revision 2, July 1981.

9. NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan,"

Section 15.6.3, "Radiological Consequences of a Steam Generator Tube Failure (PWR)," Revision 2, July 1981.

10. NUREG-0800, "U.S. Nuclear Regulatory Commission Standard Review Plan,"

Section 6.4, "Control Room Habitability System," Revision 2, July 1981.

11. NUREG-15 12, "Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003," August 1998.

12. NUREG-1431, Volume 1, Revision 3, "Standard Technical Specifications Westinghouse Plants," dated June 2004.

7.2 Precedent The Technical Specifications developed for the Westinghouse AP600 and AP1000 advanced reactor designs utilize an LCO for RCS DEX activity in place of the LCO on gross specific activity based on E. This approach was approved by the NRC for the AP600 in NUREG-1 512, "Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003," dated August 1998 and for the API000 in the NRC letter to Westinghouse Electric Company dated September 13, 2004. The curve in current TS Figure 3.4.16-1 was not included in the TS approved for the AP600 and API000 advanced reactor designs.

ATTACHMENT 2 MARKUP OF TECHNICAL SPECIFICATIONS

Definitions 1.1 1.1 Definitions (continued)

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 cycle REPORT (COLR) 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 DOlE EeUIVAELEN T -1Iv -hallbe that eeneefntrotiean9V ef 1-131 ( icrocuiesgrarn) that elnon would produceQthe same thyroid dose-as thequentity, and icotopic mixiure ef i 3i, 1-12, 1-133, I131, and 1135 actually pree Thethyroid doeo converciorn faLor useo-orinis calcul1n-shall be

-hoe listed inT- III of T:D-413-444, AEG, r9C2, "Caleulteio-of Distan- Fdcam for POwcr tand Tot Reaeter Sites" or those derived from the Jal ,vJve ;, hIItemational Qommission on R-adJulagical Pruoteiui Ponioatin 30, limits for tntaIkzz of ladi~u, ive by Wua ka ," 10i70.

- shall be the average (weighted in propoeion to the DISINTEGRA.TON EINER=GY concentrtion of cach mdionuc in the eactor coolant at t4hetime zairl~gut-euthotle-of(U Fa'orobeta and

-gamma energies per disintegration (in-k for *topes, than iodines, with half lives 15 minut, making up at Ithzr Ic4et 9556 of the total ..oniodine activity in the coolant.

ba4 .- azzyVmt 4 -An KE-iss3 /'

(continued)

CALLAWAY PLANT 1.1-2 Amendment No. 159

INSERT 1.1-2A 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 thyroid dose conversion factors from:

1) Table IlIl of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or

2) Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or

3) International Commission on Radiological Protection (ICRP) Publication 30, "Limits for Intakes of Radionuclides by Workers," Supplement to Part 1, pages 192-212, Table titled uCommitted Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," 1979, or

4) Table 2.1 of EPA Federal Guidance Report No. 11, EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988.

INSERT 1.1-2B DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (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-87, Kr-88, 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-1 33 shall be performed using the effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil", 1993.

RCS Specific Activity

3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)

LCO 3,4.16 The specific activity of the reactor coolant shall limits.

MODES 1 and 2, MODE 3 with RCS average A. DOSE EQUIVALENT

> 1.0 AiC/gm. I Verify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQUIVALENT 1-131 within the acceptable

,,region of Figure 3.4.16-1.

,DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months (continued) e /c ,7PI er . 4-5*0D CALLAWAY PLANT 3.4-40 Amendment No. 164

RCS Specific Activity 3.4.16 INSERT 3.4-40 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 RCS DOSE EQUIVALENT 1-1 31 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

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

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

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

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

< 60 pCi/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 - - NOTE ---

XE-133 not within LCO 3.0.4.c is applicable.

limit.

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.

CALLAWAY PLANT 3.4-40 Amendment No. 1xx

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION TIME C. Required Action and C.1 Be in MODE 3w*h-T-, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition itnot A/

met. Je A_

OR 8 rn1e 1,atS, 534 Aow'.

DOSE EQUIVALENT 1-131 ofFigure 3.4.1.C 1.

