2CAN030803, License Amendment Request, Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.

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License Amendment Request, Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.
ML080850908
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 03/13/2008
From: Mitchell T
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
2CAN030803
Download: ML080850908 (23)
v • d • e


Text

Entergy Operations, Inc.

1448 SA, 333 RusseIlville, AR 72802 Tel 479-858-3110 Timothy G. Mitchell Vice President, Operations Arkansas Nuclear One 2CAN030803 March 13, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment to the Technical Specifications (TS) for Arkansas Nuclear One, Unit 2 (ANO-2).

Theproposed changes would replace the current pressurized water reactor (PWR) TS 3.4.8 limit on reactor coolant system (RCS) gross specific activity with a new limit on RCS noble gas specific activity. The noble gas specific activity limit would be based on a new dose equivalent Xe-1 33 (DEX) definition that would replace the current E Bar average disintegration energy definition. In addition, the current dose equivalent 1-131 (DEI) definition would be revised to allow the use of additional thyroid dose conversion factors (DCFs).

The changes are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec." The availability of this TS improvement was announced in the FederalRegister on March 15, 2007 (72FR12217) as part of the consolidated line item improvement process (CLIIP). provides a description and assessment of the proposed changes, as well as confirmation of applicability. Attachment 2 provides the existing TS pages marked-up to show the proposed changes. Attachment 3 provides final (clean) TS pages. Attachment 4 provides a markup of the associated TS Bases, for information only.

The proposed changes do not include any new commitments.

Entergy requests approval of the proposed amendment by March 1, 2008. Once approved, the amendment shall be implemented within 90 days. Although this request is neither exigent nor emergency, your prompt review is requested. In accordance with 10 CFR 50.91, a copy of this application, with enclosures, is being provided to the designated Arkansas State Official.

2CAN030803 Page 2 of 2 If you have any questions or require additional information, please contact David Bice at 479-858-5338.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March 13, 2008.

Sincerely, TGM/dbb Attachments:

1. Description and Assessment of Proposed Changes
2. Proposed Technical Specification Changes
3. Final Technical Specification Pages
4. Technical Specification Bases Changes Markups (For Information Only) cc: Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. 0. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS 0-7 D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Division of Radiation Control and Emergency Management Arkansas Department of Health & Human Services P.O. Box 1437 Slot H-30 Little Rock, AR 72203-1437

Attachment 1 2CAN030803 Description and Assessment of Proposed Changes

Attachment to 2CAN030803 Page 1 of 4

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2).

The proposed changes would replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity. The noble gas activity would be based on DOSE EQUIVALENT XE-1 33 and would take into account only the noble gas activity in the primary coolant. The changes were approved by the NRC staff Safety Evaluation (SE) dated September 27, 2006 (ADAMS ML062700612) (Reference 1). 2 Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" was announced for availability in the Federal Register on March 15, 2007 as part of the consolidated line item improvement process (CLIIP).

2.0 PROPOSED CHANGE

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

  • Revise the definition of DOSE EQUIVALENT 1-131.
  • Delete the definition of I-AVERAGE DISINTEGRATION ENERGY.

.Add a new TS definition for DOSE EQUIVALENT XE-1 33.

  • Revise LCO 3.4.8, "RCS Specific Activity" to delete references to gross specific activity, add limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33, and delete Figure 3.4-1.
  • Revise LCO 3.4.8 "Applicability" to specify the LCO is applicable in MODES 1, 2, 3, and 4.

Modify ACTIONS as follows:

A. Reference to applicable MODES is deleted.

B. ACTION 'a' is modified to replace "specific activity" with "DOSE EQUIVALENT 1-131" and define an upper limit for DOSE EQUIVALENT 1-131 that is applicable at all power levels. A Note allowing the applicability of LCO 3.0.4.c is also added.

C. ACTION 'b' is modified to provide a Required Action for DOSE EQUIVALENT XE-1 33 with a Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. A Note allowing the applicability of LCO 3.0.4.c is also added.

D. ACTION 'c' is modified to provide a Required Action when DEI exceeds 60 pCi/gm and to reflect the change in the LCO Applicability.

