Text

Revision to Technical Specification (TS) 1.1, Definitions, and TS 3.4.16, RCS Specific Activity
	ML053070557
Person / Time
Site:	Wolf Creek
Issue date:	10/27/2005
From:	Hedges S; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	WO 05-0025
	Download: ML053070557 (54)
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7

-W6LF--CREEK

'NUCLEAR OPERATING CORPORATION Stephen E. Hedges Vice President Operations and Plant Manager October 27, 2005 WO 05-0025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Revision to Technical Specification (TS) 1.1, "Definitions," and TS 3.4.16, ARCS Specific Activity".

Gentlemen:

Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Facility Operating License Number NPF-42 for the Wolf Creek Generating Station (WCGS).

The proposed license amendment request (LAR) proposes to 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 limits 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 1-131 definition (corresponding to the lodine-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 captured in the current E definition. -The current 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

..pDC P.O. Box 411 Burlington. KS 66839 I Phone: (620) 364-8831 An Equal Opportunity Employer MIF7HCN-ET

WO 05-00025 Page 2 that value is very high (the denominator is very low) if beta emitters such as H-3 (tritium) are included in that value, as required by the.E 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.]

Wolf Creek Nuclear Operating Corporation (WCNOC) 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 TXU Power, AmerenUE, WCNOC, Pacific Gas & Electric Company, 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.

Other members of the STARS group can also be expected to submit a license amendment request similar to this one. The other license amendment requests will be submitted on a parallel basis within a short period of time of each other, with plant-specific information presented within brackets (i.e., within [ ]) in Attachment I (other than TS LCO numbers which vary between the 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 utilize 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 AP1000 in the NRC letter to Westinghouse Electric Company dated September 13, 2004.

Attachments I through VI provide the Evaluation, Markup of Technical Specification Pages, Retyped Technical Specification Pages, Proposed TS Bases Changes (for information only),

Proposed Updated Safety Analysis Report pages (for information only) and Summary of Regulatory Commitments, respectively, in support of this amendment request. Final TS Bases changes will be implemented pursuant to TS 5.5.14, 'Technical Specification Bases Control Program," at the time the amendment is implemented. A revision to the fuel element failure Emergency Action Level that reflects the approved TS 3.4.16 limits will be implemented at the time the amendment is implemented.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. The amendment application was reviewed by the WCNOC Plant Safety Review Committee.

WO 05-0025 Page 3 In accordance with 10 CFR 50.91, a copy of this amendment application is being provided to the designated Kansas State official.

WCNOC requests approval of this proposed License Amendment by June 1, 2006.

The changes proposed are not required to address an immediate safety concern. It is anticipated that the license amendment, as approved, will be effective upon issuance, to be implemented within 90 days from the date of issuance. If you have any questions concerning this matter, please contact me at (620) 364-4190, or Mr. Kevin Moles at (620) 364-4126.

Very truly yours, E. Hedges SEH/rIg Attachments:'

I-Evaluation 11 -

Markup of Technical Specification pages IlIl -

Retyped Technical Specification pages IV -

Proposed TS Bases Changes (for information only)

V -

-Proposed USAR Changes (for information only)

VI -

Summary of Regulatory Commitments cc:

V. L. Cooper (KDHE), w/a J. N. Donohew (NRC), w/a

.W. B. Jones (NRC), w/a B. S. Mallett (NRC), w/a Senior Resident Inspector (NRC), w/a

STATE OF KANSAS COUNTY OF COFFEY

) Ss Stephen E. Hedges, of lawful age, being first duly sworn upon oath says that he is Vice President Operations and Plant Manager of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

Ste n

. Hedges Vice Pr ident Operations and Plant Manager SUBSCRIBED and sworn to before me this X9 day of 004-, 2005.

I ZZ4 Notary Public

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Attachment I to WO 05-0025 Page 1 of 16 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 1-131 (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 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132,1-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in [Table Ill of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites.']"

to "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 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)

ICRP 30,1979, page 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or

4)

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

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

Attachment I to WO 05-0025 Page 2 of 16 conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil," or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.]"

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 1-131 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 (Tavg) 2 500'F."

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

TS 3.4.16 Condition A would be revised from:

DOSE EQUIVALENT 1-131 > 1.0 pCi/gm.

to DOSE EQUIVALENT 1-131 not within limit.

