Text

Application for Technical Specification Improvement to Adopt TSTF-490, Revision 0, Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec.
	ML080920781
Person / Time
Site:	Millstone
Issue date:	03/25/2008
From:	Hartz L; Dominion Nuclear Connecticut
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	08-0135
	Download: ML080920781 (44)
	v • d • e

Dominion Nuclear Connecticut, Inc.

Millstone Power StationDo ion Rope Ferry Road Waterford, CT 06385 March 25, 2008 U.S Nuclear Regulatory Commission Serial No 08-0135 Attention: Document Control Desk NSSL/MLC RD One White Flint North Docket Nos. 50-336 11555 Rockville Pike 50-423 Rockville, MD 20852-2738 License Nos. DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT. INC.

MILLSTONE POWER STATION UNITS,2 AND 3 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT TO ADOPT TSTF-490, REVISION 0, "DELETION OF E BAR DEFINITION AND REVISION TO RCS SPECIFIC ACTIVITY TECH SPEC" In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) is submitting a request for an amendment to the technical specifications (TS) for Millstone Power Station Units 2 and 3 (MPS2 and MPS3). The proposed changes would replace the current 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-133 (DEX) definition that would replace the current E Bar average disintegration energy definition.

The changes are consistent with NRC-approved Industry Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech. Spec." The availability of this TS improvement was announced in the Federal Register on March 19, 2007 (72 FR 12838) as part of the consolidated line item improvement process (CLIIP). Because neither MPS2 nor MPS3 has adopted the Standard Technical Specifications (STS), DNC is proposing minor variations and/or deviations from the TS changes described in TSTF-490, Revision 0, to provide consistent terminology and format with the MPS2 and MPS3 TS. The minor variations and/or deviations from the specific wording/format provided in TSTF-490, Revision 0, do not change the meaning, intent, or applicability of the CLIIP.

Attachments 1 and 4 provide a description and assessment of the proposed changes, as well as confirmation of applicability for MPS2 and MPS3, respectively. Attachments 2 and 5 provide the MPS2 and MPS3 marked-up TS pages, respectively. Likewise, Attachments 3 and 6 provide the marked-up TS Bases pages for MPS2 and MPS3, respectively. The marked-up TS bases

Serial No: 08-0135 Docket Nos. 50-336 and 50-423 Application to Adopt TSTF-490, Rev. 0 Page 2 of 3 pages are provided for information only. The changes to the affected TS bases pages will be incorporated in accordance with the TS Bases Control Program when this amendment request is approved.

The proposed amendment does not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92. The Facility Safety Review Committee has reviewed and concurred with the determinations herein.

Issuance of this amendment is requested no later than September 30, 2008, with the amendment to be implemented within 120 days of issuance to permit adequate time to install and test the required counting equipment.

In accordance with 10 CFR 50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.

Should you have any questions in regard to this submittal, please contact Mr. E. Thomas Shaub at (804) 273-2763.

Sincerely, Leslie N. Hartz Vice President - Nuclear Support Services COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Leslie N. Hartz, who is Vice President - Nuclear Support Services of Dominion Nuclear Connecticut, Inc. She has affirmed before me that she is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of her knowledge and belief.

Acknowledged before me this A!Sday of TR.AA ,2008.

My Commission Expires: &/, o;t 3 ,)-008

Ž I MARGARET B.BENNETT CommNotary Pblich35l30?na 0Notary PlPublic My Comminion Expires Commonwealth ofAug, 31, 2008 Virginia

Serial No: 08-0135 Docket Nos. 50-336 and 50-423 Application to Adopt TSTF-490, Rev. 0 Page 3 of 3 Attachments:

1. Evaluation of Proposed License Amendment, MPS2

2. Marked Up Technical Specification Pages, MPS2

3. Marked Up Tech Spec Bases Pages, MPS2

4. Evaluation of Proposed License Amendment, MPS3

5. Marked Up Technical Specification Pages, MPS3

6. Marked Up Tech Spec Bases Pages, MPS3 Commitments made in this letter: None cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 J. D. Hughey Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No. 08-0135 Docket No. 50-336 ATTACHMENT I APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR EVALUATION OF PROPOSED LICENSE AMENDMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

Serial No. 08-0135 Docket No. 50-336 MPS2 Evaluation for Deletion of EBAR Attachment 1 Page 1 of 3 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR EVALUATION OF PROPOSED LICENSE AMENDMENT

1.0 INTRODUCTION

In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc.

(DNC) is submitting a request for an amendment to the technical specifications (TS) for Millstone Power Station Unit 2 (MPS2). The proposed changes would replace the current limits on primary coolant gross specific activity with limits on primary coolant noble gas activity. The noble gas activity would be based on DOSE EQUIVALENT XE-133 and would take into account only the noble gas activity in the primary coolant. The changes were approved by the NRC's Safety Evaluation (SE) dated September 27, 2006 (ADAMS ML062700612) (Reference 1). Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" was announced for availability in the Federal Register on March 19, 2007 as part of the consolidated line item improvement process (CLIlP).