Table 4.4-4 is deleted and replaced by revising SR 4.4.8 (renumbered as 4.4.8.1) to verify the limit for DOSE EQUIVALENT XE-1 33 and new SR 4.4.8.2 to verify the limit

'for DOSE EQUIVALENT i-131. A Note is added to allow entry into MODES 2, 3, and 4 prior to performance of the SR.

Attachment to 2CAN030803 Page 2 of 4 Wording and format changes are included in the ANO-2 TS markup and clean pages which differentiate from that illustrated in TSTF-490 becauselANO-2 is not an improved TS (ITS) plant. The proposed ANO-2 TSs capture the requirements of TSTF-490 while maintaining the wording and formatting associated with custom TSs. provides a markup of the affected TSs delineated above. Attachment 3 provides the final (clean) TS pages associated with these changes. includes a markup of the impacted TS Bases pages for information only. The TS Bases will be revised in accordance with the TS Bases Control Program (TS 6.5.14) during implementation of TSTF-490 once approved by the NRC.

Note that TSTF-490 was written and approved considering the most current version of the Improved Technical Specifications (ITS) with approved TSTFs at the time. As described above, this included reference to Limiting Condition for Operation (LCO) 3.0.4.c, which was incorporated in the ITS under TSTF-359, Mode Change Limitations. Entergy has submitted a proposal to adopt TSTF-359, which is currently under review by the NRC (Reference 4).

Therefore, the above changes associated with LCO 3.0.4 are dependent on the NRC approval of Entergy's TSTF-359 adoption proposal.

Therefore, the markup and final versions of TS 3.4.8 included in Attachments 2 and 3, and the TS 3.4.8 Bases markup included in Attachment 4 are shown as ifthe ANO-2 TSTF-359 submittal has been approved (minus the addition of a new amendment number in the associated page footer). Entergy will submit new markup and final TS pages upon NRC request should the TSTF-359 application be revised or should the approval of either of these applications be significantly delayed. If the aforementioned applications are approved as anticipated, Entergy will submit new final TS pages that include the new amendment numbers upon request from the NRC.

3.0 BACKGROUND

The background for this application is as stated in the model safety evaluation (SE) in the NRC's Notice of Availability published on March 15, 2007 (72FR12217), the NRC Notice for Comment published on November 20, 2006 (71 FR67170), and TSTF-490, Revision 0.

4.0 TECHNICAL ANALYSIS

In the model SE, the NRC included statements which would require the licensee to identify specific information in support of adopting TSTF-490. The following provides ANO-2 specific information in this regard.

1. Section 3.1.1 of the model SE includes a list of acceptable dose conversion factors (DCF) for use in the determination of dose equivalent iodine (DEI) in relation to dose consequence analyses. The ANO-2 analyses employ Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites."

Attachment to 2CAN030803 Page 3 of 4

2. In the first paragraph of Section 3.1.2 of the model SE, a bracketed list of isotopes is provided that designates the noble gases that may be used in the determination of dose equivalent xenon (DEX). All isotopes depicted within these brackets are currently considered in the ANO-2 calculation of DEX. This is captured in the proposed ANO-2 definition of DEX to be included in TS Section 1.0.
3. Section 3.1.2 of the model SE also provides two possible determination methods for DEX.

ANO-2 uses the average disintegration energies derived from the dose conversion factors provided in ICRP Publication 2, "Report of ICRP Committee II on Permissible Dose for Internal Radiation" for non-fuel damage events. This is also depicted in the proposed definition of DEX in TS Section 1.0.

4. Section 3.1.3 states that it is incumbent on the licensee to ensure that the site specific limits for both DEI and DEX are consistent with the current steam generator tube rupture (SGTR) and main steam line break (MSLB) radiological consequence analyses. The ANO-2 analyses value for DEI is currently and will remain 1.0 pCi/gm. The ANO-2 analyses value for DEX is 1200 pCi/gm. This is depicted in the attached TS markups.

Entergy Operations, Inc. (Entergy) has reviewed References 1, 2 and 3, and the model SE published on November 20, 2006 (71 FR67170) as part of the CLIIP Notice for Comment.

Entergy has applied the methodology in Reference 1 to develop the proposed TS changes.