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

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

to "Verify DOSE EQUIVALENT 1-131 5 60 pCi/gm."

1 ]

TS 3.4.16 Condition B would be revised from:

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

to "DOSE EQUIVALENT XE-1 33 not within limit."

Attachment I to WO 05-0025 Page 3 of 16 TS 3.4.16 Required Action B.1 would be revised from:

"Be in MODE 3 with Tavg < 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 1-131 > 60 pCi/gm."

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

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

to "C.1 Be in MODE 3.

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 ." Note: 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 .

Attachment I to WO 05-0025 Page 4 of 16 Surveillance Requirement (SR) 3.4.16.1 would be revised from:

"Verify reactor coolant gross specific activity S100/ E pCi/gm."

to

'--------------------------------------NOTE---------------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-1 33 specific activity s [500] 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 P -

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 II and IlIl provide the TS markups reflecting the above changes and the retyped TS. Attachment IV provides an information-only copy of the associated TS Bases changes.

Attachment V provides an information-only copy of related USAR changes. Attachment VI 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 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.

Attachment I to WO 05-0025 Page 5 of 16 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, 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 assume both an accident-initiated iodine spike release to the primary coolant occurring immediately after the accident (Case 1) and a scenario whereby a reactor transient occurs prior to the postulated accident raising the primary coolant iodine value to the maximum permitted by technical specifications at full power operations (Case 2). The results of the SGTR radiological consequence analyses are described in [USAR Section 15.6.3]. The results of the MSLB radiological consequence analyses are described in

[USAR Section 15.1.5].

[

]

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 Regulatory Guide 1.109, Revision 1, 1977, have been used in radiological consequence analyses performed to date.] Any of the NRC-approved thyroid dose conversion factor references cited in the revised definition of DOSE EQUIVALENT 1-131 can 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 for the Case 1 radiological consequence evaluation for the SGTR analysis.

[The Case 2 radiological consequence analyses for SGTR and MSLB assume a reactor transient has occurred prior to the postulated accident and has raised the primary coolant iodine concentration to the maximum value permitted by the technical specifications at full power operations.]

This [corresponds to] 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%. TS Figure 3.4.16-1 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 [greater than 500] pCi/gm DEX. These assumptions are discussed further in [USAR Tables 15.1-3 and 15.6-4]. The initial DEX concentrations were calculated assuming [1% failed fuel] and using (whole body 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, or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, 19771.

Attachment I to WO 05-0025 Page 6 of 16 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.

[Three] 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 1OO/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 noniodine activity in the coolant."

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% failed fuel],

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 DEl to as high as (300] pCi/gm DEI at [20]% RTP.

Below [20]% 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) are included in that value, as required by the E 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

Attachment I to WO 05-0025 Page 7 of 16 compared to 100/I 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 Chanqes 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, [Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, and 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].

EPA Federal Guidance Report No. 11 [and ICRP Publication 30] 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 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 are 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 essentially meaningless. The use of E also results in a TS limit that can vary during operation as different values for F are determined, resulting in different values for the gross specific activity limit (1 00/P /uCi/gm).

Attachment I to WO 05-0025 Page 8 of 16 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 concentration, short half life, or small dose conversion factors. The excluded isotope [Kr-83m, Kr-85, Kr-89, Xe-131m, and Xe-137] contributes less than 2% of the whole body dose contributions from noble gases in the accident analysis.] 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 111.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081, "External Exposure to Radionuclides in Air, Water, and Soil," 1993, [or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977].

These dose conversion factors are applicable for determination of DEX. The use of the dose conversion factors for air submersion listed in Table 111.1 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 the iodine specific activity in terms of DEI and noble gas specific 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 Condition A and SR 3.4.16.2 for iodine specific activity.

The proposed change states "RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits. The DEI limit of <1.0,pCi/gm is contained in current Condition A and SR 3.4.16.2. In addition, the limit of 1.0 pCi/gm is consistent with the current SGTR and MSLB radiological consequence analyses discussed in Section 3.1 above.

Attachment I to WO 05-0025 Page 9 of 16 The DEX limit of [500] pCi/gm contained in revised SR 3.4.16.2 is [more limiting than the current SGTR and MSLB radiological consequences discussed in Section 3.1 above. The noble gas specific activity limit is established based on the maximum accident analysis activity corresponding to 1% fuel defects with sufficient margin to accommodate the exclusion of those isotopes based on low concentration, short half life, 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 concentration, 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 design basis accident radiological consequences analysis and would not fluctuate with variations in the calculated value of E during normal operation as is currently the case.

