CPSES-200060390, Revision to Technical Specification 1.1, Definitions and 3.5.16, RCS Specific Activity

Revision to Technical Specification 1.1, Definitions and 3.5.16, RCS Specific Activity

ML060580225
Person / Time
Site:	Comanche Peak
Issue date:	02/21/2006
From:	Madden F TXU Power
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
00236, CPSES-200060390, LAR 06-001, TXX-06038
Download: ML060580225 (53)

Text

Power TXU Power Mike Blevins Ref: 10CFR50.90 Comanche Peak Steam Senior Vice President &

Electric Station Chief Nuclear Officer P. 0. Box 1002 (EO1)

Glen Rose, TX 76043 Tel: 254 897 5209 Fax: 254 897 6652 mike.blevinsitxu.com CPSES-2000600390 Log # TXX-06038 File # 00236 February 21, 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 LICENSE AMENDMENT REQUEST (LAR)06-001 REVISION TO TECHNICAL SPECIFICATION 1.1, "DEFINITIONS" and 3.4.16, "RCS SPECIFIC ACTIVITY"

Dear Sir or Madam:

Pursuant to I OCFR50.90, TXU Generation Company LP (TXU Power) hereby requests an amendment to the CF'SES Unit 1 Operating License (NPF-87) and CPSES Unit 2 Operating License (NPF-89) by incorporating the attached change into the CPSES Unit 1 and 2 Technical Specifications (TS). This change request applies to both Units.

The proposed change will revise TS 1.1 entitled "Definitions" and TS 3.4.16 entitled "RCS Specific Activity". This License Amendment Request 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 P - AVERAGE DISINTEGRATION ENERGY definition. In addition, the current DOSE EQUIVALENT I-131 definition (corresponding to the Iodine-131 isotope) would be revised to allow the use of alternate, NRC-approved thyroid dose conversion factors.

This change is being proposed in order to implement an RCS specific activity Limiting Condition for Operation (LCO) that reflects the whole body radiological consequence analysis assumptions. Those assumptions are sensitive to the noble gas activity in the primary coolant, but not to the other, non-gaseous activity currently captured in the current definition. The current E definition includes radioisotopes A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon

• Palo Verde

• South Texas Project

• Wolf Creek

TXX-06038 Page 2 of 4 that decay by the emission of both gamma and beta radiation. Current Ccndition 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 veiy low) if beta emitters such as H-3 and F-18 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, an objective which is not attained with the current construct of the LCO.

TXU Power 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, Wolf Creek Nuclear Operating Corporation, PG&E, STP Nuclear Operating Company, and Arizona Public Service Company.

PG&E's Diablo Canyon Power Plant is the lead STARS plant for this amendment request. 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 infornmation presented within brackets (i.e., within []) in Attachment 1 (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 API000 advanced reactor designs utilize an LCO for RCS I)OSE 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 AP 1000 in the NRC letter to Westinghouse Electric Company dated September 13, 2004. provides a detailed description of the proposed changes, a technical analysis of the proposed changes. TXU Power's determination that the proposed changes do not involve a significant hazard consideration, a regulatory analysis of the proposed changes and an environmental evaluation. Attachment 2 provides the affected Technical Specification (lTS) pages marked-up to reflect the proposed

TXX-06038 Page 3 of 4 changes. Attachment 3 provides proposed changes to the Technical Specification Bases for information only. These changes will be processed per CPSES site procedures. Attachment 4 provides retyped Technical Specification pages which incorporate the requested changes. Attachment 5 provides retyped Technical Specification Bases pages which incorporate the proposed changes.

TXU Power requests approval of the proposed License Amendment coincident with or after the approval of LAR 05-005 (TAC No. MC8163/8164), submitted via TXX-05127 on 08/22/05, to be implemented within 120 days of the issuance of the license amendment. LAR 05-005 contains the revised CPSES radiological consequence analyses referenced in Attachment 1. The approval date was administratively selected to allow for NRC review but the plant does not require this amendment to allow continued safe full power operations.

In accordance with I OCFR50.91 (b), TXU Power is providing the State of Texas with a copy of this proposed amendment.

Should you have any questions, please contact Mr. Robert A. Slough at (254) 897-5727 I state under penalty of perjury that the foregoing is true and correct.

Executed on February 21, 2006 Sincerely, TXU Generation Company LP By:

TXU Generation Management Company LLC Its General Partner Mike Blevins By: Au t-(

lt

irec, AW.

Maadsden Director, Regulatory Affairs

TXX-0603 8 Page 4 of 4 RAS Attachments

1. Description and Assessment

2. Proposed Technical Specifications Changes

3. Proposed Technical Specifications Bases Changes (for information)

4. Retyped Technical Specification Pages

5. Retyped Technical Specification Bases Pages (for information) c -

B. S. Mallet, Region IV M. Thadani, NRR Resident Inspectors, CPSES Ms. Alice K. Rogers Environmental & Consumer Sa:fety Section Texas Department of State Health Services 100 West 49th Street Austin, Texas 78756-3189

Attachment I to TXX-06038 Page 1 of 21 ATTACHMENT 1 to TXX-06038 DESCRIPTION AND ASSESSMENT

Attachment I to TXX-06038 Page 2 of 21 LICENSEES EVALUATION

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECEDENTS

8.0 REFERENCES

to TXX-06038 Page 3 of 21

1.0 DESCRIPTION

By this letter, TXU Generation Company LP (TXU Power) requests an amendment to the CPSES Unit 1 Operating License (NPF-87) and CPSES Unit 2 Operating License (NPF-

89) by incorporating the attached change into the CPSES Unit 1 and 2 Techrnical Specifications. Proposed change LAR 06-001 is a request to revise Technical Specifications (TS) 1.1, "Definitions" and 3.4.16 "RCS Specific Activity" for Comanche Peak Steam Electric Station (CPSES) Units 1 and 2. The proposed changes would replace the current TS 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 (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 all ow alternate, NRC-approved thyroid dose conversion factors.

No changes to the CPSES Final Safety Analysis Report are anticipated at this time as a result of this License Amendment Request.

2.0 PROPOSED CHANGE

The proposed change would:

2.1 Revise the definition of DOSE EQUIVALENT I-13 1 in TS 1.1 as follows:

"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 I-1 31, 1-132, 1-133, 1-134, and I-135 actually present.

The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or from 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]."

2.2 Replace the TS Section 1.1 definition for E - AVERAGE DISINTEGRATION ENERGY with the following new definition for DEX:

"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 [Ar-4 1, Kr-83m, Kr-

Attachment I to TXX-06038 Page 4 of 21 85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xc-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 detennination of DOSE EQUIVALENT XE-133 shall be performed using [effective dose conversion factors for air submersion listed in Table III. 1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil", or using the dose conversion factors from Table B-I of Regulatory Guide 1.109, Revision 1, NRC, 1977.]