Entergy has also concluded that the justifications presented in TSTF-490, Revision 0 and the model SE prepared by the NRC staff are applicable to ANO-2 and justify this amendment for the incorporation of the changes to the ANO-2 TS.

5.0 REGULATORY ANALYSIS

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on March 15, 2007 (72FR12217), the NRC Notice for Comment published on November 20, 2006 (71FR67170), and TSTF-490, Revision 0.

6.0 NO SIGNIFICANT HAZARDS CONSIDERATION Entergy Operations, Inc. (Entergy) has reviewed the proposed no significant hazards consideration determination published in the FederalRegister on March 15, 2007 (72FR12217) as part of the CLIIP. Entergy has concluded that the proposed determination presented in the notice is applicable to Arkansas Nuclear One, Unit 1 (ANO-1) and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(a).

Attachment to 2CAN030803 Page 4 of 4 7.0 ENVIRONMENTAL EVALUATION Entergy Operations, Inc. (Entergy) has reviewed the environmental consideration included in the model SE published in the Federal Register on March 15i 2007 (72FR1 2217) as part of the CLIIP. Entergy has concluded that the staffs findings presented therein are applicable to Arkansas Nuclear One, Unit 1 (ANO-1) and the determination is hereby incorporated by reference for this application.

8.0 REFERENCES

1. NRC Safety Evaluation (SE)'approving TSTF-490, Revision 0 dated September 27, 2006
2. Federal Notice for Comment published on November 20, 2006 (71 FR 67170)
3. Federal Notice of Availability published on March 15, 2007 (72FR12217)
4. Entergy letter to NRC dated October 22, 2007, Technical Specification Changes Regarding Mode Change Limitations and Associated Bases Using the Consolidated Line Item Improvement Process (TSTF-359), Arkansas Nuclear One, Unit 2 (2CAN100701)

(TAC No. MD7174)

Attachment 2 2CAN030803 Proposed Technical Specification Changes

DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

AZIMUTHAL POWER TILT - T2 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT 1-131 1.18 DOSE EQUIVALENT 1131 shall be that coneRnFration of 1131 (ýCiirarm) which aloRn would produce the same thyroid dose as the quantity and iso~topic miqXturo of 1-Z13, 1132, 1133, 1134 and 1135 actually preSent. The thyroid dose conVersion factors; us*d foF th;i* .6_÷la*i o alcu shall be those listed iR Table 111of TID 181,4 "*alculation at Distance Farctrs for Poe'-r and Test Reactor Sites."DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and I-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites."

DOSE EQUIVALENT XE-133E AVERAGE DISINTEGRATION ENERGY 1.19 Eshall be the aVerage (weighted in prE)opoRio to the cocnrto f e~ach radionuclide in the rcactor coolIant at the time of sampling) of the sum of the average beta and gamma an*egies per disintegration (in ,MV)for isotopes, ether than iodinos, with half lives,greater than 15 minues main up at least 95% of the total non io-dine acivtyi the eeeeait7.DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using the average disintegration energies derived from the dose conversion factors provided in ICRP Publication 2, "Report of ICRP Committee II on Permissible Dose for Internal Radiation" for non-fuel damage events.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

ARKANSAS - UNIT 2 4-1 Amendment No. -67,2&5,266,

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and
b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

ARKANSAS - UNIT 2 4-1 Amendment No. 4-57-,265,2-66,

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 TheRCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity of the prima','y coolant shall be within limitsed-teG.

a. 1.0 pni/gram DOSE EQUIVALENT 1131, and
b.
  • 00 E pn/i*

t*/g*ram.

APPLICABILITY: MODES 1, 2, 3, and 4,-apd-5.