TS 3.4.16 Applicability 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 '5 1.0 [tCi/gm" with the words "not within limit" to be consistent with the revised TS 3.4.16 LCO format. The DEl limit of < 1.0 jCi/gm is contained in SR 3.4.16.2.

TS 3.4.16 Required Actionr 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 piCVgm for DEL.

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 [300] pCi/gm at [20]% 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.

Attachment I to WO 05-0025 Page 10 of 16

[1 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 limits. 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 limits as discussed above. The DEX limit of [500] jaCi/gm is contained in revised SR 3.4.16.1.

The limit of [500] pCi/gm is [established based on the maximum accident analysis activity corresponding to 1% fuel defects with sufficient margin to accommodate the exclusion of those isotopes based on low concentration, short half life, or small dose conversion factors.] 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 USAR Tables 15.6-5, 15.6-5A, and 15.1-4 at WCGS 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.

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 /JCi/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.

Attachment I to WO 05-0025 Page 11 of 16 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 value of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is consistent with other TS which have 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 S [500] PiCi/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.

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 [300] pCi/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.

Attachment I to WO 05-0025 Page 12 of 16 4.3 Summary In summary, the proposed changes will revise the definition of DOSE EQUIVALENT 1-131, 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 1-131, add a new LCO limit to TS 3.4.16 for DOSE EQUIVALENT XE-133, increase the Completion Time of Required Action B.1, delete TS Figure 3.4.16-1, and revise the TS 3.4.16 Conditions and Required Actions 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 1-131 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, and 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, in which 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 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 1-131, delete the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition for DOSE EQUIVALENT XE-1 33, revise TS 3.4.16 to specify an LCO limit on DOSE EQUIVALENT 1-131, add a new LCO limit to TS 3.4.16 for DOSE EQUIVALENT XE-133, increase the Completion Time of 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

[Wolf Creek Nuclear Operating Corporation (WCNOC)] 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:

Attachment I to WO 05-0025 Page 13 of 16 (1)

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Response

No The proposed changes would add new thyroid dose conversion factor references to the definition of DOSE EQUIVALENT 1-131, eliminate the definition *of C 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, replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT 1-131, 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)

The proposed change does not 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)

The proposed change does not 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 full power operation, which is

Attachment I to WO 05-0025 Page 14 of 16 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, [WCNOC] 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, that 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, that 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, that 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," that 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 [WCGS] 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 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.

Attachment I to WO 05-0025 Page 15 of 16

6.0 ENVIRONMENTAL CONSIDERATION

[WCNOC] 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 Evaluation Compliance with 10 CFR Part 50, Appendix I," October 1977.

7.

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

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.

Attachment I to WO 05-0025 Page 16 of 16

11.

NUREG-1512, '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," 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-1512, "Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design, Docket No.52-003," dated August 1998 and for the AP1 000 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 APIOQO advanced reactor designs.

1 to WO 05-0025 Page 1 of 9 ATTACHMENT 11 MARKUP OF TECHNICAL SPECIFICATION PAGES

Attachment II to WO 05-0025 Page 2 of 9 Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

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

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

CORE ALTERATION 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.

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

cim DOSE EQUIVALENT 1-131 (continued)

Wolf Creek - Unit I 1.1-2 Amendment No. 123

Attachment II to WO 05-0025 Attachment II to6W0 05-0025 Page 3 of 9 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)

ICRP 30,1979, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or

4)

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

INSERT 1.1-2B DOSE EQUIVALENT XE-1 33 DOSE EQUIVALENT XE-133 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-1 35m, Xe-1 35, and Xe-1 38 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, 1993, 'External Exposure to Radionuclides in Air, Water, and Soil," or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.

1 to WO 05-0025 Definitions Page4of9 1.1 1.1 Definitions (continued)

J(PISIN~iGRATI IEE Y

othertno a

sytt ha f ives 15 mites, fakin~up' (

a e

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 the methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

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

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 (SG) to the Secondary System;

b.

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

Wolf Creek - Unit 1 1.1 -3 Amendment No. 423, 131

Attachment II to WO 05-0025 RCS Specific Activi Page5of9

3.