2.3 Revise TS Limiting Condition for Operation (LCO) 3.4.16, "RCS Specific Activity," 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."

2.4 Delete the current TS Figure 3.4.16-1, "Reactor Coolant DOSE EQU[VALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER".

2.5 Revise the Applicability of TS 3.4.16 from:

"MODES I and 2, MODE 3 with RCS average temperature (Tavg) > 5 000F" to "MODES 1, 2, 3, and 4."

2.6 Revise TS 3.4.16 Condition A from:

"DOSE EQUIVALENT I-13:L > 0.45 pCi/gm."'

to "DOSE EQUIVALENT I-13 not within limit."

2.7 Revise TS 3.4.16 Required Action A.I from:

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

Attachment I to TXX-06038 Page 5 of 21 to "Verify DOSE EQUIVALENT 1-131 < 60 piCi/gm."

2.8 Revise TS 3.4.16 Condition 13 from:

"Gross specific activity of the reactor coolant > I 00/f ptCi/gm."

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

2.9 Add a Note to state that LCO 3.0.4.c is applicable to TS 3.4.16 Required Action B.I and revise TS 3.4.16 Required Action B.1 from:

"Be in MODE 3 with Tavg < S00°"

to

"----------------NOTE--------------

LCO 3.0.4.c is applicable.

Restore DOSE EQUIVALENT XE-133 to within limit."

2.10 Revise TS 3.4.16 Required Action B.1 Completion Time from "6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />" to "48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />".

2.11 Revise TS 3.4.16 Condition C from:

"Required Action and associated Completion Time of Condition A not met OR DOSE EQUIVALENT I-13 1 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."

2.12 TS 3.4.16 Required Action for Condition C would be revised from:

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

to

Attachment I to TXX-06038 Page 6 of 21 "C.1 Be in MODE 3."

AND C.2 Be in MODE 5."

2.13 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 <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />". The completion time for Required Action C. I would remain 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.14 Add a Note to state that Surveillance Requirement (SR) 3.4.16.1 is only required to be performed in MODE I and revise SR 3.4.16.1 from:

"Verify reactor coolant gross specific activity < 100/E [tCi/gm."

to

---

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

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity < [500]

PCi/gm."

2.15 Delete the current SR 3.4.16.3.

In summary, the proposed changes will revise the definition of DEI, delete the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition for DEX, revise TS 3.4.16 to specify an LCO limit on DEI, add a new LCO 3.4.16 limit for DEX, increase the completion time of 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 I entry, and SR 3.4.16.3 is deleted.

The TS Bases for LCO 3.4.16 would be revised to reflect the proposed changes and to incorporate the bases for the proposed changes. The proposed TS Bases changes are included for information only in Attachment 5. The proposed TS change is noted on the marked up TS pages provided in Attachment 2. The proposed retyped TS pages are provided in Attachment 4.

Attachment I to TXX-06038 Page 7 of 21

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) j'outside containment] accident since these events involve the release of primary coolant activity [outside containment]. 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 the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> immediately following an accident, or at the low population zone outer boundary for the duration of a radiological release, is specified in 10CFRI00. I1. 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.

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 [9 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-Waler Nuclear Power Reactors."

The SGTR and MSLB radiological consequence analyses establish the acceptance limits for the TS 3.4.16 limits on RCS specific activity. These analyses consider two cases of RCS iodine specific activity. Case I assumes that an accident-initiated iodine spike occurs which results in an increase in the rate of iodine release from those fuel rods containing [preexisting] cladding defects to the primary coolant immediately after [the postulated] MSLB or SGTR. Case 2 assumes that a pre-accident iodine spike occurs due to a transient prior to the MSLB or SGTR. The results of the SGTR radiological consequence analyses are described in [FSAR Chapter 15.6.3]. The results of the MSLB radiological consequence analyses are described in [FSAR Chapter 15.1.5].

to TXX-06038 Page 8 of 21

[The Case I radiological consequence analyses for SGTR and MSLB assume that the initial reactor coolant iodine specific activity corresponds to an isotope mixture that bounds the proposed LCO 3.4.16 limit. The initial isotopic mix is taken from FSAR Table 15.1-4, for the SGTR analyses, and from FSAR Table 15.1-4A, for the MSLB analyses. Note that FSAR Tables 15.1-4 and 15.1-4A will be combined in new FSAR Table 15.1-4 upon approval and subsequent implementation of LAR 05-005 (Reference 8.13, TAC No. MC8163/8164) which was submitted via TXX-05 127 on 08/22/05.]

This analysis assumption provides the basis for the iodine specific activity limit of 0.45 ltCi/gm contained in current TS 3.4.16 Condition A and SR 3.4.16.2.

Thyroid dose conversion factors based on [Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites,"] 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 I-131 may be used in future analyses after this amendment is approved. The radiological consequence analyses submitted with LAR 05-005 (Reference 8.13) has been updated to use the Thyroid Dose Conversion Factors in ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers", 1979.

Case 1 also assumes an accident-initiated iodine spike that increases the rate of iodine release from the fuel rods containing [preexisting] cladding defects to the primary coolant immediately after [the postulated] MSLB or SGTR. The iodine spiking factor is assumed to be 500 for the Case I radiological consequence evaluations for MSLB. The iodine spiking factor is assumed to be 325 for the Case 1 radiological consequence evaluation for the SGTR.

The Case 2 radiological consequence analyses for SGTR and MSLB assume the initial reactor coolant iodine specific activity is a factor of 60 higher than Case I due to a pre-accident iodine spike caused by a transient prior the accident. This bounds the allowable RCS specific activity value of 63) pCi/gm contained in current TS Figure 3.4.16-1 for RATED THERMAL POWER (RTP) between 80% and 100% RTP. 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 Cases I and 2 radiological consequence evaluations for SGTR and MSLB, the noble gas specific activity in the reactor coolant is assumed to be that associated with 1% fuel defects (increased by 10% for conservatrsm). This Noble Gas isotopic mix is equivalent to [715] ptCi/gm DEX. The initial DEX concentrations were calculated using [whole body dose conversion factors for air submersion based on Table III. I of EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil"].

to TXX-06038 Page 9 of 21 3.2 RCS Specific Activity The RCS specific activity level is used in design basis accident analyses to determine the thyroid and whole body radiological consequences of accidents that involve the release of RCS activity. For events that also include fuel damage, the dose contribution from the initial activity in the RCS is insignificant.

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

additional NRC-approved source[s] of thyroid dose conversion factors [are] being added to the revised definition.

LCO 3.4.16 specifies the limit for RCS gross specific activity as I100/E [tCi/gm.