ACTION:

MODES 1, 2 and 3*:Note: The provisions of Specification 3.0.4.c are applicable to ACTION a and b.

a. With the spccific actiVity Of the primar' coolant 1.0 pCi/gram DOSE EQUIVALENT 1-131 not within limit:
1. Verify DOSE EQUIVALENT 1-131 < 60 pCi/gm once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and
2. Restore DOSE EQUIVALENT 1-131 within limit withinfer ,mne-thaR 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s-du*i~-e continuous time intoal Or exceeding tho limit In;e ,how .. FOguro 3.4 1, be in at least HOT STANDBY with Tae 500 0 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The proviSions of Specification 3.0.4.cA aFe apphiGable.
b. With the DOSE EQUIVALENT XE-133 not within limit, restore specific actiVity Of the prima4y coolant > 4094 pi/gram, be in at least HOT STANDRY T.r 500F within 6 ho-rs.DOSE EQUIVALENT XE-1 33 within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

ARKANSAS - UNIT 2 3/4 4-18 Amendment No. 92, Next Page is 3/4 4-22

REACTOR COOLANT SYSTEM To be moved to Ar'-r*A..

,,. Cotued previous page MODSF 2r'1, 2 3, 41,'aR3 dn5,.

c. With the requirementsof ACTION a and/or b not met, or with DOSE EQUIVALENT 1-131

> 60 pCi/gm, be in at least HOT STANDBY in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />speoific a.tiVity Of tho prim=a,'y coIant >1.0 p-igram DOSE EQUIVALENT 1.131 o 1001E p'i.gram, prform the sampling and analysis,rqu61Frots' of item 4 a) of Tabhlo 4.4- until the Gpe9iG& actiVity of the primnar' coolant is roctorod to witi SURVEILLANCE REQUIREMENTS 4.4.8.1 Verify reactor coolant DOSE EQUIVALENT XE-133T-he specific activity-4f4he-ima4y coolant shall be detormninod to be within the limits by pcrformac oftesapig and analysis progrFa of Tab'l 4,4-4

, 1200 pCi/gm once every 7 days.*

4.4.8.2 Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 1.0 pCi/gm:*

a. once every 14 days, and
b. between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after THERMAL POWER change of > 15% RATED THERMAL POWER within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.
  • Only required to be performed in MODE 1.

PuK-PA10~

/A0 .4A 4 A-4W i! AmendMen X

TABLE 4,.4-4 PRIMARY COO',_NT SPECIFIC ACTIVI.T SAMPLE AND ANALYSIS PROGRA\M TYPEANDANIALYSI OF MEASUREMENT SAMPLE AlNALYSIS .

FRQ*U-IFNCI MODES ANDr IN WHICH ANlAILYSIS SAMPLE REQUIRED 4, Gross

- ActiVity Detrmrination At least once per 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />"s 2L. Isotopic AnalyssfrDS 1 peF 14 days 4-EQIUIVA1:::ýl ENT 131 Co-Qnc-ntfraton 3, Radiochom~ical for F=Dctc~rinatio 1p-6--moths 4 4- Isotopic Analysis for lodino a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,  !#,2#f,3#f,41#f,5#

Inc'luding 1131,1 133, and 1 135 whenever the DOSE EQUIVALENT 1-131 exceeds 1.0 ~t~i~gram, and b) One sample bc~et 12, 2 and 6 hourFs fo"lWing a THERMAL oxcooding 15 percont of the RATED THERMAL POWER Within a end hour J* I I._t:lLL _ ....  : :_t _ _*:. _:L._ t ZL _ ._ _:.__

IT 6nRIl Mne spedcino actvt oRY inc prMer Iyte c n I restorca WInInI its imits.

J'n

-- A. . .CM(% l M ý A I k I.-4.

SampleJI te boa tLkRUu aite1 at 1111 ...

Ipp A- 22 Ir- LJ JII

-..- fuyb. Ul F;Cvve rD~~ IUU CIPbfU biiotlU It!Ut~U! Wctb-f k' ;kr+;rn 1 nr Ai kn r.nrIn-,

n twl  %ýr t=t . .ý . . - - . - - I tv " M 0 r.

ADIVAKIQAQ I INIIT'1) "IIA A4 On"

250 A

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UNACCEPTABLE H OPERATION H

~'150 -H-H I [ [ I E I E I i I-ý-

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03 I

S100 C,,-

I I II H

H I I H

>50 Ol ACCEPTABLE I I OPERATION --

0I r01 0

0 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.41

  • A*lmd p*I lilia I pLI'IP ! *1*4 IP*__!___ _'____,L u.i ....... E,%....... JL _J*

WuuI! 4uwl 11.44 1 Imp r-ii :1primar: 6ee.ooipnt~v-l DOIiC :vitv Limit Versus= Ar-erent of RATIED THIIERMAL POWER with the rirnr*: ,oolantspe*twii ActiUity > 1.0-tiuvram Dr.. Equivalent 1-131 A IIAKICI'

~ AC I kllT "ir.iui OIA A 04 APL 94-..- 4

Attachment 3 2CAN030803 Final Technical Specification Pages

,,,. t,-

DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

AZIMUTHAL POWER TILT - T, 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT 1-131 1.18 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites."