_h O _ C /5 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~

I _ oly orA 1:t ---- ;CW medcFf.zQ/.ot _ _

SURVEILLANC REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant wross specific latit, 7 days

-- _1001 ~6 uurvgm.cbr -/:f3 KEA$~

act;*<avow .~c;t ________.

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 ptCi/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 CALLAWAY PLANT 3.4-41 Amendment No. 133

Re

!U pecir,Activity-5.4.1 -

SU6RVE!66ANGE REeUIREMEN~TS (feontinuezJ 6YR\.'ELLANiCE FRE__UENeY-SR 3.4.1C.B -^c----------------NOTE -- --------------------------

No~airedto be performed until 31 days aftera mnm effective full power days ad2 ys of MODE 1 open have elapsed since eco was last subcritical 8 hous Determine Ef os ODE 1 after a +84-deys minimum of 2 itv ulpwrdyad 20 days of MODE Ortonhv elapsed since th ~ctor I t subcritical for > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

-- _l . _'PAN 3ALLA~AY LANTAmen;dmQ bnt 4!

13

RGS Speefie r ....

300 (25, 290) 250 I 200 UNACCEPTABLE OPERATION V4 150

'I 100

.01 ACCEPTABLE OPERATION (100, 60)

I 4 (Bus 50

/ . . . \

/

. . _I 20 30 40 50 60 70 80 90 \0\

DL.j. huw an u.aUL PI1R.

- Ul _ I g - --

ro-&- -.. -rl1..I

^*t/to__A_-!

Limit Vciuas fPe RATED TI ACRE AltHrovE

^. . A V YI A_

LANt

ATTACHMENT 3 RETYPED TECHNICAL SPECIFICATIONS

9o Definitions 1.1 1.1 Definitions (continued)

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 cycle REPORT (COLR) 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 thyroid dose conversion factors from:

1) Table IlIl of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or

2) Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or

3) International Commission on Radiological Protection (ICRP) Publication 30, "Limits for Intakes of Radionuclides by Workers," Supplement to Part 1, pages 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," 1979, or

4) Table 2.1 of EPA Federal Guidance Report No. 11, EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factois for Inhalation, Submersion, and Ingestion,"

1988.

(continued)

CALLAWAY PLANT 1.1-2 Amendment No.

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE-1 33 DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (microcuries per gram) that alone would produce the same acute dose to the wMole body as the combined activities of noble gas nuclides Kr-85m, Kr-87, Kr-88, Xe-1 33m, 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-1 33 shall be performed using the effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "Extemal Exposure to Radionuclides in Air, Water, and Soil", 1993.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its ESF actuation TIME 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 leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from 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);

fi-nnfiniumd)

CALLAWAY PLANT 1.1-3 Amendment No.

Definitions 1.1 1.1 Definitions LEAKAGE (continued) b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water 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 any one 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, component, or device to perform its specified 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 of the FSAR;

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

c. Otherwise approved by the Nuclear Regulatory Commission.

(continued)

CALLAWAY PLANT 1.1-4 Amendment No.

Definitions 1.1 1.1 Definitions (continued)

PRESSURE AND The PTLR is the unit specific document that provides the TEMPERATURE LIMITS reactor vessel pressure and temperature limits, including REPORT (PTLR) heatup and REPORT (PTLR) cooldown rates, the power operated relief valve (PORV) lift settings, and the Cold Overpressure Mitigation System (COMS) arming temperature, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.6. Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature (PrT) Limits," and LCO 3.4.12, "Cold Overpressure Mitigation System (COMS)."

QUADRANT POWER QPTR shall be the ratio of the maximum upper excore TILT RATIO (QPTR) 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 rate to the (RTP) reactor coolant of 3565 MWt.

REACTOR TRIP SYSTEM The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME 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:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of 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 hot zero power temperatures.

(continued)

CALLAWAY PLANT 1.1-5 Amendment No.

T Definitions 1.1 1.1 Definitions (continued)

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.

TRIPACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of all devices in the channel (TADOT) required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the iequired setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

CALLAWAY PLANT 1.1-6 Amendment No.

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 1-1 31 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

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

ACTIONS CONDITION REQUIRED ACTION TIME A. DOSE EQUIVALENT - NOTE 1-131 not within limit. LCO 3.0.4.c is applicable.