E is defined as:

"B 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 > [10] minutes, making up at least 95% of the total non-iodine 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 I00/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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. As currently presented, the value for the maximum allowable iodine concentration during the 48-hour period of elevated activity is a function of power level as provided in TS Figure [3.4.16-1]. In accordance with the figure, as power is reduced below 80% RTP, the allowable RCS iodine concentration increases from 60 pCi/gm DEI to as high as [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.

Attachment I to TXX-06038 Page IO of 21 3.3 Purpose of 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 tritium (H-3) and Fluorine-18 (F-18) are included in that value, as required by the E de Finition. [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 I 00/B 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, [an objective which] is not attained with the current construct of the LCO.]

4.0 TECHNICAL ANALYSIS

4.1 TS Changes 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.

[Two] new thyroid dose conversion factor reference[s are] added to the definition.

The new reference[s are] ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," and Table 2.1 of EPA Federal Guidance Report No. I 1, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and I)ose Conversion Factors for Inhalation, Submersion, and Ingestion. EPA Federal Guidance Report No. II is referenced in Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003, Section C, "Regulatory Position," Subsection 4, "Dose Calculational Methodology,"

Subsection 4.1, "Offsite Dose Consequences," assumption 4.1.2 as acceptable for determining thyroid dose from inhalation. [The thyroid dose conversion factor values contained in Table 2.1 of EPA Federal Guidance Report No. 11 are to TXX-06038 Page 11 of 21 provided to three significant digits. The thyroid dose conversion factor values contained in ICRP 30 Supplement to Part 1, pages 192-212, and Table 2.1 of EPA Federal Guidance Report No. II are the same when the EPA Federal Guidance Report No. 1 values are rounided to two significant digits.]

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

Deletion of Definition for E - AVERAGE DISINTEGRATION ENERGY and Addition of New Definition for DEX The current TS I.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 plart operation there are typically only a small number 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 from 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 E are determined resulting in different values for the gross specific activity limit (100/E pCi/gm).

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 B - AVERAGE DISINTEGRATION ENERGY and addition of a new definition for I)EX will result in TS 3.4.16 requirements for RCS specific activity which are consistent with the assumptions contained in the radiological consequence analyses.

Attachment I to TXX-06038 Page 12 of 21 The new definition for DEX is similar to the definition for DEI. The determination of DEX will be perfonned 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, Xc-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 factor. 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 shal I 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," 1993 [or using the dose conversion factors from Table B-I 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 III. I of EPA Federal Guidance Report No. 12 is endorsed by Regulatory Guide 1.195, Subsection 4.1, assumption 4.1.4 as acceptable for determining whole body doses because of the uniform body exposure associated with semi-infinite cloud dose modeling.

TS 3.4.16 LCO Revision The TS 3.4.16 LCO is modified to specify that the iodine specific activity, in terms of DEI, and noble gas 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 1DOSE EQUIVALENT XE-133 specific activity shall be within limits". The DEI limit of

< [0.45] jtCi/gm is contained in current Condition A and SR 3.4.16.2. In addition, the limit of [0.45] gCi/gin is consistent with the current SGTR and MSLB radiological consequence analyses discussed in Section 3.1 above.

The DEX limit of [500] [Ci/gm is contained in revised SR 3.4.16.1 and is

[consistent with the current SGTR and MSLB radiological consequences analyses discussed in Section 3.1 above.]

to TXX-06038 Page 13 of 21 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 analyses and would not fluctuate with variations in the calculated value of E during normal operation as is currently the case.

TS 3.4.16 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 I 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 fo: decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal [or non-existent]. 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 "> 0.45,uCi/gm" with the words "not within limit" to be consistent with the revised TS 3.4.16 LCO format.

The DEI limit of < 0.45 pCi/gm is contained in SR 3.4.16.2.

TS 3.4.16 Required Actionrl A.I [I Revision TS 3.4.16 Required Action A. I is modified to remove the reference to Figure

[3.4.16-1] and insert a limit of less than or equal to 60 paCi/gm for D'I.

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 pennitted 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 I analysis assumption, [which corresponds to the 60 pCi/gm specific to TXX-06038 Page 14 of21 activity limit associated with 100% RTP operation as discussed in Section 3.1 above.] Therefore, TS 3.4.16 Required Action A. l should be based on a limit of 60 gtCi/gm to be consistent with the assumptions contained in the radiological consequence analyses. It is rot expected that plant operation at reduced power levels would result in iodine specific activity levels that exceed the 60 [LCi/gm upper limit defined for full power operation.

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 TS 3.4.16 LCO which requires the DEX specific activity to be within limits as discussed above. The DEX limit of [500] jiCi/gm is contained in the revised SR 3.4.16.1.

The limit of [500] [tCi/gm is [consistent with the current SGTR and MSLB radiological consequence analyses which assume the steady state initial RCS noble gas specific activity is that associated with 1% fuel defects for both the Case I and Case 2 analyses.] 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 the revised TS 3.4.16 Required Action B. 1 will require restoration of DEX to within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This is consistent with the completion time for current Required Action A.2 for DEL. [Since the radiological consequences reported for MSLB and SGTR 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 the 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for the revised Required Action B. l 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 modes(s), relying on Required Action B. I 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 which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

to TXX-06038 Page 15 of 21 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 ptCi/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.l is changed to require the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and a new Required Action C.2 is added which requires the plant to be in Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. These changes are consistent with the changes made to the TS 3.4.16 Applicability. The revised LCO is applicable throughout all of Modes I through 4 to limit the potential radiological consequences of an SGTR or MSLB that may occur during these Modes. Therefore, Condition C needs to default to a Mode 5 end state for TS 3.4.16 to no longer be applicable.

A new TS 3.4.16 Required Action C.2 Completion Time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 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 <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> 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 is < [500] jiCi/gm. This change provides a surveillance for the new LCO limit added to TS 3.4.16 for I)EX.

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 sun 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 I. I is not detecited, it should be assumed to be present at the minimum detectable activity. This is consistent with the new TS I.I Definition for DEX and will ensure a conservative calculation of DEX when noble gas nuclides are not detected.

to TXX-06038 Page 16 of 21 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 those modes, prior to entering Mode 1, similar to the current surveillance SR 3.4.16.2 for DEI.

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 detennine 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 jiCi/gm DEI). This concentration is more restrictive on plant operation than the current LCO which allows operation up to [300] [tCi/gm DEI as indicated in Figure 3.4.16-1. The proposed changes eliminate the potential for the radiological consequences of a postulated accident to exceed those previously calculated.

4.3 Summary In summary, the proposed changes will revise the definition of DOSE; EQUIVALENT I-131, delete the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition for DOSE EQUIVALENT XE-133, revise TS 3.4.16 to specify an LCO limit on DOSE EQUIVALENT I-131, add a new LCO limit to 'S 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.