2.0 PROPOSED CHANGE

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

Revise the definition of DOSE EQUIVALENT 1-131 in MPS2 TS 1.19.

Replace MPS2 TS 1.20, "E Bar, AVERAGE DISINTEGRATION ENERGY" with new MPS2 TS 1.20, "DOSE EQUIVALENT XE-1 33."

Revise LCO 3.4.8 to replace the gross specific activity limit with a DOSE EQUIVALENT XE-1 33 limit.

Revise LCO 3.4.8 "Applicability" to specify the LCO is applicable in MODES 1, 2, 3, and 4.

" Delete Figure 3.4-1, "DOSE EQUIVALENT 1-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0 gCi/gram DOSE EQUIVALENT 1-131," and Table 4.4-2, "PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM."

Modify the LCO 3.4.8 ACTIONS as follows:

A. Renumber ACTION (d) as ACTION (a), consistent with format of TSTF-490, and delete the references to a gross specific activity limit and Table 4.4-2.

B. Modify ACTION (b), formerly ACTION (a), to delete the reference to Figure 3.4-1, and define an upper limit for DOSE EQUIVALENT 1-131 that is applicable at all power levels.

C. Modify ACTION (c), formerly ACTION (b), to delete the reference to Figure 3.4-1, and revise the required actions (Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ) to agree with TSTF-490.

Serial No. 08-0135 Docket No. 50-336 MPS2 Evaluation for Deletion of EBAR Attachment 1 Page 2 of 3 D. Add a new ACTION (d) to provide a condition and required action (48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months restoration time) for DOSE EQUIVALENT XE-1 33.

E. Modify former ACTION (c), now ACTION (e), to replace reference to gross specific activity limit with DOSE EQUIVALENT XE-133 limit, and provide contingency actions (Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ) in the event the 48-hour restoration time of new ACTION (d) is not met, consistent with TSTF-490.

Revise Surveillance Requirement (SR) 4.4.8 to replace Table 4.4-2 with two new requirements. SR 4.4.8.1 will verify DOSE EQUIVALENT XE-1 33 is within limits once per 7 days, and SR 4.4.8.2 will verify DOSE EQUIVALENT 1-131 is within limits once per 14 days, and between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a thermal power change of >15% within a 1-hour period. A Note is added to both SRs, consistent with TSTF-490, to allow entry into MODES 2, 3, and 4 prior to performance of the respective SRs.

3.0 BACKGROUND

The background for this application is as stated in the model SE in NRC's Notice of Availability published on March 19, 2007 (72 FR 12838), the NRC Notice for Comment published on November 20, 2006 (71 FR 67170), and TSTF-490, Revision 0.

4.0 TECHNICAL ANALYSIS

DNC has reviewed References 1, 2 and 3, and the model SE published on November 20, 2006 (71 FR 67170) as part of the CLIIP Notice for Comment. DNC has applied the methodology in Reference 1 to develop the proposed TS changes. DNC has also concluded that the justifications presented in TSTF-490, Revision 0 and the model SE prepared by the NRC staff are applicable to MPS2 and support incorporation of this amendment into the MPS2 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 19, 2007 (72 FR 12838), the NRC Notice for Comment published on November 20, 2006 (71 FR 67170), and TSTF-490, Revision 0.

6.0 NO SIGNIFICANT HAZARDS CONSIDERATION DNC has reviewed the proposed No Significant Hazards Consideration determination published in the Federal Register on March 19, 2007 (72 FR 12838) as part of the CLIIP.

DNC has concluded that the proposed determination presented in the notice is applicable to MPS2 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).

Serial No. 08-0135 Docket No. 50-336 MPS2- Evaluation for Deletion of EBAR Attachment 1 Page 3 of 3 7.0 ENVIRONMENTAL EVALUATION DNC has reviewed the environmental consideration included in the model SE published in the FederalRegister on March 19, 2007 (72 FR 12838) as part of the CLIIP. DNC has concluded that the staff's findings presented therein are applicable to MPS2 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 19, 2007 (72 FR 12838)

Serial No. 08-0135 Docket No. 50-336 ATTACHMENT 2 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR MARKED-UP TECHNICAL SPECIFICATION PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

DEFINITIONS AZIMUTHAL POWER TILT - Tq 1.18 AZIMUTHAL POWER TILT shall be the difference between the maximum power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core.