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

< 60 pCi/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 ------ NOTE-----------------

XE-133 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.

(continued)

CALLAWAY PLANT 3.4-40 Amendment No.

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION 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. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR DOSE EQUIVALENT 1-131

> 60 pCi/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 -- NOTE-----

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-1 33 7 days specific activity

225,uCi/gm.

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

1.0 pCi/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 CALLAWAY PLANT 3.4-41 Amendment No.

ATTACHMENT 4 PROPOSED TECHNICAL SPECIFICATION BASES CHANGES (for information only)

a, RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND Te wmaximum doao to th1PAohec! body and the thyroid that an individual at th sIe Ondar' can recoive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during er egcident ir spc-ified in 10 CFP 100 (Rf. 1). Thl.. itspoeifiaeivity scnourc that the doCGE arc he!d to a small fraotn of the 10 CFR 100 limits during annaly--ad tran;&crF and eeeidents.-=Vr.fR-r7 A The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the -ffoite-rdieectivity dose consequences in the event of a sk^

rrei,,a4CCfI4)or steam generator tube rupture (SGTR) accident.

The LCO *ntains specific activity limis for both DOSE EQUIVALENT 1-131 and rotc cpoofic activity. Th oellonoblelovols ar iLnded to li;.,;t

-tho hour dosc athe citc beundary to as mall froetion sfhei, 10 Crn 100 dose guidelino limitc. The limits in th LECO aore tandordized, based on

+arametrio coauaticna of-offsitc rodicoativity dozo conouonces fof-

-4,pra citc locotions.

Thc pa, latione ihowod thepotontia? offsitc dosc Icvel for a SGT.R accident Uweoro fractin of the 10 CFR 100 nn !ppropritlsall doec guideline limits. Each eovaluation eseurnea a broad range of 3itc-applicable atmospheric dispesion faet in o prammetric evaluation.

APPLICABLE The LCO Him its on the specific activity of the reactor coolant ensure that SAFETY the resulting hour doaca atth citc bedy wi not -xceed a sall ANALYSES -42atiom of tho 10 6r 100 dosc guideline limit3;fllevuing a 3ST-R-

,&ccident. The SGTCR-cofety snolyss (R1f. 2) assumes t# cpcificr 44iniil activity of thc rfactor ecoeant isgrfatr than the LCOlimit and assumec I an exictin%roactor coolant oteam pMenrtor (stC) tube leealge rate of 1 gpm. The safetyinlj I u s the initial specific efivityofthe condary coolant iSgroater than the limit of 0.1 jjgDO 9 I

-Tho enalyci for-tho SCTR accident-ostaUlihes the aOOetfl limits for CS s~pefietivit Rfno c to this analysJ-usuod to -----

ohanges3 to toe unit hat could affe- RCS cpeeifis activity, as they relate ietohle aoccptnc !Q.imits.

Thlve nA2YS F tW9 easor of rea49or cooiant cpecr.!R Usprformed eet ivty. - se I amcu eu et oneuifecnt argoe ioembspiKe I (continued)

CALLAWAY PLANT B 3.4.16-1 Revision 1

INSERT A The maximum dose to the whole body and the thyroid 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 100.11 (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.

INSERT B 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).

INSERT C offsite and control room doses meet the appropriate SRP acceptance criteria following an SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the initial iodine specific activity of the reactor coolant is greater than the LCO limit (see the discussion of Case 1 below), and a pre-accident reactor coolant steam generator (SG) tube leakage rate of I gpm exists. The safety analyses assume the initial iodine specific activity of the secondary coolant is 10% of the Case 1 reactor coolant iodine specific activity, greater than the limit of 0.1 pCi/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 plant that could affect RCS specific activity, as they relate to the acceptance limits.

The safety analyses consider two cases of reactor coolant iodine specific activity. In Case 1, the initial reactor coolant iodine specific activity corresponds to an isotope mixture that bounds the SR 3.4.16.2 limit for both tight and open fuel defects. The isotopic mix is based on the initial RCS concentrations from FSAR Table 15A-5. This table provides conservative values for the iodine isotopic spectrum that bound the RCS concentrations which could be expected with either tight or open fuel defects. Since the assumed iodine spectrum represents bounding values for different types of fuel defects, the initial radioiodine inventory exceeds the SR 3.4.16.2 limit of 1.0 pCi/gm.