Attachment I to TXX-06038 Page 17 of 21 The revised definition of DOSE EQUIVALENT I-131 allows the use of thyroid dose conversion factors which are acceptable for determining thyroid dose. The above changes will result in TS 3.4.16 requirements for RCS specific activity which are consistent with the assumptions contained in the radiological consequence analyses. The primary purpose of the TS 3.4.16 LCO o01 RCS specific activity is to support the dose analyses for design basis accidents, in which the whole body dose is; primarily dependent on the noble gas specific activity, not the non-gaseous activity currently captured in the E definition. The TS 3.4.16 Conditions, Required Actions, and Surveillance Requirements are revised accordingly to support the deletion of the requirements for gross specific activity based on E and the addition of the new LCO limit for DOSE EQUIVALENT XE-133. The proposed changes do not impact the radiological consequences of any design basis accident.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration

[TXU Power] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10CFR50.92, "Issuance of amendment," as discussed below:

1.

Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response

No The proposed changes to add new thyroid dose conversion factor reference[s] to the definition of DOSE EQUIVALENT 1-13 1, eliminate the definition of E - AVERAGE DISINTEGRATION ENERGY, add a new definition of DOSE EQUIVALENT XE-133, replace the Technical Specification (TS) 3.4.16 limit on reactor coolant system (RCS) gross specific activity with a limit on noble gas specific activity in the form of a Limiting Condition for Operation (LCO) on DOSE EQUIVALENT XE-133, replace TS Figure 3.4.16-1 with a maximum limit on DOSE EQUIVALENT I-131, extend the Applicability of LCO 3.4.16, and make corresponding changes to TS 3.4.16 to reflect all of the above are not accident initiators and have no impact on the probability of occurrence for 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 to TXX-06038 Page 18 of 21 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.

Do the proposed changes 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 assurned in the current radiological consequence analyses.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

Do the proposed changes 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 more restrictive on plant operation than the existing allowable RCS iodine specific activity at lower power levels.

Therefore the proposed change does not involve a reduction in a margin of safety.

Based on the above evaluations, [TXU Power] concludes that the pro Dosed amendment(s) present no significant hazards under the standards set forth in I OCFR50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.

Attachment I to TXX-06038 Page 19 of21 5.2 Applicable Regulatory Requirements/Criteria The regulatory guidance documents associated with this LAR 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.

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) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

[TXU Power] has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection or surveillance requirement.

[TXU Power] has evaluated the proposed changes and has determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amount of effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure.

Attachment I to TXX-06038 Page 20 of 2]

Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in I OCFR51.22 (c)(9). Therefore, pursuant to IOCFR51.22 (b), [an environmental assessment of the proposed change is not required.]

7.0 PRECEDENT The Technical Specifications developed for the Westinghouse AP600 and APlI000 advanced reactor designs utilize an L[CO 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 ftor the API000 in the NRC letter to Westinghouse Electric Company dated September 13, 2004.

The curve in current TS Figure 3.4.16-1 was not included in the TS approved for the AP600 and API000 advanced reactor designs.

8.0.

REFERENCES 8.1 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.

8.2 Environmental Protection Agency (EPA) Federal Guidance Report No. 12, EPA-402-R-93-08 1, "External Exposure to Radionuclides in Air, Water, and Soil,"

1993.

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

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

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

8.6 Regulatory Guide 1.109, Revision 1, "Calculation of Annual Doses lo Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," October 1977.

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

to TXX-06038 Page 21 of 21 8.8 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.

8.9 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.

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

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

8.11 NUREG-1 512, "Final Safety Evaluation Report Related to the Certif cation of the AP600 Standard Design, Docket No.52-003," August 1998.

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

8.13 TXU Power letter, logged TXX-05127, "License Amendment Request (LAR)05-005, Revision to Technical Specification 3.7.10, Control Room Emergency Filtration/Pressurization System (CREFS)", from Mike Blevins to USNRC, dated August 22, 2005.

to TXX-06038 Page 1 of 7 ATTACHMENT 2 to TXX-06038 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

Pages 1.1-3 3.4-44 3.4-45 3.4-46 3.4-47 INSERTS to TXX-06038 Page 2 of 7 1.1 Definitions (continued)

Definitions 1.1 DOSE EQUIVALENT 1-131 JINSERT 1 POS E EQUIVALENT 1-131 shall be that concentration Af

-,.4-1-31 (microcuriGs/gram) that alone would produce the same thYoid dose as the quantity and isotopic mixturWe of 1 131, 4-1 1133, 1 134, and 1 135 atually prosent The thyFeid dCoe-nvnversion factors used for this calculation shall-be theoA listedr in Tabe hlll of TID 11 811, AE44 1AEG Calculation elf istance Factors for Power and Test React tes," or thAses listed in Table E 7 of Regulatory Guide 1.109, Rev. 1, NRCT" IA 7 IINSERT 2P 1

^-shaII be the average (weighted in proportion to the GORGentration of each radionuclide in the reaWmto.-GeeaRt at the-tine of sampling) of the sum of the average beta aRd e cnorgitcs pcr disintcgration (in M oA) nF, ieetepe mthpr-thanRiodinre, with half lies.-_ o10 minRutes nakirg up at least )5% of the total noniodine activity in the cEoIant.

ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME 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 methodology for verification have been previously reviewed and approved by the NRC.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 1.1 -3 Amendment No. 64 to TXX-06038 Page 3 of 7 RC'S Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 Th-pn-nIfbn.r+R-0+,,

pf flqa r-ncnnnr 31:an+ qh-q11 him wifhin limit--

RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity 11 to.

APPLICABILITY:

MODES 1-and-2 RACnn= '

-4k DC'(Q IT

\\ ~,

1rfloC A.j

-M "I

\\

-a*/

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT 1-131

---

Note--------------------

> 0-4 m

LCC) 3.0.4.c is applicable.

A. 1 Verify DOSE Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> EQUIVALENT 1-131 Awithin the aceptab

<6 region of Fmigure 3.4.16- -1.

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

14-O (continued)

COMANCHE PEAK - UNITS I AND 2 3.4-44 Amendment No. 44)0 to TXX-06038 Page 4 of 7 RC'S Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME lDOSE EQUIVALENT XE-133 not within limit D f-

--- fs,

__+;-,;t+ of no ;- h^nnc Q.^,;+k LD.

A, the reactor coolant s! Inns ttoikjR4 N SERT 3,

10T g hours 4-4 C. Required Action and associated Completion Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 in tho unac cptablqjeg ion of Figur 3.1.16 1.

1> 60 gLCi/gmI C.1 Be in MODE 3 With Ta e 500F a

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Be in MODE 3 AND C'.2 Be in MODE 5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolar 9s pecific activity 7 days

<4 0QjE pCi/gm.