AZIMUTHALPOWERTILT [Maximum power in any core quadrant (upper or lower)]_

L Average power of all quadrants (upper or lower) _

DOSE E QUIVALENT 1-131 %Lýiji*

1.19 DOSE EQUIVALENT 1-131 shall be that concentr ion of -131 micro-curie/gram "

.ie alone would produce the same G4

doselý n'i, 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The hr_ , ".or.-- i f^ttz:.Lzz. 4 for t"^i* MAA1*tiw- s*1÷ .1be t..... hsi-4 t --n-lt,-;

.... Federal Guidance Report No. 11

Q+4 "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

E-AVE EGRATION ENERGY 1.20 E shall be the average sum of t erine for isotopes, other th , with half lives greater than 15 the "Tne activity in the coolant.

STAGGERED TEST BASIS 1.21 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by.t dividing the specified test interval into n enual subinterval and WSW LZO ..........

b. The testing of one system, subsystem, train or other designated component at the beginning o f each subinterval.

FREQUENCY NOTATIO1T 1.22 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall corresp ond to the intervals defined in Table 1.2.

MILLSTONE - UNIT 2 en ment No. 4-04,--6, 29&

Attachment 2 MPS2 Marked-Up Technical Specification Pages INSERT 1.20 DOSE EQUIVALENT XE-133 1.20 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (micro-curielgram) 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-1 33 shall be performed using effective dose conversion factors for air submersion listed in Table [i1.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

REACTOR COOLANT SYSTEM 46PON-40, t999--

SPECIFIC ACTIVITY LIMITING CONDITION FOR OPER~ATION 3.4.8 The specific activity of the primary coolant shall be limited to:

/gram DOSE EQUIVALENT 1-131, and

. 6WOmICA,

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b.IT APPLICABILITY: MODES 1, 2, 3, ACTION: 3NELC-TC z

MODES 1, 2, and 3*:

a. With the specific activity of the primary coolant > 1.0 pU ram DOSE EQUIVALENT 1-131 but within the allowable limit (be and to e left of the line) shown on Figure 3.4-1, operatio y continue fo to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . .Specification 3.0.4 is not icable.

b. With the cific activity of the primary olant > 1.0 pCi/gram DOSE EQUIVAL 1-131 for more than 48 rs during one continuous time interval or eeding the limi me shown on Figure 3.4-1, be in HOT STANDBY with < 515°F thin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c. With the specific acivi he primary coolant > 100/E pCi/gram of gross specific a vity, be HOT STANDBY with Tavg < 515°F within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

MODES 1; 2, 3, 4 an:

d. Wi the specific activity of the primary coolan* 1.0 pCi/gram SE EQUIVALENT 1-131 or > 100/E pCi/gram of gr specific activity, perform the 'ampl ing and analysis require ._ts of item 4 a) of Table 4.4-2 until the specific activity of the p *ary coolant is restored to within its limits.

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MILLSTONE - UNIT 2 3/4 4-13 Amendment No. ZY*,

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Attachment 2 MPS2 Marked-Up Technical Specification Pages INSERT 3.4.8 ACTION

a. With the specific activity of the primary coolant > 1.0 jLCi/gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 *< 60 gCi/gram once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With the specific activity of the primary coolant > 1.0 p.Ci/gram DOSE EQUIVALENT 1-131 but __60 pCi/gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 pCi/gram limit.

Specification 3.0.4 is not applicable.

c. With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or > 60 pCi/gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

d. With the specific activity of the primary coolant > 1100 [tCi/gram DOSE EQUIVALENT XE-1 33, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-1 33 to within the 1100 pCi/gram limit.

Specification 3.0.4 is not applicable.

e. With the specific activity of the primary coolant > 1100 RCi/gram DOSE EQUIVALENT XE-133 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months durng one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

REACTOR COOLA.!,"T" SYSTEM SURVEILLANCE REOUI REMENTS vt Aw~4 a 'TAEWWL.- ?OvkfRl C.i~

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4. sotopic An ysis for oi 0 a) 0 e per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,wheniever the Iticludin -131, 1-13-,and 1-135. ecifiC ati exceeds 1.0 1 grm D ,EQUIVAL 1-1.32.1, b) ne sampi etween 2 an~d 6h~ f1J-ollowil ,/

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Serial No. 08-0135 Docket No. 50-336 ATTACHMENT 3 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR MARKED-UP BASES PAGES (INFORMATION ONLY)

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2

-DDCR08* ?#ir: 005

.44ft*is,200-REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.2 REACTOR COOLANT SYSTEM OPERATIONAL OPERATIONAL LEAKAGE REFERENCES 1 10 CFR 50, Appendix A, GDC 30.