DOSE EQUIVALENT XE-1 33 1.19 DOSE EQUIVALENT XE-1 33 shall be that concentration of Xe-1 33 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-1 31 m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using the average disintegration energies derived from the dose conversion factors provided in ICRP Publication 2, "Report of ICRP Committee II on Permissible Dose for Internal Radiation" for non-fuel damage events.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and
b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

ARKANSAS - UNIT 2 1-4 Amendment No. 1-57,255,266,

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 specific activity shall be within limits.

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

ACTION:

Note: The provisions of Specification 3.0.4.c are applicable to ACTION a and b.

a. With the DOSE EQUIVALENT 1-131 not within limit:
1. Verify DOSE EQUIVALENT 1-131 -560 pCi/gm once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and
2. Restore DOSE EQUIVALENT 1-131 within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
b. With the DOSE EQUIVALENT XE-133 not within limit, restore DOSE EQUIVALENT XE-1 33 within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
c. With the requirements of ACTION a and/or b not met, or with DOSE EQUIVALENT 1-131

> 60 pCi/gm, be in at least HOT STANDBY in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.8.1 Verify reactor coolant DOSE EQUIVALENT XE-1 33 specific activity -51200 pCi/gm once every 7 days.*

4.4.8.2 Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity < 1.0 pCi/gm:*

a. once every 14 days, and
b. between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after THERMAL POWER change of > 15% RATED THERMAL POWER within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.
  • Only required to be performed in MODE 1.

ARKANSAS - UNIT 2 3/4 4-18 Amendment No. P2, Next Page is 3/4 4-22

Attachment 4 2CAN030803 Technical Specification Bases Changes Markups (For Information Only)

i REACTOR COOLANT SYSTEM BASES 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosioh of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY BACKGROUND The limitations On the Gpecific actiVity of the prim~arY coolant cnsuro that the resulting 2 hou doses at the site boUndary will net exceed an appropriately small fraction of Part 100 limits follo)Wing a steam goncrator tube ru1pture accident in conjunction with an assumed Steady state prmW to econdary steamR gGeneator leakage rate of 1.0 GPMI and a concrrenF8t l)ss Of offsite electical power. The values for the limits onA s-pecific activity represent interim limits based upon a parame~tric evaluation by the NRC of typical Site locations. These values, are G 1ie 1. Itiv IH 6pUUIG iia1 6119 p1F

  • l L 1 ill*la Of i*ii* sUal
  • I 61I19 , UG*aI 1 Si teI boun~dary location and mneteorological conditions, were nOt considered in thisP0:1 evalution. The NRCis finaliing site specificG citeria Which will be used as tlhe basis for t reevaluation of the specific activity lim~its; of this site. This reevaluation may reGult in higher limnitS.

The AGTION statement permitting POWER OPERATIONr*t continue for limited time with the ri)maryco*,--nt'. specific activity " 1.0 Pigram DOSE EQUIVALENT I 131, but within the allowable l.im.it shown Figure 3.4 1, accomAodates p6'o ible iodine spiking phenomenon which May occur fo9llowing changes in THERMAL POWER.

Reducig T-avg to 500 0F prevents the release Of activity should a steam; gencratrF tube ruptur Ie since the saturation prossur~e of the prim~ary coolant is below the lift pressure of the atmospheric steam relief valves. The suR'eillanc rqiemets provide adequate assurance that exc*eivne specific atvity levels in the prim;aP cooIant will be det*6tdd in sfficient timeRto take co,.rective acto. Information obtaired oRn iodine spiking will be used to assess the parameti~g apssocar*-ted with spiking phenom~ena. A reductionR in freqUency of istpcanalyses following power changes, may be permiss6ible if justified by the data obtained.