/

50 IDOSE EQUIVALENT XE-1331 I

(continued)

Only required to be performed in MODE 1.

COMANCHE PEAK - UNITS 1 AND 2 3.4-45 Amendment No. 64 to TXX-06038 Page 5 of 7 RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.2

---

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

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 0.45 pCi/gm.

14 days AND 1 102 Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWNER change of Ž: 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period SR 3.4.16.3 iN l

t Not required to be performed until 31 days after a minimum of A2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for Ž 18 Gos Dotorminc E from a samrple taken in MODE 1 after a minimum of 2 effective full power days and 20 days of h AInrI nnIr.rninn k-%

hn

,t

-np-rl einrp th-r-,nrtnr wm 484 days I^ -

A-..

ow vl v

VfWAc-~

a rvs Vl i

wA Z-

-+1 Fo An L ------

ias1t Sufr4Iti:iI

'nr lo rag COMANCHE PEAK - UNITS 1 AND 2 3.4-46 Amendrnent No.40 to TXX-06038 Page 6 of 7 RG peGif4G Atl-t PERcENT or RATED THERMAL POWER grev 3.4.16 1 (page 1 of 1)

Reactor Coolant DOSESEQUIVALENT! 131 Specific Activity I imit Vorsus Porcont of RATED-THPERIW4,AL POWER I-IRAA K If'I 1r MrAIZ I IKIITOe it AKIl% ^,

'2 A A7 A A_# A_

Kl-EVA I r : ~ r I J I I I

~ ~ I

.~

' I

~r r ~ ~ r I 1 I 0

to TXX-06038 Page 7 of 7 INSERTS INSERT 1 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same close 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 Table IlIl of TID-14844, AEC, 1 962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP-30, 1979, Supplement to Part 1, page 192-212, Table tilled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or from 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 2 DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (microcLiries 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-1 33m, Xe-1 33, Xe-135m, Xe-1 35, and Xe-1 38 actually present. If a specific noble gas nuclide s not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil", or using the dose conversion factors from Table B-1 of Regulatory Guide 1.109, Revision 1, NRC, 1977.

INSERT 3

---

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

LCO 3.0.4.c is applicable.

B.1 Restore DOSE EQUIVALENT XE-1 33 to within limit.

to TXX-06038 Page 1 of 9 ATTACHMENT 3 to TXX-06038 PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES (Markup For Information Only)

Pages B 3.4-101 B 3.4-102 B 3.4-103 B 3.4-104 B 3.4-105 B 3.4-106 INSERTS INSERTS to TXX-06038 Page 2 of 9 RC'S Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES

ThJKing

[INSERT 1 BACKGROUND The maximum dose to thL whole body and th t yroid that a individual at t

e boundary can reive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> durThg an acciden s specified Exnlusion (Ref. 1). The limits on specific activity ensure that the are hoto a-small fraction of t 10 GFR I100- imits during 10011 analyzed transients and accidents.

\\v: --

-- ely limitedi The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) or a main steam lire break (MSLB) accident.

PDOSE EQUIVALENT XE-133 S-The ontains specific activity limits for both DOSE EQUIVALENT 1-1 31 and P c

GifiG aGtivity. The allowable levels are intended to NSRT3 limit tho 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the site boundary to a small fraction ef the 10 CFR 100 dose-giledimFts. The limits in the LCO are specific to CPSES due to th im>nlementafion of the iternate steam generator tube repair criteria.

j,or more conservative than, 1 25 25 offsite and Control RoomI

/

APPLICABLE SAFETY ANALYSES The LCO' orthe sp c

the resulting or dogs at4h Lctivity of the reac/r coolant ensures that ot i

a k

A A

i H im H 1631LIt; J-uurARi.

'..IIi RR)L fX ::(PIfR :i sRn4IIi j

fraction of the 10 CIF);100 des guideline imits'folfowing a SGTR or a MSLB accident. 1,Ife SGTR sa ety analysis (Ref. 2) assumes the specific activity of the re ctor coolant a the LCO limit and an existing reactor coolant steai generator (SG) tube leakage rate of 1 gpm. The MSLB safety anaqlis (Ref. 3) assumes the specific activity of the reactor coolant athfie LCO limit and an existing reactor coolant steam generator (SG) tube leakage! rate of 27.8 gpm in the affected steam generator and 450 gpm combined in the unaffected steam generators. The safety analysis for both accidents assumes the specific activity of the secondary coolant at its limit of 0.1 pCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.18, "Secondary Specific Activity."

25 1 25 (continued) 1 The referenced safety analysis reports the doses for the SGTR assuming DEI-1 31 to ba 1.0 pCi/gm.

The doses reported using this value have been shown to be conservative relative to those that would be calculated using a DEI-131 value of 0.45 pCi/gm.

2 To obtain the maximum benefit from the steam generator alternate repair criteria, the MISLB radiological consequences analysis assumes a leak rate to the faulted steam generator during the accident that results in calculated consequences approaching a small fraction (10%) of the 10 CFR 100 guideline values for the accident initiated spike. This leak rate provides a maximum primary-to-secondary leak rate limit against which the predicted end-of-cycle leakage is compared.

25 COMANCHE PEAK - UNITS 1 AND 2 B 3.4-1 01 Revision 2a to TXX-06038 I

RC'i Specific Activity Page 3 of 9 B 3.4.16 and SGTR BASES APPLICABLE SAFETY ANALYSES (continued)

The analysis for the MSLBaccidents establishes the acceptance limits for RCS specific activity. However, the SGTR accident analog consequences are gignificant. Reference to these analyses are used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

Each of the above analyses must consider two cases of reactor coolant specific activity. One case assumes specific activity at 0.45 pCi/gm r -.

DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that nINSERT 51 icr the-11 1 activity in the reactor coolant by a factor of about 500 immediately after the acGident. The second case assumes the initial reactor coolant iodine activity at 60.0 pCi/gm DOSE EQUIVALENT 1-131 due to a pre-acciclent iodine spike caused by an RCS transient. In both cases the nobl man cv ithe reactoFr coolant des --is IINSERT 6

W;OI4;8RlY en Is Iase nthe fraction of f,21ueldG 17uL=

0i L

~WIN ll.

rW fd!u cd toCi/gm DOSE EQUIVALENT I 131. 1he TCO limit of a~ayse&.

These analyses also assume a loss of offsite power at the same time as the SGTR or the MSLB event. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low rhessuizer pressure signal or an RoS overtemperature N-16 signal.

he MSLB causes a reduction in reactor coolant temperature and pressure. The temperature decrease causes an increase in reactor power. The power increase will trip the reactor on high neutron flux or overpower N-16. The pressure decrease will initiate a reactor trip on either low pressurizer pressure or the safety injection signal initiated by low pressurizer pressure, low steam generator pressure, or high containment pressure.