2 Regulatory Guide 1.45, May 1973.

3 FSAR, Section 14 4 NEI 97-06, "Steam Generator Program Guidelines."

5 EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

3/4.4.7 DELETE 3/4.4.8 SPECIFIC ACTIVITY e lmitations on the specific activity of the primary coolant ensure that the res hour. doses at0BUNDARY will not exceed an appropriately small o art 100.

limits following a steam g ueru ueacciet periods with the primary coo pecific activityn> 1.0 4 1 OSE EQUIVALENT 1-13P1, but within the all imit shown, on Figure 3.4- 1, accommodates po odespkn phen n which may occur following changes in THERMAL POWER RE~PLACE if 1*4-t. P69-S S/f MILLSTONE - UNIT 2 B 3/4 4-4 Amendment No. 44-5-, 494, 266,

Attachment 3 MPS2 Marked-Up Bases Pages (Information Only)

INSERT 3/4.4.8 BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration 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-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 (Ref. 2).

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure the resulting offsite and control room doses meet the appropriate SRP acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of 150 gpd exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 p.Ci/gm DOSE EQUIVALENT 1-131 from LCO 3.7.1.4, "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.

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.0 gCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively. The second case assumes the initial reactor coolant iodine activity at 60.0 giCi/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 1100 g.Ci/gm DOSE EQUIVALENT XE-1 33.

The SGTR analysis assumes a rise in pressure in the ruptured SG causes radioactively contaminated steam to discharge to the atmosphere through the atmospheric dump valves or the main steam safety valves. The atmospheric discharge stops when the turbine bypass to the condenser removes the excess energy to rapidly reduce the RCS pressure and close the valves. The unaffected SG removes core decay heat by venting steam until the cooldown ends and the Shutdown Cooling (SDC) system is placed in service.

The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. 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 SDC system is placed in service.

Operation with iodine specific activity levels greater than l1Ci/gm but less than or equal to 60.0 pCi/gm is permissible for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1 g+/-Ci/gm LCO limit. Operation with iodine specific activity levels greater than 60 pCi/gm is not permissible.

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

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

LCO The iodine specific activity in the reactor coolant is limited to 1.0 gCi/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 1100 gCi/gm DOSE EQUIVALENT XE-1 33. The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).

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

APPLICABILITY In MODES 1, 2, 3, and4, 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 SRP acceptance criteria (Ref. 2).

In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS~and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.

ACTIONS

a. and b.

With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of four hours must be taken to demonstrate that the specific activity is < 60 ýtCi/gm. Four hours is required to obtain and analyze a sample. Sampling is continued every four hours to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were an iodine

spike, the normal coolant iodine concentration would be restored within this time period.

Also, there is a low probability of a SLB or SGTR occurring during this time period.

A statement in ACTION b. indicates the provisions of LCO 3.0.4 are not applicable. This exception to LCO 3.0.4 permits entry into the applicable MODE(S), relying on ACTIONS a.

and b. while the DOSE EQUIVALENT 1-131 LCO is not met. This exception is acceptable due to the significant conservatism incorporated into the RCS 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.

If the required action and completion time of ACTION b. is not met, or if the DOSE EQUIVALENT 1-131 is > 60 g.Ci/gm, the reactor must be~brought to HOT STANDBY (MODE

3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN (MODE 5) within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

d.

With the RCS DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-1 33 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period.

Also, there is a low probability of a SLB or SGTR occurring during this time period.

A statement in ACTION d. indicates the provisions of LCO 3.0.4 are not applicable. This exception to LCO 3.0.4 permits entry into the applicable MODE(S), relying on ACTION d.

while the DOSE EQUIVALENT XE-133 LCO is not met. This exception is acceptable due to the significant conservatism incorporated into the RCS 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.

e.

If the required action and completion time of ACTION d. is not met, the reactor must be brought to HOT STANDBY (MODE 3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN (MODE 5)

Within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS 4.4.8.1 Surveillance Requirement 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 Requirement provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance Requirement allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The surveillance 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-1 8 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the Surveillance Requirement 4.4.8.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-I 33 is not detected, it should be assumed to be present at the minimum detectable activity.

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

4.4.8.2 This Surveillance Requirement 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 of between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change Ž- 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

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

REFERENCES

1. 10OCFR 50.67.

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

3. ESAR, Section 14.1.5.

4. FSAR, Section 14.6.3.

REACTOR COOLANT SYSTEM BASES II ing T.Vg to < 515°F prevents the release of activity should a steam genera be rupture since the saturation pressure of t primary coolant is he lift pressure of the atmo c steam relief valves. The surveil equirements .e adequate assurance that excessive specific activity lev e primary coolant will be detected in sufficient time to orrec I tion. Information" obtained on iodine spik 1 be used to assess ameters associated with iodines pi enomena. A reduction in frequency o 0 ic t analyses owingpower changes may be permissible if justified o.d,t ained.