Case 1 also assumes an accident-initiated iodine spike that increases the rate of iodine release from the fuel rods containing cladding defects to the primary coolant immediately after an SLB or SGTR. The iodine spiking factor is assumed to be 500 for the Case I radiological consequence evaluations for SLB and 335 for the Case I radiological consequence evaluation for both SGTR radiological consequence analyses.

Case 2 radiological consequence evaluations for SLB and SGTR assume the initial reactor coolant iodine specific activity is a factor of 60 higher than Case I due to a pre-accident iodine spike caused by a transient prior to the accident.

In both Case 1 and Case 2 radiological consequence evaluations, the noble gas specific activity in the reactor coolant is assumed to be greater than the 225 pCigm DOSE EQUIVALENT XE-1 33 limit in SR 3.4.16.2. The dose analysis assumptions are discussed further in Tables 15.1-3 and 15.6-4 of Reference 4.

ij -Q RCS Specific Activity B 3.4.16 BASES APPLICABLE --Hiatinese tl ,atecow~iodi iei lease- nto the ieacter coolar a-SAFETY -facter ef'aou eimmadiato!y'eftzrtJc

~ 2-2-t.

ANALYSES (continued) 3scumc the initial rcctar coolant iodine activity is a factor Of C6higher than CaEO 1 duo to a prc cscidentidi~nc pilec eaused by anRCS trancicnt. In-both cMes, the nelo acati-vity in the reaetor coolant o"sumc 1%faled fel iL-to I

R _aNFp- esfiansap di3Zu d fuli Itr ;n Tlbe 15.-4 s l 'kercicc 2. -

Thevanalysis also assumes a loss of offsite power at the same time as the reactor trip i T prn c . The SGTR causes a reduction in reactor coolant inventory. The reduction initi tes a reactor trip from a low pressurizer pressure sign~ ArA/ g-r D 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 atmospheric steam dump valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends .,o( VA khlR .ryr4-e, Xrf a4cS ;n -

-The safety analysis chow.c the radiological con.^encca of ZGTR

-2^cident 2re within -the SRPP 16.6.3 froetionc of thc Re-fcrcn1 JU;vd-uideline Uimit^. Oporatiore with iodcic pecifie eetiyity levels geeatcr than thm LCO limit jJ prrias, if ths ectivity levels -d not xceed the limit ehown in rFigufe 3.1.1 1.i,itho 2pplicable specification, for more than-18 hourc. The ^afety nlyoio has co ntand pit lcide opikfing -,-casc.- I o ~rde thc eboev eirit permisebl" iodinc lICYlC shown in f-gur-o3.4.16 1 arc eooptable becw'> nf the IEr prob bility of a SGTR eccidont eeetifring durirq the established 43 hour4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months time limit. Thoe

-oeurfenee of an GCTR aoeident at the e permissible leveIS coutd

-isrrasethe slto bourndary dose lcve6,but-lbo within 10 CFR100 ffYRr The limits on RCS specific activity are also used for establishing standardization in radiation shielding and planqradiation protection practices. /e,-rcnn e/

RCS specific activity satisfies Criterion 2 of IOCFR50.36(c)(2)(ii).

LCO Jhe specific-odine ctivity is limited to 1.0 ci;l ECC EQUIVALEIT 1--131, end thc gross opAcific a n thc rooztor coolant ip t (continued)

CALLAWAY PLANT B 3.4.16-2 Revision 1

INSERT D in the analysis of an SGTR with a failed ASD on the faulted steam generator. In the analysis of an SGTR with a failed AFW flow control valve on the faulted steam generator, reactor trip and safety injection are assumed to occur at the time of the tube rupture to maximize the potential for overfilling the ruptured steam generator.

INSERT E 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 steamline pressure. The affected SG blows down 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 pCi/gm for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

INSERT F The iodine specific activity in the reactor coolant is limited to 1.0 pCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 225 pCi/gm DOSE EQUIVALENT XE-1 33. 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 an SLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2).