For the SGTR and the MSLB, the coincident loss of offsite power causes the steam dump valves to close to protect the condenser. For the SGTR, INSERT 7 a rise in pressure in the ruptured SG discharges radioactively ontaminated steam to the atmosphere through the SG atmospheric relief vWand the main steam safety valves. A failure to close of the ato relief valve on the affected SG is also assumed. The un

~

move core decay heat by venting steam to the atmosphere ends. For the MSLB, an uncontrolled (i.e, released to atmo ere) blowdown of only one steam generator is assumed. The unaff ed SGs remove core decay heat by venting steam to the atmosphere until th coodown ends. Radioactively contaminated steam is released to the atmosphere through the faulted.3G as well as the intact SGs assuming the primary to secondary leak rates shown above.

25 25 39 25 25 25 The applicable safety analysis shows the radiological consequences of either an SGTR or an MSLB accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not

_xce.the imtitsshown in Figure 3.1.16 1, in the applicalietspecificatin.

for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The safety analysis has concurrG_-.An4d pre accden iod spiking levels up to 60.0 pCi/gm DOSE EQUIVALENT l

'&A vdR spkR

.v 1 25 60.0 gCi/gm DOSE EQUIVALENT I-1 31 (continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.4-102 Revision 39 to TXX-06038 Page 4 of 9 RCS Specific Activity B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued)

Tho nemi;fnder of thp RhoyoA l~irt oemiihe oine levpUl swn on

- - I - "', " - I- -

+-k1-

^f +k-I-,

-f Cr_-rD

lAZ ItA

'4 z

n A n

-l CtIt7_CrCrt7IVW, -,

I --

v I-Or WSTLB accIgeni occurrIno auFRGn Ine ISISa2IisneG

, SAI

.s P.

o j -Vty Ul v

-W I rT-4 F hA =r time limit. The 125 e-pe rmip ib levele

-;I k-

-+k; occurrence oT an SGT or an MSLB accioent at thc on-.1, inrrA-.

th+o ritA

.lr-A-rA.

..1r--^

1,-t +

'U I I v vv It:;v-Is r

t Dan X v-t-----

10 CFR 100 dose ^,uideline limits.

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

RCS specific activity satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).

/ -/in the reactor coolant The e

i activity I imited to 0.45 pCi/gm DOSE EQUIVALENT 25 1-131, a rh s p ecific acti-ity in the ractor coolanF i lim itted to the

_ l.

- vrr.__ _Iv--v

.- IJ 1 -

4v rWf w_J

.: f--A

.v eL I LCO WI11DfU u11 ri U1i1-IUU'IU V

-(J UJ U

11W44fr~-

INERT 8 energy of the sum of the avera;ge beta and gamma energies of the coolant RsUlides} Thp irnit on DO SEE EQUIVALENT I 131 enSUr4-thc 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose to an individual at the site boundary during tl Accident (DBA) will be within the allowed thyroid dose gu deline. Since the nahobeasatvtinteratrcoatdin airailaclaavo 39 v

ic I-c.

I-

-1.1 pGi/gm DOSE EQ'UIVALENT 1 131, the limit on DOSE EQUIVALENT l131 also ensures the : hour whole body dose to an individual-at the site boundary during the DBA will be a small fraction of the allowed whole bed" dose. The limit oln aross AeeAifin activitv is not directly naplicnhIp to anv I

11-_

I I-calculatedl The SGTR accident analysis ef. 2) and the MSLB accident analysis (Ref. 3) show that the 2 heu site beundary dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of ap SGTR, lead to site boundary doses that exceed the 10 CR 1001 dhe guideline limi4s.

1 25 COMANCHE PEAK - UNITS 1 AND 2 B 3.4-103 Revision 38 to TXX-06038 Page 5 of 9 RC"; Specific Activity B 3.4.16 BASES (continued:

APPLICABILI ITY In MODES 1 is-2, and in MODE 3 with RCS average t4Ffnpenatur e

~' 5000F, operation within the LCO limits for DOSE EQUIVALENT 1-1 31 is nw~ess sa~

aki the potential consequences of an SGTR and an i

MSLB to within 0# acceptable site boundary dose values While not

+e-u VAab y CPSES radiological accident aRalyset-414eL appw_

a y defined above for DOSE EQUIVALENT I I131 is also applied hpcfic activity (100/ E PCi/gm) LCO.

ISRP acceptance criterial INERT For operation in AOng:DE 3 with RcS average temperatu re 4 5001F, and in

'MODES 4 and 5, the offsite release of radioactivity in the event ef a SGTR is unlikely since the saturation pressure of the reactsocoanti-s below the lift pressure settings of the main steam safety -ind relief valves.

cFa oepeatieR MAQDEa 3 with RCS average temperaturc 5000F, a-d in MODES 4 and 5, the offsito release of radioactivity in the event ef aR MSLB is unlikely. This is because the saturation pressure of the reactor coolant is sufficiently low that the primary to secondary pWe6seue difereRnce created w

%hen the affected SG deprsuizos i t -Iae enough to cause significant SG tube damage. Also, for th-unaffeted SGs the offsite release of radioactivity in the event of an MgSB-is nikely Rinse the saturation pressure of the reactor coolant is bolowA the lift 39 1 25 presr_

setti w

---steam safety and relief va-vas ACTIONS A. 1 and A.2 ISERT 10 A Note permits the use of the provisions of LCO 3.0.4.c. This allowanc permits entry into the applicable MODE(S) while relying on the ACTION%.

This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an even which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be taken to demonstrate that th i~eimts ef Figure 3.1.16 1 are not exceeded. The Completion Time of hours is required to obtain and analyze a sample. Sampling is done t continue to provide a trend.

specific activity is < 60.0 Ci/gm The DOSE EQUIVALENT 1-131 must be restored to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Ccmpletion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> iq equired, if the limit violation resulted from normal iodine spiking.

34 JIN E1III 16 25 (continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.4-104 Revision 39 to TXX-06038 Page 6 of 9 RCS Specific Activity B 3.4.16 BASES ACTIONS (continued)

B.1 With the gross specific activity in excess of the allowed limrit, the unit must jINSERT 12 I ie placed in a MODE in which the requirement does not apply.

I ---

ah ne canace withir b hours to MODu vi and Ads average temperature JINSERT 13l A

A s

-<50uOl= lowers V; e-saturati4n pressure of the reactor coo rant to an 25 d Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable

_1 Ai

-i

-o

.-.,A I..1

___1l 1

_1.-_11.

aGGeptable level.