6 REACTOR COOLANT SYSTEM (RCS) 3.4.1 XRCS Specific Activity LCO 3.4.1 The specific activity of the reactor coolant shall be within ts.

APPLICABILITY:

ODES 1 and 2, DE 3 with RCS average temperature (Ta.

500OF.

ACTIONS\\/

CONDITION REQUIRET9ACTION COMPLETION TIME A.

DOSE EQUIVALENT I-131---------

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

> 1.0 PCi/gm.

LCO 0.4

. is applicable.

6-A.1 Vify DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months EQ VALENT I-131 within the ac eptable region of Figure

..16-1.

/

AND\\

A.2 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

/

~EQUIVALENT 1-131\\x

/

~within limit.\\

B.

Gross s ecific activity of B.1 Be in MODE 3 with hours the r ctor coolant Tavg < 5000F.

>X/ E iCi/gm....

\\

/

disontiued)

I Wolf Creek - Unit I 3.4-4 1 Amendment No. 423,155

Attachment II to WO 05-0025 Page 6 of 9 RCS Specific Activity 3.4.16 INSERT 3.4-41 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:

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

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT 1-131

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

not within limit.

LCO 3.0.4c. 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 gCi/gm.

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

B.

DOSE EQUIVALENT

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

XE-133 not within limit.

LCO 3.0.4c. is applicable.

B.1 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT XE-1 33 to within limit.

(continued)

Wolf Creek - Unit 1 3.4-41 Amendment No. 12-3,1466,

Attachment II to WO 05-0025 Page 7 of 9 RCS Specific Activity 3.4.16 C. Required Action and associated Completion Time of Condition A not met.

OR 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months DOSE

.--r.>-FIo C-1di nequ~rec( 4ro br. -erbr Arv SURVEILLANCE SR 3.4.16.2 FIArn INV I C Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 1.0,uCi/gm.

14 days 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 Wolf Creek - Unit I 3.4-42 Amendment No. 123 1 to WO 05-0025 Page 8 of 9

Attachment II to WO 05-0025 P e9f9 RCS Specific Activi 3.4 300 E

U t 250 a

t o 200 IA.

ml 5a 150 a0 t

q-100 Y.

.1.

"I 50 a

a

+

\\

UNACC PTABLE

__OP ATION I

I__

ACCEPABLE

____OPERATION Y

= lDN/

lI n

20 30 40 SO 60 70 80 90 100 PERCENT OF RATED THER L POWER FIGURE 3.4.16-1 (page 1 of 1) eactorCoolant DOSE EQUIVALENT 1-131 Specific Activi Limit Versus Percent of RATED THERMAL POWER Wolf eek-Unit 1 3.4-44 Amendment

o. 123

Attachment III to WO 05-0025 Page 1 of 4 ATTACHMENT 11 RETYPED TECHNICAL SPECIFICATION PAGES

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions NOTE---------------------------------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)

ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AFD shall be the difference in normalized flux signals between the top and bottom halves of an excore neutron detector.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

(continued)

Wolf Creek - Unit 1 1.1 -1 Amendment No. 123,

fills Definitions 1.1 1.1 Definitions (continued)

CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

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

CORE ALTERATION 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.

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

DOSE EQUIVALENT 1-131 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 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 (continued)

Wolf Creek - Unit I 1.1 -2 Amendment No. 123,

Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT 1-131 (continued)

3)

ICRP 30, 1979. page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity,' or

4)

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

DOSE EQUIVALENT XE-133 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME LEAKAGE 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, 'Extemal Exposure to Radionuclides in Air, Water, and Soil," 1993, or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.

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

LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve (continued)

Wolf Creek - Unit I 1.1-3 Amendment No. 423,

I Hill Definitions 1.1 1.1 Definitions (continued)

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

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 (SG) to the Secondary System;

b.

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

c.

Pressure Boundary LEAKAGE LEAKAGE (except SG 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).

(continued)

Wolf Creek - Unit 1 1.1-4 Amendment No. 4S3, l

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity, LCO 3.4.16 APPLICABILITY:

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

MODES 1, 2,3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

DOSE EQUIVALENT 1-131

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

not within limit:

LCO 3.0.4c. 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 jCI/gm.

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

B.

DOSE EQUIVALENT-NOTE-------------------

XE-1 33 not within limit.

LCO 3.0.4c. is applicable.

B.1 Restore DOSE 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months EQUIVALENT XE-1 33 to within limit.