3/4.4.9 PRESSURE/TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to with-stand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 4.0 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. In addition, during heatup and cooldown evolutions, the RCS ferritic materials transition between ductile and brittle (non-ductile) behavior. To provide adequate protection, the pressure/temperature limits were developed in accordance with the 10CFR5O Appendix G requirements to ensure the margins of safety against non-ductile failure are maintained during all normal and anticipated operational occurrences. These pressure/temperature limits are provided in Figures 3.4-2a and 3.4-2b and the heatup and cooldown rates are contained in Table 3.4-2.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner, wall to tensile at the outer wall. These thermally induced compressive stresses atf the inside wall tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure- temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients estab-lished during heatup produce tensile stresses at.the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermally induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the, heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. Subsequently, for the cases in which the outer wall of the vessel becomes- the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

MILLSTONE - UNIT 2 B 3/4 4-5 Amendment No.440-

Serial No. 08-0135 Docket No. 50-423 ATTACHMENT 4 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR EVALUATION OF PROPOSED LICENSE AMENDMENT DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No. 08-0135 Docket No. 50-423 MPS3 Evaluation for Deletion of E BAR Attachment 4 Page 1 of 3 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR EVALUATION OF PROPOSED LICENSE AMENDMENT

1.0 INTRODUCTION

In accordance with the provisions of 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) is submitting a request for an amendment to the technical specifications (TS) for Millstone Power Station Unit 3 (MPS3). 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's Safety Evaluation (SE) dated September 27, 2006 (ADAMS ML062700612) (Reference 1).

Technical Specification Task Force (TSTF) change traveler TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to RCS Specific Activity Tech Spec" was announced for availability in the Federal Register on March 19, 2007 as part of the consolidated line item improvement process (CLIIP).

2.0 PROPOSED CHANGE

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

Revise the definition of DOSE EQUIVALENT 1-131 in MPS3 TS 1.10.

Replace MPS3 TS 1.11, "E Bar, AVERAGE DISINTEGRATION ENERGY" with new MPS3 TS 1.11, "DOSE EQUIVALENT XE-1 33."

'Revise LCO 3.4.8 to replace the gross radioactivity limit with a DOSE EQUIVALENT XE-133 limit.

0 Revise LCO 3.4.8 "Applicability" to specify the LCO is applicable in MODES 1, 2, 3, and 4.

a Delete Figure 3.4-1, "DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >1 p.Ci/gram DOSE EQUIVALENT 1-131,"

and Table 4.4-4, "REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM."

Replace the existing LCO 3.4.8 ACTIONS with the following:

A. An ACTION to verify DOSE EQUIVALENT 1-131 less than an absolute upper limit of 60 PCi/gram for all power levels once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when DOSE EQUIVALENT 1-131 is >1 AtCi/gram.

B. An ACTION permitting continued operation for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if DOSE EQUIVALENT'1-131 is >1 ACi/gram, but less than the specified upper limit, while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1 ItCi/gram limit.

Serial No. 08-0135 Docket No. 50-423 MPS3 Evaluation for Deletion of E BAR Attachment 4 Page 2 of 3 C. An ACTION to place the plant in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , if DOSE EQUIVALENT 1-131 is >1 PCi/gram for more than 48 continuous hours, or exceeds the absolute upper limit of 60 j.Ci/gram.

D. An ACTION permitting continued operation for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if the new DOSE EQUIVALENT XE-133 limit is exceeded while efforts are made to restore DOSE EQUIVALENT XE-1 33 to within the new LCO limit.

E. An ACTION to place the plant, in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , if DOSE EQUIVALENT XE-133 exceeds the LCO limit for more than 48 continuous hours.

0 Revise Surveillance Requirement (SR) 4.4.8 to replace Table 4.4-4 with two new requirements. SR 4.4.8.1 will verify DOSE EQUIVALENT XE-133 is within limits once per 7 days, and SR 4.4.8.2 will verify DOSE EQUIVALENT 1-131 is within limits once per 14 days, and between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a thermal power change of Ž_15% within a 1-hour period. A Note is added to both SRs, consistent with TSTF-490, to allow entry into MODES 2, 3, and 4 prior to performance of the respective SRs.

3.0 BACKGROUND

The background for this application is as stated in the model SE in NRC's Notice of Availability published on March 19, .2007 (72 FR 12838), the NRC Notice for Comment published on November 20, 2006 (71 FR 67170), and TSTF-490, Revision 0.

4.0 TECHNICAL ANALYSIS

DNC has reviewed References 1, 2 and 3, and the model SE published on November 20, 2006 (71 FR 67170) as part of the CLIIP Notice for Comment. DNC has applied the methodology in Reference 1 to develop the proposed TS changes. DNC has also concluded that the justifications presented in TSTF-490, Revision 0 and the model SE prepared by the NRC staff are applicable to MPS3 and support incorporation of this amendment into the MPS3 TS.