RCS Specific Activity B 3.4.16 BASES LCO number of pCJi equal u1 ;Jid b, vcrago dicintogtion (continued) -energy of the sum if th-leberage beta and gnamm en of the CAgioc

-cool2nt nuclide ). The limit en DOSE EQUIVALENfT 1-131 ensttresthc

-2 hour thyroid doco to an individual at th cito boundary during the Dein Basiq Accid (DBA) will bo a mall fraction-of tho al:o'..d thyroid dosc.

-The limit en gress specific activity ensurae the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months whole ltbs dedto an individual at the cite boundary during the CBA will be a fraction amwll ef thc allewed-wholc body desc.

Tho GCTR accidgnt mm lycis (Ref, 2) shows that thc 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Gite boundary' cda Inle'.e aro withccptablc imits. Violation of the LGO may result -

in ractor coolant radioactivity leve t could, in the event ef an CGCTR,

'-ladto siteoe idaun;derlses that-emeeed the 10ECeFR 10~ee-~

APPLICABILITY In4MODEG end 1 2, andmin MODE e with RGCS acrage temwpraturo-

-Ž509F, operation within the LCOErnitfor DOSE EQUIVALEI4T -131 and groess specifie ctivity are necsar,' tc eentainthe ,potential

-concgucncc of-an GCTR4to within the accptable site boundary doo For operation in MODE 3Keith RC vcrago temperaturc - 6000F, and int MODE-S 4 and 6, the eftite ca 4f radioatty in the event of an CGT-R i_ unlikcly sine the caturation prcsurc of the reactor coolant is below the lift pressure setings f tithe nwin steamfi safety and ati 1.suPl MI it--

eteam dump valves.

ACTIONS A.1AandA.2 / A2 rle4~-k bDotc A Note permits the use of the provisions o LCO 3..4.c. Thie permits entry into the applicable MODE ile felyin

/o"ve This allowance is acceptable due to the einificant conservatism 0e. Is 4. '14 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 transiertecific activity excursions while the plant remains at, or proceeds-to, power operation.

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 Jome-ef-urc 3.4.1 C-i are nt ,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 4mple. Sampling is done to continue to provide a trend. Lr7eci'./rc acTv~7 St c G0SuC;/,r (continued)

CALLAWAY PLANT B 3.4.16-3 Revision 5a

INSERT G 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 an 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.

RCS Specific Activity B 3.4.16 BASES ACTIONS A.1 and A.2 (continued) A'JrA'7 /-/

The DOSE EQUIVALENT 1-131 must be restor d to within limitwithin 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion lime of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is quir- thlmi viuolatinn rPsi llteP from normal iodine spileing and is aocpta bncaue of-the Inw prnhability of n TR c c y Jum ;n thi egperiod.

B. Avow ve

-e 4, &

-n1 ewt ?/

-Alh the gross specific activity in exeessfZ the allowed limit, the unit most be placed in MO19BE end Tavg -must be reduced to -- 568^F within

-Ch nCot safety saturction pr and e thlercaotor s otoet duwmpxa vos and prevents venting the SC to the environment in n 3GTR event. Tho allowed Completion lmo cf Chours is reasonable, based en opoeting oxporionce, to rcach MODE 3 beow SOOTF from full powcr

-conditicns in an orderl a.Kde an vfithout challenging plant systewms CA1 *net C.2 r > 6- u^/)

4Me If eRequired Action and 4he associated Coinetion Time of Condition not meteor if the DOSE EQUIVALENT 1-131 is nthe unaceeptablfregio.

of Figu.re 3.1.16 1, the reactor must be brought to MODE 3eR4dRAe a'.'orngo temPQPatuPr muct bo to -Soo

-ducodwithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months The Completion Time6 bewR easonable, based on operating experience, to reach QG E 3below SGoar from full power con tions in an orderly manner a dwithout challenging plant systems. I

-tMz iIsu.rc-afji . /

SURVEILLANCE SR 3.4.16.1 ,4 n 3S AoDrt)

A) MAai7,5 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the ee cific activity of the reactor coolant at least once every 7 days. While basically;* aquontiw furc of radienuclides wvith half livos longer than 15 minutce,9oxcluding iodinec,1Fiis measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in feseespecific activity.