I (continued)

Wolf Creek - Unit 1 3.4-41 Amendment No. -123, 155,

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 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 gCi/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1

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

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT X-133 7 days specific activity < 500 gCi/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 giCi/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 I

Wolf Creek - Unit 1 3.4-42 Amendment No. 123, -55,

Attachment IV to WO 05-0025 Page 1 of 12 ATTACHMENT IV PROPOSED TS BASES CHANGES (for information only)

Attachment IV to WO 05-0025 RCS Specific Activity Page2of12 B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND (8

ximnu se to~e yvho bd/nd tled thyroid tha~a idvidul at

[tsite bo kary ca iecIV ~ r 2 surs '(ring ariiaccip nt ispeci dJ

,pn10 CFF 100 (R~f)

Tpelimit 'on s

~cific icv ty nsur that t~e ose re heldt I

sm,'fat ofth10 CF0R101 mitsuring/,

ayzdtriet ap s

ci.ns./

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the o

y dose consequences in the event of a (seoam lt'rsbrettk(st)orA steam generator tube rupture SGTR) accident.

The LCO contains specific activit limits for both DOSE EQUIVALENT 1-131_gnj-r-sp~cci~f.MI~~e~loa ee n

to I rheFFour Sse ajphe site i ounda7/(o a sm,# fractico tehe 1CFR 00 d0s uleline "wits.

t~lim it's ir he LC Yre sta r adze se on Xaati yuatino offsit fadioac iy dos rcoseqecso

/

y Si s ite h

at o

s/n Th~ramet ' eva lua shd lv Th rmr vlu s sh d the tetial site d9elv o

S R acc ent were n appr riately all frac on of ti 10 C 100 se gui eline lim'. Eac valuatic assum s a broSu rang of sit a lcl tohrcdprin cors i r ~rcev uatio APPLICABLE The LCO limits on the s ecific activitv of the reactor coolant ensure that SAFETY ANALYSES the esultin out bo dary will ot excpdd a small fracti oft 10 CF 100 d e guide* e limits Ilowin ha SGTR ac dent.

he SfR safet analysi Ref. 2) sumes e specif activi the r actor cplant at tpe LCO lit and a existin eactor 2olant stea gener or (SG) e leak e rate of gpm.

he safeyana a umes t specifi ctivity ohe seco dary co ant is a its limit o

,1 c/f OE QUVA NT 1-13,rmL£ 3.7.1 'econdpf Tea yssfor th S'R cdent esta shes the cpae limit fr RCSYpecific ac ity. Refer nce to thi nalysis used t assess ch ges to th unit that c Id affect S spec' activit, as they elateI o eaccepta e limis Wolf Creek - Unit 1 B 3.4.16-1 Revision 0

Attachment IV to WO 05-0025 Page 3 of 12 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 a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the'reactor coolant is at or more conservative than the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 pCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.18, "Secondary Specific Activity."

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

The analyses consider two cases of reactor coolant specific activity. One case assumes specific activity at 1.0 jlCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases, by a factor of 500, the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB or SGTR, respectively. The second case assumes the initial reactor coolant iodine activity at 60 ItCi/gm DOSE EQUIVALENT 1-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas specific activity is assumed to be equal to or greater than 500 ftCi/gm DOSE EQUIVALENT XE-133.

Attachment IV to WO 05-0025 RCS Specific Activity Page4of 12 B 3.4.16 BASES APPLICABLE e

alysis i erforme or two cas of reacto oolant spe fic SAFETY ANALYSES ac ity. 0 case ass es specif jactivity at.0 gCi/gm SE (continued)

UIVAJ NT 1-131 ith a conc rent large idine spike at increasp he ratof iodine r ease into e reactor c olant by a ctor of abo 500 im diately afte he accid t. The sec, nd case a umes the i i ial re ctor coola iodine ac ity at 60.0 i/gm DO EQUIVALENT 1-131 ue to a pr accident imne spike'used by RCS transjefit. In bot cases, t noble ga ctivity in t reactor c lant assun)6s 1 % fail fuel, w ich closely quals the CO limit of 00/

iCi/gfor gr lysis also assumes a loss of offsite power at the same time as the reactor trip The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signabl-93 The loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG Vne 9" 91discharges radioactively contaminated steam to the atmosphere through Ao~eeA u scrle the SG atmospheric relief valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown end; h

ty anal shows Ie radiologic conseque es of a GTR acci nt are thin a sm fraction of t Referenc 1 dose ideline Ii

s. Ope tion with ipmine specific ctivity leve greater an the 0

it is p missible, if e activity le els do not ceed the Ynits sho n in Figure

.4.16-1, i e applicabl specificatio for more an 48 urs.