The proposed DOSE EQUIVALENT XE-133 limit is based on the revised reactor coolant radionuclide concentrations used in the Stretch Power Uprate (SPU) analyses submitted in DNC letter 07-0450 dated July 13, 2007. The reactor coolant radionuclide concentrations associated with the current specific activity limit of 100/E-bar correspond to greater than 1% failed fuel.

However, the reactor coolant radioiodine concentrations associated with the specific activity limit of 1 p.Ci/gram DOSE EQUIVALENT 1-131 correspond to 0.29% failed-fuel. For the SPU analyses, the more limiting condition of 0.29% failed fuel was used to define the reactor coolant radionuclide concentrations for both activity limits (100/E Bar and I gCi/gram DOSE EQUIVALENT 1-131). By comparison, if the current (non-SPU) licensing basis reactor coolant radionuclide concentrations were used to specify the new DOSE EQUIVALENT XE-133 limit, the limit would be 985 fiCi/gram, vice the proposed 81.2 ýtCi/gram.

Serial No. 08-0135 Docket No. 50-423 MPS3 Evaluation for Deletion of E BAR Attachment 4. Page 3 of 3

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 19, 2007 (72 FR 12838), the NRC Notice for Comment published on November 20, 2006 (71 FR 67170), and TSTF-490, Revision 0.

6.0 NO SIGNIFICANT HAZARDS CONSIDERATION DNC has reviewed the proposed No Significant Hazards Consideration determination published in the Federal Register on March 19, 2007 (72 FR 12838) as part of the CLIIP. DNC has concluded that the proposed determination presented in the notice is applicable to MPS3 and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).

7.0 ENVIRONMENTAL EVALUATION DNC has reviewed the environmental consideration included in the model SE published in the FederalRegister on March 19, 2007'(72 FR 12838) as part of the CLIIP. DNC has concluded that the staff's findings presented therein are applicable to MPS3 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 19, 2007 (72 FR 12838)

Serial No. 08-0135 Docket No. 50-423 ATTACHMENT 5 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR MARKED-UP TECHNICAL SPECIFICATION PAGES DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

DEFINITIONS$

CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

1) Capable of being closed by an OPERABLE containment automatic isolation valve system*, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.

b. All equipment hatches are closed and sealed,

c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of the Containment Leakage Rate Testing Program, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or O-rings) is OPERABLE.

1.8 DELETED CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

DOSE EQUIVALENT 1-131 iV 1.10 DOSE EQUIVALENT 1-131 shall be that c centration of I- (microCurie/gram) alone would produce the same,@BE liliai dose a the. -_-131, 1-132, 1-133, 1-134, and 1-135 actually present. The tih::.d ..... a..... f....

.................U. J ,. fse..1id...r. ih..Ai... in Federal Guidance Report No. II (FGR 11),

"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

In MODE 4, the requirement for an OPERABLE containment isolation valve system is satisfied by use of the containment isolation actuation pushbuttons.

MILLSTONE - UNIT 3 1-2 Amendment No. 8, 4-7, 4-86, 24-6, 5 P2e1e6&oA4.

AA

Nj'-b ,

&t-epiJkf wbe) &w-&fwcýJý lA ~ EP kTate

DEFINITIONS E-AVE UTEGRATION ENERGY 1.11 E shall be the average (wei ht the concentration of each radionuclide in the sample o te average beta and gamma e er disintegration for the radionuclides in the sample. A 1.12 DELETED I ,EL' ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.,

the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

1.14 DELETED FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1. 1.

LEAKAGE 1.16 LEAKAGE shall be:

L.16.1 CONTROLLED LEAKAGE CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals, and 1.16.2 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

.. b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or MILLSTONE - UNIT 3 1-3 Amendment No. 84, 8, -26, 4-8, 246, 220,-"tY

Attachment 5 MPS3 Marked-Up Technical Specification Pages INSERT 1.11 DOSE EQUIVALENT XE-133 1.11 . DOSE EQUIVALENT XE-133 shall be that concentration of Xe-1 33 (micro-Curie/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-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, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

REACTOR COOLANT SYSTEM .I-

',r - i 84- 488 01 3/4.4.8 SPECIFIC ACTIVITY I IMITING cONnaITIN FOR OPFRATION 3.4'.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microCurie per gram DOSE EQUIVALENT 1-131, and

b. Less than or equal icroCurie5 per gr-am. r.ne VILM APPLI CABILITY: MODES 1, 2, 3,4 ... p ACTIC"'I-Fi J

a. ith the specific activity of the reactor coolant greater n I1 roCurie per gram DOSE EQUIVALENT 1-131 for more t 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during e continuous time interval, or exceeding limit line shown on Figure .. 1, be in at least HOT STANDBY wi avg less than 500*F within rs; and avg

b. Withethe specific acti of the actor coolant greater than 100/E microCuries per gra , in at least. HOT STANDBY with Tav less than 500*F within 6-hoursav MODES 1, 2, 3, 4, and 5:

With the spec& c activity of the reactor coolant gr. er than 1 m~icroCu *per gram DOSE EQUIVALENT 1-131 or greater t. 100/E micro-Curie er gram, perform the sampling and analysis requiremen o0 I 4.a) of Table 4.4-4 until the specific activity of the reac coolant is restored to within its, limits.