I+/-Reallowei hased on operatina( experieRce. 10 FeaGn NMUUA=.i 9eleW-aUUi-` fFrom Wui

-r -I-- -- --

r nower conditions in an orderly manner and without challcaqgin elan V

/

Iand C.l 2>

60.0 Cia d

If quired Action and the associated Comple ion Ti e of Condition A is not met or if the DOSE EQUIVALENT 1-131 i

-a-e-table region ef Figure 3.41.6 1, the reactor must be brought t6 MODE 3 wi4 RCS average t emperatue-00-OF within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ThevCompletion Times reasonable, based on operating exp rience, to reach bfrom full power conditions nan orderly manner and out challenging p ant systems.

an DE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> ar te reouired niant conditions i

-- -- r, SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of theqeS specific activity of the reactor coolant at least once every ys. While ba ;rially a quantitative maeasure of radionniulides with half lives longer than 10 minutes, excluding iodines, JThis measurement is the noble gas sum of the degassed gamma activities and the gaseous gamma activities mree taken. This Surveillance provides an indication of any increase in specific activity.

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

The SuCr'eillrne is applicable in MODES 1 and 2, and in MOE 3 with T~-sgeaten than equal te 5002 F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during the time.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.4-105 Revision 2a to TXX-06038 Page 7 of 9 RC'S Specific Activity B 3.4.16 BASES SURVEILLANCE SR 2

specific a REQUIREMENTS 0

(continued)

This Surveilla ce is performed in MODE 1 only to ens re iodin emains witbiv limit ring normal operation and following fast p/ver changes he n

fuel fa luee is more apt to occur. The 14 day Frequ ncy is adequate to trend changes in the iodine activity level, considering gFess activity is monitored every 7 days. The Frequency, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power change 2 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following fuel failur; samples at other times would provide inaccurate results.

Tiodinespike initiation SR 3.4.16.3 A radiochemical analysis for E determination is required c.we.484-day6 I(6 months) with the plant operating in MODE 1 equilibrium conditions.

The E detormination directly relates to the LCO and is required te verify plant operation within the specified gross actiV;ty LCO "iFmitb The aRaInY for E is a measurement of the average energies per disinegratioR fo isotopetewit lhalives longer than 10 minutes, excluding-iTdines. The Fr=equency of 184 days reoGgni_;zer d-oe noRt chang rFaiAdly This SR has been modified by a Note that indicates sampic ris-requiFed to be performed WithiR 31 days after a minimum of 2 e A'ie-full -pew days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />. This ensures that the radioactive materials are at equilibriuim so the flysior f

is representative andi net skewed by a crud burst or ether similar nabnorma eveRtm REFERENCES

1.

10 CFR 1()0.11, 1973.

2.

FSAR, Section 15.6.3.

3.

FSAR, Section 15.1.5 1 25 COMANCHE PEAK - UNITS 1 AND 2 B 3.4-106 Revision 2X to TXX-06038 Page 8 of 9 INSERTS INSERT 1

,or at the Low Population Zone outer boundary for the radiological release duration, INSERT 2 Doses to the Control Room operators must be limited per GDC 19.

INSERT 3 ensure that offsite and Control Room doses meet the appropriate acceptance criteria in the Standard Review Plan.

INSERT 4 meet the appropriate Standard Review Plan acceptance criteria.

INSERT 5

,by a factor of [500 or 335], the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a MSLB or SGTR, respectively.

INSERT 6 specific activity is assumed to be the equivalent of 1% fuel defects which corresponds to [715]

pci/gm DOSE EQUIVALENT XE-133.

INSERT 7 RHR system is placed in service.

INSERT 8 noble gas specific activity in the reactor coolant is limited to [500] pci/gm DOSE EQUIVALENT XE-133, as contained in SR 3.4.16.2 and SR 3.4.16.1 respectively. The limits on specific activity ensure that offsite and Control Room doses will meet the appropriate Standard Review Plan acceptance criteria.

INSERT 9 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, monitoring of RCS specific activity is not required.

to TXX-06038 Page 9 of 9 INSERTS INSERT 10 Required Actions A.1 and A.2 while the DOSE EQUIVALENT 1-131 LCO limit is not met.

INSERT 1 1 acceptable since it is expected that, if there were no iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.

INSERT 12 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.

INSERT 13 A NOTE permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

INSERT 14 If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-1 3:3 in Specification 1.1, Definitions," is not detected, it should be assumed to be present ait the minimum detectable activity.

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

INSERT 15 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.

to TXX-06038 Page 1 of 5 ATTACHMENT 4 to TXX-06038 RETYPED TECHNICAL SPECIFICATION PAGES Pages 1. 1-3 3.4-44 3.4-45 3.4-46 to TXX-06038 Page 2 of 5 Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT XE-1 33 ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME 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 Table IlIl of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity," or from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radicnuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersicn, and Ingestion."

DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-i 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-133 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 effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, "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.

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

(continued)

COMANCHE PEAK - UNITS 1 AND 2 1.1 -3 Amendment No. 64 to TXX-06038 Page 3 of 5 RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 APPLICABILITY:

RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 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.

LCC) 3.0.4.c is applicable.

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

• 60 pCi/gm.

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

(continued)

COMANCHE PEAK - UNITS 1 AND 2 3.4-44 Amendment No. 4-99 to TXX-06038 Page 4 of 5 RCS' Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. DOSE EQUIVALENT

---

Note--------------------

XE-1 33 not within limit.

LCO) 3.0.4.c is applicable.

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

C. Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A or B AND!

not met.

C.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR DOSE EQUIVALENT 1-131

> 60 pCi/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1

---

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

7 days Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT XE-1 33 specific activity < 500 piCi/gm.

(continued)

I COMANCHE PEAK - UNITS 1 AND 2 3.4-45 Amendment No. 64 to TXX-06038 Page 5 of 5 RCS Specific Activity 3.4.16 ACTIONS (continued)

-

I--- --

SURVEILLANCE FREQUENCY SR 3.4.16.2

---

NOTIE------------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 0.45 siCi/gm.

14 days 1 102 AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of > 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period SR 3.4.16.3 DELETED DELETED COMANCHE PEAK - UNITS 1 AND 2 3.4-46 Amendment No. 49 to TXX-06038 Page I of 7 ATTACHMENT 5 to TXX-06038 RETYPED TECHNICAL SPECIFICATION BASES PAGES Pages B 3.4-101 B 3.4-102 B 3.4-103 B 3.4-104 B 3.4-105 B 3.4-106 B 3.4 REACTOR COOLANT SYSTEM (RCS) to TXX-06038 Page 2 of 7 RC'S Specific Activity B 3.4.16 B 3.4.16 RCS Specific Activity BASES BACKGROUND 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident, or at the Low Population Zone outer boundary foir the radiological release duration, is specified in 10 CFR 100.11 (Ref. 1).

Doses to the Control Room operators must be limited per GDC 19. The limits on specific activity ensure that the doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsi:e radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) or a main steam lire break (MSLB) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33. The allowable levels are intended to ensure that offsite and Control Room doses meet the appropriate acceptance criteria in the Standard Review Plan. The limits in the LCO are specific to CPSES due to the implementation of the alternate steam generator tube repair criteria.