Mfe ,vS/e ad Trending the results o' this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. Tho Surveillance is applieeble in MODES I and 2, and in (continued)

CALLAWAY PLANT B 3.4.16-4 Revision 0

INSERT H 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 an SLB or SGTR occurring during this time period.

INSERT I With the DOSE EQUIVALENT XE-1 33 greater than the LCO limit, DOSE EQUIVALENT XE-1 33 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 an 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 Action B.1 while the DOSE EQUIVALENT XE-1 33 LCO limit is not met. This allowance is acceptable due to the 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.

RCS Specific Activity B 3.4.16 BASES SURVEILLANCE SR 3.4.16.1 (continued)

REQUIREMENTS MODE 3 A w'ith Tpvg a ctk 500-f. The 7 day Frequency considers the unlikelihood of a gross fuel failure during Utheime.

CT SR 3.4.16.2 wpe^-wa y{- -Me L-Co This Surveillance is performed to ensure iodineMremains withifimit during normal operation and following fast power changes when fel-failure is-r/;1 ¶e r f kPna, 4 more apt to occur. The 14 day Frequ yeyjsjdequate to trend changes in the iodine activity level, considerir 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 2 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 iodine levels peak during istime s followin H rsamples at other times would provide inaccurate results. Nn-lteo mod-ifies this SR to altGw-"atryr

.aton~ pprafinn in MODE3r an inM E3with S T-,,g i- 6900F prior to rf- 1hR. This allows the s ervillanee to be perffemed in these-

-M9BES, Adi tv uof Nentein M85iE- 1. -tad, ne rpf eke rn#- f4-k0nj

_ A } a__

=/V~j -4eA-T XR341.

Aadiochemical analysis for E determination is required every 184 days (6 m hs) with the plant operating in MODE 1 equilibrium condi The C de ination directly relates to the LCO and is requicto verify plant operation hin the specified gross activity LC it. The analysis for Eis a measurem f the average energi disintegration for isotopes with half lives on han 15 mil s, excluding iodines. The Frequency of 184 days recogni es not change rapidly.

This SR has been modif y Note that es sampling is required to be performed wIt 1 days after a minimum o fective full power days and 20 d of MODE I operation have elapsed si the reactor was last critical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures that tadioactive matehs are at equilibrium so the analysis for P is representative d not flowed by a crud burst or other similar abnormal event.

REFERENCES 1. 10 CFR 100.1 1, 1973.

FSAR, Section 15.6.3.

I (2, JX4*.dA-a4 ket frilea 10/4,n (Jw), /5; Al'0-

'A (44CA) Ah/ Jechk, I5ja.3 (Xd7-,x).

rhe) r 4t -1 I-r CALLAWAY PLANT B 3.4.16-5 Revision 0

I INSERT J If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 in Specification 1.1, "Definitions," is not detected, it should be assumed to be present at the minimum detectable activity.

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.

INSERT K 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.

i T -t ATTACHMENT 6

SUMMARY

OF REGULATORY COMMITMENTS

iL. -, *'j,- -

SUMMARY

OF REGULATORY COMMITMENTS The following table identifies those actions committed to by AmerenUE in this document. Any other statements in this submittal are provided for information purposes and are not considered to be commitments. Please direct questions regarding these commitments to Mr. Dave E. Shafer, Superintendent Licensing, (314) 554-3104.

COMMITMENT Due Date/Event Changes to the Fuel Clad Degradation Emergency Action Within 90 days of Levels will be implemented within 90 days after NRC amendment approval.

approval of the amendment application.

v t e Letter Sequence Request
TAC:MD1814, (Approved, Closed)
Initiation Request Acceptance... Administration Questions Answers Results Approval Other: ML063490274
MONTHYEAR2006-12-18 [Table View]

ULNRC-05281, Application for Amendment to Facility Operating License NPF-30, Adoption of Industry Traveler TSTF-490

Text

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

Conclusion:

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

SUMMARY

SUMMARY

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ULNRC-05281, Application for Amendment to Facility Operating License NPF-30, Adoption of Industry Traveler TSTF-490

Text

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

Conclusion:

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

SUMMARY

SUMMARY

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