U4%R~ 54.cmbrA MILLSTONE - UNIT 3 3/4 4-28

Attachment 5 MPS3 Marked-Up Technical Specification Pages INSERT 3.4.8 ACTION

a. With the specific activity of the reactor coolant > 1.0 microCurie per gram DOSE EQUIVALENT 1-131, verify DOSE EQUIVALENT 1-131 *60 microCuries per gram once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. With the specific activity of the reactor coolant > 1.0 microCurie per gram DOSE EQUIVALENT 1-131 but < 60 microCuries per gram, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1.0 microCurie per gram limit. Specification 3.0.4 is not applicable.

c. With the specific activity of the reactor coolant > 1.0 microCurie per gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or >

60 microCuries per gram DOSE EQUIVALENT 1-131, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

d. With the specific activity Of the reactor coolant > 81.2 microCuries per gram DOSE EQUIVALENT XE-133, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT XE-1 33 to within the 81.2 microCuries per gram limit. Specification 3.0.4 is not applicable.

e. With the specific activity of the reactor coolant > 81.2 microCuries per gram DOSE EQUIVALENT XE-1 33 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS

+ab *_A .4-h4-j~b'promac ft;;~ig rd l~i~cr T~bR1.1-1.

MILLSTONE -. UNIT 3 3/4 4-29

-Hlweýf ii, iyg+d 6E-T1-1331 RREAA COOLANT I CIFIC ACTI LIMIT VERSUS RATED TH AL POWER H THE REAC COOLANT SPECIFIC cI/or DOSE E LENT 1-13 MILLSTONE - UNIT 3 3/4,4-30 Amendment No. 46Q--

REACTOR\COOLANT SPE&FIC ACTIVITY"6AMPLE AND ANALYSK§ PROGRAM TYPE OF MEASURE NT SPLE AND MODES AND ANALYSIS N,\ FREQUENCY\ AND A

1. Gr s Radloact. vity At 14st once per"72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 2#

De rminatf on

2. Isoto c Analysis or DOSE EQ VA- 1 per 1\days.

LENT 1- 31 Concent tion

3. Radiochem al for E terminationk- 1 per 6 mon hs*1

4. otopic An ysis for I dine a) Once per hours, \, 2 In luding I-i 1, 1-133, nd 1-135 whenever tie specific activity eA.eeds 1.

pCi/gram DOSE EQUIVALENT IP4;31 W. .or IOl/E PCi/ ram of .

gross radloact *vity, and

" b) ne sample betwe n 2 1, 2, "d 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months follo ing a HHERMAL POWER c nge

\ex eeding 15%

of Of he RATED THERMA POWER within a 1-ho period.

-September i9, t,95ý9,

, _AAGAELE 4.4-,C4ontinuedL r adiochemic analysis f shallI sist of the quant.' ive measurement of the s ific acti for e . radionuclide, e.. -t for radionucljies with f-lives s than minutes and a radioiodines, Wis ntified in e react coolant. The cific activiti - r these individual di"onucl shall be used.4 the determina l* of - for the reacto cooant s le. etermina of the contre ors to C shall bo b upon thoo-1energy peaks Ad ifiable with a.

confidence leve

- Sample e taken after inimum of 2c and 20 days o WER OPERATION hay apsed since rea r was last s. ritical for 48 - rs or longer. T provisions of Spec.'cation . T04e not applica

Until the s ific activi of the React oolant System i stored within its limi,.

-fP(%1& PAW.

MILLSTONE - UNIT 3 3/4 4-32 AMENDMENT NO.-&&

Serial No. 08-0135 Docket No. 50-423 ATTACHMENT 6 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING DELETION OF E BAR MARKED-UP BASES PAGES (INFORMATION ONLY)

DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

LDD CIt IX,. uu-IvIr.)-'JLt)

~ 15, 2~Cc REACTOR COOLANT SYSTEM BASES 3747.47 DELETED 3/4.4.8 SPECIFIC ACTIVITY

-itaions on the'specific activity of the reactor coolant ensure that t EAB, LPZ and contro r ifl not exceed 10 CFR 5 egu atory Guide 1.183 dose criteria following a steam generat cident in conjunction with an assumed steady-state react -ary steam generator leakage rate o . values MILLSTO.N~E - UNfIT 3 B 3/4 4-5 Amendment No. 204.