The maximum dose 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 for the radiological release duration, is specified in 10 CFR 100.11 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

ARKANSAS - UNIT 2 B 3/4 4-5 Amendment No. 92,12*1,21 Rev. 44-9,

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the dose consequences in the event of a steam line break (SLB) or steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in 10 CFR 100 and GDC 19.

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure that the resulting offsite and control room doses meet the appropriate acceptance criteria following a SLB or SGTR accident.

The safety analyses assume the specific activityof the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 1 gpm exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 pCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.1.4, "Secondary Specific Activity."

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

ARKANSAS - UNIT 2 B 3/4 4-5 Amendmont No. 92,124,242 Rev. 4,49,

REACTOR COOLANT SYSTEM BASES 3/4.4.8 SPECIFIC ACTIVITY (continued)

APPLICABLE SAFETY ANALYSES (continued)

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.0 pCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB or SGTR (by a factor of 500). The second case assumes the initial reactor coolant iodine activity at 60.0 pCi/gm DOSE EQUIVALENT 1-131 due to an iodine spike caused by a reactor or an RCS transient prior to the accident. In both cases, the noble gas specific activity is assumed to be 1200 pCi/gm DOSE EQUIVALENT XE-1 33.

The SGTR analysis assumes that offsite power is lost at the same time the tube rupture occurs.

Thus, radioactively contaminated steam discharges to the atmosphere through the atmospheric dump valves or the main steam safety valves. The atmospheric discharge from the ruptured SG stops when the operator isolates the SG at 30 minutes into the event. The unaffected SG then removes core decay heat by venting steam until the cooldown ends and the Decay Heat Removal (DHR) system is placed in service.

The SLB radiological analysis assumes thatoffsite power is lost at the same time as the pipe break occurs outside containment. The affected SG blows down completely and steam is vented directly to the atmosphere. The unaffected SG removes core decay heat by venting steam to the atmosphere until the cooldown ends and the DHR system is placed in service.

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

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

RCS Specific Activity satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3).

LCO The iodine specific activity in the reactor coolant is limited to 1.0 pCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 1200 pCi/gm DOSE EQUIVALENT XE-133. The limits on specific activity ensure that offsite and control room doses will meet the appropriate acceptance criteria (Ref. 1).

The SLB and SGTR accident analyses (Ref. 2) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a SLB or SGTR, lead to doses that exceed the acceptance criteria (Ref. 1).

ARKANSAS - UNIT 2 B 3/4 4-6 A.mondmont No. 124,242 Rev. 48-,

V..

REACTOR COOLANT SYSTEM BASES 3/4.4.8 SPECIFIC ACTIVITY (continued)

APPLICABILITY In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-1 33 is necessary to limit.the potential consequences of a SLB or SGTR to within the acceptance criteria (Ref. 1).

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

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

ACTIONS A Note permits the use of the provisions of LCO 3.0.4.c for ACTION a and b. This allowance permits entry into the applicable MODE(S), relying on Required Actions 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.

ACTION a 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 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 continued every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limit 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, ifthere were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

ACTION b With the DOSE EQUIVALENT XE-1 33 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit 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, ifthere were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

ACTION c If the Required Action and associated Completion Time of ACTION a and/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 the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-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.

ARKANSAS - UNIT 2 B 3/4 4-7 Amendment No. 49,424,243,2-42 Next Pago is B 3/4 414 Rev. 7-,4,

REACTOR COOLANT SYSTEM BASES 3/4.4.8 SPECIFIC ACTIVITY (continued)

SURVEILLANCE REQUIREMENTS A Note modifies the SRs 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 4.4.8.1 SR 4.4.8.1 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 the 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. The 7 day Frequency considers the low probability of a gross fuel failure during this time; Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 4.4.8.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

SR 4.4.8.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering noble gas activity is monitored every 7 days.

The Frequency, between 2 and 6 hours6.944444e-5 days <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.

REFERENCES

1. 10 CFR 100.11.
2. CALC-98-E-0036-04.
3. 10 CFR 50.36.

ARKANSAS - UNIT 2 B 3/4 4-8 Rev.

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