The afety anal is has conc ent and pr accident i ine spik g level u

60.0 iC ;gm DOSE E IVALENT 131.

The rem nder of the ove limit per issible 10 ne levels own in Figur

.4.16-1 are cceptable b ause of t low proba ity of a TR acc ent occurri during the e ablished 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months time H it. The o urrence of n SGTR accient at thesIpermissible vels co

\\ncrease th ite boundary ose levels, ut still be wi in 10 CFR 100/

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

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

Wolf Creek - Unit 1 B 3.4.16-2 Revision 1

Attachment IV to WO 05-0025 Page 5 of 12 INSERT D in the analysis of an SGTR with a failed atmospheric relief valve 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 ptCi/gm for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Attachment IV to WO 05-0025 RCS Specific Activity Page 6 of 12 B 3.4.16 BASES LCO The ecifi odine a vity is lim'd to 1.0 p gm DOSE E6UJVALEN I-I 1, an he gros specific tivity in the actor coola pis limited t he mbe of tCi/g equal to 0 divided b itE(average isintegratio ener of the m of the erage beta nd gamma ergies of t co ant nuc es). The imit on DOS EQUIVAL T 1-131 en

-res the hour thyoid dose t an individu at the site undary durkg the Des n Basis ccident (D ) will be a all fractionf the allowe thyroid d e.

The mit on gro specific ac ity ensures e 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months w,(ole body se to a individual the site bo dary dunn e DBA wile a small action o eallowe hole body se.

The S R accident nalysis (Re. 2) shows t t the 2 hor site bo dary dos levels are wiin accepta e limits. Vi tion of.th LCO ma result eactor coola radioactivi levels that uld, in th event ofn SGT, ead to site b ndary dos that excee he 10 CF,100 do guid e

APPLICABILITY i

S nd 2, d in MO 3 with)CS avera e

e ur

Ž 0°F, eration ithin the CO limit for DOSE QUIVA NT 1-13 d gro specif activity re neces ry to conin the p ential cons uences f an SG R to with' the acce able sit oundary ose val s.

or ope tion in DE 3 w RCS av age te erature 000F, a d in MOD S 4 and

, the off& e releaseI f radioaivity in th event of SG;R is likely si e the sa ration pr sure of e reacto oolant i elo he Iipesrsettings the maita t and mospher eir ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit samrles 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 thed w) sprfic odhM-¶, s (F~ig e3.1 W

e p6t px(ce4ded) The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months isfiuireduif themit,

Wolf Creek - Unit I B 3.4.16-3 Revision 0

Attachment IV to WO 05-0025 Page 7 of 12 INSERT F The iodine specific activity in the reactor coolant is limited to 1.0 pCi/gm DOSE EQUIVALENT I-131, and the noble gas specific activity in the reactor coolant is limited to 500 pCi/gm DOSE EQUIVALENT XE-133. The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).

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

INSERT G In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 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.

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.

Attachment IV to WO 05-0025 RCS Specific Activity Page8of12 B 3.4.16 BASES ACTIONS A.1 and A.2 (continued)

A Note permits the use of the provisions of LCO 3.0.4c. This allowanc permits entry into the applicable MODE(s) We -

nr

.1 Cf 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 lant remains at, or proceeds to power operation.

,rely ng on Reolute Ae A.A oLA&.2 w'lAel

+Ar-,e Vc OE E4UIVA\\£4T l-l!A I al-oit is Fre met.