1 25 25 APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensures that the resulting offsite and Control Room doses meet the appropriate Standard Review Plan acceptance criteria following a SGTR or a MSLB accident. The SGTR safety analysis (Ref. 2) assumes the specific activity of the reactor coolant at, or more conservative than, the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm.

The MSLB safety analysis (Ref. 3) assumes the specific activity of the reactor coolant at. or more conservative than, the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 27.8 gpm in the affected steam generator and 450 gpm combined in the unaffected steam generators. The safety analysis for both accidents assumes the specific activity of the secondary coolant at ts limit of 0.1 pCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.18, "Secondary Specific Activity."

(continued)

I 25 25 1 The referenced safety analysis reports the doses for the SGTR assuming DEI-1 31 to be 1.0 pCi/gm.

The doses reported using this value have been shown to be conservative relative to those that would be calculated using a DEI-1 31 value of 0.45 pCi/gm.

2 To obtain the maximum benefit from the steam generator alternate repair criteria, the MISLB radiological consequences analysis assumes a leak rate to the faulted steam generator during the accident that results in calculated consequences approaching a small fraction (10%) of the 10 CFR 100 guideline values for the accident initiated spike. This leak rate provides a maximum primary-to-secondary leak rate limit against which the predicted end-of-cycle leakage is compared.

25 COMANCHE PEAK - UNITS 1 AND 2 B 3.4-1 01 Revision 2X to TXX-06038 RCOS Specific Activity Page 3 of 7 B 3.4.16 BASES APPLICABLE The analysis for the MSLB and SGTR accidents establish the acceptance l 25 SAFETY limits for RCS specific activity. Reference to these analyses are used to I

ANALYSES assess changes tD the unit that could affect RCS specific activity, as they (continued) relate to the acceptance limits.

Each of the above analyses must consider two cases of reactor coolant 25 specific activity. One case assumes specific activity at 0.45 pCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases, by a factor of 500 or 335, the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a MSLB or SGTR, respectively. The second case assumes the initial reactor coolant iodine activity at 60.0 pCi/gm DOSE EQUIVALENT 1-131 due to a pre-acciclent iodine spike caused by an RCS transient. In both cases, the noble gas specific activity is assumed to be the equivalent of 1 % fuel defects which corresponds to [715] pCi/gm DOSE EQUIVALENT XE-133.

These analyses also assume a loss of offsite power at the same time as 25 the SGTR or the MSLB event. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature N-16 signal. The MSLB causes a reduction in reactor coolant temperature and pressure.

The temperature decrease causes an increase in reactor power. The power increase will trip the reactor on high neutron flux or overpower N-

16. The pressure decrease will initiate a reactor trip on either low pressurizer pressure or the safety injection signal initiated by low pressurizer pressure, low steam generator pressure, or high containment pressure.

For the SGTR and the MSLB, the coincident loss of offsite power causes 25 the steam dump valves to close to protect the condenser. For the SGTR, a rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG atmospheric relief valves and the main steam safety valves. A failure to close of the atmospheric relief valve on the affected SG is also assumed. The unaffected SGs remove core decay heat by venting steam to the atmosphere until RHR system is placed in service. For the MSLB, an uncontrolled (i.e., released to atmosphere) blowdown of cnly one steam 25 generator is assumed. The unaffected SGs remove core decay heat by venting steam to the atmosphere until RHR system is placed in service.

Radioactively contaminated steam is released to the atmosphere through the faulted SG as well as the intact SGs assuming the primary to secondary leak rates shown above.

The applicable safety analysis shows the radiological consequences of 125 either an SGTR or an MSLB accident are within a small fraction of the Reference 1 dose guideline limits. Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed 60 pCi/gnmi DOSE EQUIVALENT 1-131 for more tWan 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.4-102 Revision 39 to TXX-06038 Page 4 of 7 RCS Specific Activity B 3.4.16 BASES APPLICABLE The limits on RC'; specific activity are also used for establishing SAFETY standardization in radiation shielding and plant personnel radiation ANALYSES protection practices.

(continued)

RCS specific activity satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).

25 39 LCO The iodine specific activity in the reactor coolant is limited to 0.45 pCi/gm 25 DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to [500] pCi/gm DOSE EQUIVALENT XE-1 33, as contained in SR 3.4.16.2 and SR 3.4.16.1 respectively. The limits on specific activity ensure that offsite and Control Room doses will meet the appropriate Standard Review Plan acceptance criteria.

The SGTR accident analysis (Ref. 2) and the MSLB accident analysis (Ref. 3) show that the calculated dose levels are within acceptable limits.

Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a MSLB or SGTR, lead to doses that exceed the SRP-acceptance criteria.

j125 (continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.4-103 Revision 39 to TXX-06038 Page 5 of 7 RCS Specific Activity B 3.4.16 BASES (continued)

APPLICABILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 is necessary to limit the potential consequences of an SGTR and an MSL.B to within the SRP acceptance criteria.

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, monitoring of RCS specific activity is not required.

ACTIONS A. 1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LOC) limit, samples at intervals of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> must be taken to demonstrate that the specific activity is

• 60.0 pCi/gm. The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were no iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on Required Actions A.1 and A.2 while the DOSE EQUIVALENT 1-131 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

(continued) 39 25 34 16 1 25 COMANCHE PEAK - UNITS 1 AND 2 B 3.4-1 04 Revision 38 to TXX-06038 RC'S Specific Activity Page 6 of 7 B 3.4.16 BASES ACTIONS B.1 (continued)

With the DOSE EQUIVALENT XE-133 in excess of the allowed limit, DOSE EQUIVALENT XE-1 33 must be restored to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.

A NOTE permits the use of of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required 25 Action B.1 while the DOSE EQUIVALENT XE-1 33 LCO limit is not met.

This allowance is acceptable due to the significant conservatism incorporated into :he specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

C.1 and C.2 If the Required Ac~tion and the associated Completion Time of Condition A or B is not met or if the DOSE EQUIVALENT 1-131 is > 60).0 pCi/gm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant at least once every 7 days. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in noble gas specific activity.

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

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

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

(continued)

COMANCHE PEAK - UNITS 1 AND 2 B 3.4-105 Revision 25 to TXX-06038 Page 7 of 7 RC"; Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days. The Frequency, I between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power change ' 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

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

SR 3.4.16.3 DELETED I

REFERENCES

1.

10 CFR 1C)0.11, 1973.

2.

FSAR, Section 15.6.3.

3.

FSAR, Section 15.1.5 1 25 COMANCHE PEAK - UNITS 1 AND 2 B 3.4-106 Revision 2a