B.1 Wth tegro pecifi ctiv ty iecess oteaoedlimit, eunit ust b

aced a MODin whic the requi ment ds not a ly.

he c nge wit 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MOD and R S avera temper ture

<50 F lowerIthe sat ation pre ure of e reactor oolant low the se oints of e mainteam safa and a ospheric elief vales and events nting t SG to th environ ent in an GTR ev nt. Th tallowed omplefn Time of hours ireasona e, base on ope ting expernce, to ach MO9 3 beloy5000F fr full poer con tions anderly mnner and ithout ch i

ant syst C. 16a~,)A (7RRequired Action and e associated Comi etion Time of Condition A is (g) not met or if the DOSE EQUIVALENT 1-131 is h

the reactor must be brough ODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Tim rs'reasonable, based on operating ex erience, to reach from full power conditions in an orderly manner and without lo.< coX' tlosy challenging plant systems.

t

=

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure n

of the

's~ pecific activit of the reactor coolant at least once eve 7dafs.l ey sie

~lyo q n atis'e eaeras rs ihncid Ii~sn n~Spin scl~i in n,is measurement is the Wolf Creek - Unit 1 B 3.4.16-4 Revision 19

Attachment IV to WO 05-0025 Page 9 of 12 INSERT I With the DOSE EQUIVALENT XE-133 in excess of the allowed limit, DOSE EQUIVALENT XE-133 must be restored to within limits 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 3.0.4c. This allowance permits entry into the applicable MODE(s), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

Attachment IV to WO 05-0025 RCS Specific Activity Page 10 of 12 B 3.4.16 BASES SURVEILLANCE SR 3.4.16.1 (continued)

REQUIREMENTS sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any ipecific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

T tfvas0aThe7ay0Frequ-ency considers the unlikelihood of a gross fuel failure during Otime.

SR 3.4.16.2 eilce ii by Not. T No difi t* i

/

ur a eto 10 nt into nd era on i MOE E Ffi

-to rfor ingis s eil ince-equ-em,nt.

srvanc is perf nsur remains w iti du n

al operation and following fast power changes when f

more apt to occur. The 14 day Frequency is adequate to trend changes-in the iodine activity level, considering Qactivity is monitored every (I 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 this time following samples at other times would provide inaccurate results.

radioc mical a alysis fo deter ation is r 4red eve 184 days (6 moths) wit e plant peratingj MODE I quilibrium nditions.

Thy deternation d' ctly rela s to the L and is re uired to very ant oper ion withi the spe ied gross ivity LCO mit. The a lysis for E is measur ent of t average ergies pe disintegrati for isot es with h f lives lo er than 1 minutes, e cluding iodi es. The Fquency 84 days cognize does not ange rapiy.

This S as been odified by Note that icates s pling is r uired to be erformed ithin 31 d s after a mjimum of effective I power da and 20 ys of MO I operati have ela sed sinc e reactor fas last su critical for least 48 h rs. This sures t the radio tive materials &re at equili rium so the nalysis f f is re esentative nd not or othe rmal e t.

Wolf Creek - Unit 1 B 3;4.16-5 Revision 0

Attachment IV to WO 05-0025 Page 11 of 12 INSERTJ If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-1 33 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 operating 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.

Attachment IV to WO 05-0025 Page 12 of 12 RCS Specific Activity B 3.4.16 BASES REFERENCES

1.

10 CFR 100.1 1, 1973.

Q(

,@)

USAR, Section 15.6 7.

sR ALFtzAA5%.s., A Wolf Creek - Unit 1 B 3.4.16-6 Revision 0

Attachment V to WO 05-0025 Page 1 of 4 ATTACHMENT V PROPOSED USAR CHANGES (for information only)

Attachment V to WO 05-0025 WOLF CREEK Page 2 of 4 CHAPTER 11.0 RADIOACTIVE WASTE MANAGEMENT 11.1 SOURCE TERMS This section presents the design bases for determining the source terms for radioactive releases from the plant, for shielding within the plant, and for accident analysis performed in Chapter 15.0.

The source terms used for releases, shielding, and accident analyses are based on 0.12, 0.25, and 1.0 percent fuel defects, respectively.

Actual release data is contained in Annual Radioactive Effluent Release Reports filed with the NRC in accordance with Offsite Dose Calculation Manual (ODCM) requirements. Data supporting Chapter 15 accident analyses is not considered historical and is maintained current.

11.1.1 RADIOACTIVE CONCENTRATIONS AND RELEASES Reactor coolant and secondary coolant specific activities for an assumed 0.12-percent fuel defects and an assumed 100 pounds per day primary-to-secondary leakage are listed in Table 11.1-1.

The basis for calculating these sources is Regulatory Guide 1.112.

Compliance with Regulatory Guide 1.112 is discussed in Table 11.1-3.

Appendix ll.lA provides a description of the input used.

The decontamination factors applied are based on Regulatory Guide 1.112. A description of liquid leakage rates, process paths, and associated component activity levels is contained in Section 11.2 and Appendix l1.lA.

A description of gaseous leakage rates, process paths, and associated activity levels is contained in Appendix ll.lA and Sections 11.3 and 9.4. In-plant airborne activity concentrations and other data regarding the ventilation systems are provided in Sections 12.3 and 12.4.

11.1.2 SHIELDING Reactor coolant and secondary coolant source terms used for shielding are based on 0.25-percent fuel defects.

The source terms and the parameters used to calculate the source terms are given in Table 11.1-4 and Appendix ll.1A, respectively.

Table 11.1-6 provides the isotopic composition of the contained sources for radioactive waste management systems and for large, potentially radioactive outside storage tanks.

11.1.3 ACCIDENT ANALYSIS SOURCE TERMS Except for a LOCA and a fuel handling accident, the specific activity used for accident analysis releases is based on operating with 1-percent fuel defects.

Table 11.1-5 provides the isotopic (Wath rsur s In oL.

RC$ Olxwit 1Ad& god'e I 3.4.L(~,

t of s o

T0SE ELWaVALE'W XE-1335.

112.-

1 Rev. 14 l

Attachment V to WO 05-0025 WOLF CREEK Page 3 of 4 composition of the reactor coolant based on 1 percent fuel defects.

l 11.1-6 provides the inventory of the contained sources for radioactive waste management systems and for large, potentially radioactive outside storage tanks.

Sources for the LOCA are based on TID 14844.

Sources for the fuel handling accident are based on Regulatory Guide 1.25.

Chapter 15.0 provides a complete discussion and a listing of the source terms for each accident analyzed.

a`0

.IecA 1- ;g j tn dei n;l~oo O6 RuvLWrX-3 te oJCn^

11.1-2 Rev. 0

Attachment V to WO 05-0025 Page 4 of 4 WOLF CREEK TABLE 11.1-5 I

Reactor Coolant and Secondary Coolant Specific Activities -

1% Fuel Defects" Reactor Coolant Class 1 gaiLgnM Secondary Coolant

. 11cm Kr-83m*-

Kr-85m Kr-851K Kr-87 Kr-88 Kr-s 9*

Xe-131mf Xe-133 Xe-133m Xe-135m Xe-135 Xe-137-r Xe-138 5.54E-01 2.26E+00 9.41E+00 1.47E+00 4.26E+00 1.21E-01 3.41E+00 2.90E+02 5.37E+00 6.04E-01 9.82E+00 2.24E-01 8.15E-01 1.92E-05 7.1OE-05 2.96E-04 4.62E-05 1.34E-04 3.71E-06 1.07E-04 9.12E-03 1.73E-04

8. 62E-05 3.20E-04 6.90E-06 2.55E-05 Total noble gas Class 2 Br-83 Br-84 Br-85 I-130 I-131 I-132 I-133 1-134 I-135 Total halogens 3.29E+02 1.09E-01 5.82E-02 6.86E-03 3.57E-02 2.80E+00 3.14E+00 4.93E+00 7.52E-01 2.88E+00 1.47E+01

1. 04E-02 1.98E-04 4.38E-05 6.00E-07 9.62E-05 8.46E-03 5.97E-03 1.39E-02 8.27E-04 7.05E-03

3. 66E-02

(+E~uv i CA-spx;S!~

X"DI;^t\\O SE C-QwVlLErar XE-t3l )

Rev. 13 l

Attachment VI to WO 05-0025 Page 1 of 1 LIST OF COMMITMENTS The following table identifies those actions committed to by WCNOC 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.. Kevin Moles at (620) 364-4126.

COMMITMENT Due Date/Event The proposed changes to the Technical Specification Bases Within 90 days of and USAR will be implemented within 90 days of NRC NRC approval approval.

A revision to the fuel element failure Emergency Action Level Within 90 days of that reflects the approved TS 3.4.16 limits will be implemented NRC approval at the time the amendment is implemented.

WO 05-0025, Revision to Technical Specification (TS) 1.1, Definitions, and TS 3.4.16, RCS Specific Activity

Contents

Text

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

Response

Response

Response

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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WO 05-0025, Revision to Technical Specification (TS) 1.1, Definitions, and TS 3.4.16, RCS Specific Activity

Text

Subject:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

Response

Response

Response

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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