-NR-e Yetb.. 't-.ýuu w ltdgemuern: O701jiM

Attachment 6 MPS3 Marked-Up Bases Pages (Information Only)

INSERT 314.4.8 BASES BACKGROUND The maximum dose that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 50.67 (Ref. 1). Doses to control room operators must be limited per GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration 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-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 (Ref. 2).

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure the resulting offsite and control room doses meet the appropriate SRP acceptance criteria following a SLB or SGTR accident. The safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant is at the LCO limits, and an existing reactor coolant steam generator (SG) tube leakage rate of I gpm exists. The safety analyses assume the specific activity of the secondary coolant is at its limit of 0.1 iCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.1.4, "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.

The safety analyses consider two cases of reactor coolant iodine specific activity. One case assumes specific activity at 1.0 piCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the rate of release of iodine from the fuel rods containing cladding defects to the primary coolant immediately after a SLB (by a factor of 500), or SGTR (by a factor of 335), respectively. The second case assumes the initial reactor coolant iodine activity at 60.0 jiCi/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 81.2 pCi/gm DOSE EQUIVALENT XE-1 33.

The SGTR analysis also assumes a loss of offsite power at the same time as the reactor trip. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature AT signal.

The loss of offsite power causes the steam dump valves to close to protect the condenser.

The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the

atmosphere through the SG power operated relief valves and/or the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends and the Residual Heat Removal (RHR) system is put in service.

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

Operation with iodine specific activity levels greater than 1jlCi/gm but less than or equal to 60.0 ýiCi/gm is permissible for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while efforts are made to restore DOSE EQUIVALENT 1-131 to within the 1 ýtCi/gm LCO limit. Operation with iodine specific activity levels greater than 60 PCi/gm is not permissible.

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

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

LCO The iodine specific activity in the reactor coolant is limited to 1.0 p.Ci/gm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 81.2 pCi/gm DOSE EQUIVALENT XE-133. The limits on specific activity ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).

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

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

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

ACTIONS

a. and b.

With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of four hours must be taken to demonstrate that the specific activity is < 60 giCi/gm. Four hours is required to obtain and analyze a sample. Sampling is continued every four hours to provide a trend.

V' The DOSE EQUIVALENT 1-131 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period.

Also, there is a low probability of a SLB or SGTR occurring during this time period.

A statement in ACTION b. indicates the provisions of LCO 3.0.4 are not applicable. This exception to LCO 3.0.4 permits entry into the applicable MODE(S), relying on ACTIONS a.

and b. while the DOSE EQUIVALENT 1-131 LCO is not met. This exception is acceptable due to the significant conservatism incorporated into the RCS 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.

If the required action and completion time of ACTION b. is not met, or if the DOSE EQUIVALENT 1-131 is > 60 PLCi/gm, the reactor must be brought to HOT STANDBY (MODE

3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN (MODE 5) within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

d.

With the RCS DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period.

Also, there is a low probability of a SLB or SGTR occurring during this time period.

A statement in ACTION d. indicates the provisions of LCO 3.0.4 are not applicable. This exception to LCO 3.0.4 permits entry into the applicable MODE(S), relying on ACTION d.

while the DOSE EQUIVALENT XE-133 LCO is not met. This exception is acceptable due to the significant conservatism incorporated into the RCS 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.

e.

If the required action and completion time of ACTION d. is not met, the reactor must be brought to HOT STANDBY (MODE 3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN (MODE 5) within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS 4.4.8.1 Surveillance Requirement 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 activitie's and the gaseous gamma activities in the sample taken. This Surveillance Requirement provides an -indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance Requirement allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The surveillance 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-1 8 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the Surveillance Requirement 4.4.8.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-I133 is not detected, it should be assumed to be present at the minimum detectable activity.

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

4.4.8.2 This Surveillance Requirement 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 of between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change :- 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

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

Surveillance Requirement to be performed in those MODES, prior to entering MODE 1.

REFERENCES

1. 10OCER 50.67.

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

3. FSAR, Section 15.1.5.

4. FSAR, Section 15.6.3.

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) for the limits on specific activity represent limits based upon a parametric evalhation by the NRC typical site locations. These values are conservative in that specific site parameters. of the Mi one site,*such as SITE BOUNDARY location and meteoroilgical conditions, were not consid in this evaluation.

The A ION statement permitting POWER OPERATION-to continue for ited time periods with the r tor coolant's specific activity greater than 1 microCurie/gr DOSE EQUIVALENT 1-131, ut within the allowable limit shown on Figure 3.4-1 ccommodates

,possibleiodinespiking p, omenon which may occur following chaingin THERMAL The sample analysis for dete ining the gross specifi ctivity and E can exclude the radioiodines because of the low reactor olant limit of 1 icroCurie/gram DOSE EQUIVALENT 1-131, and because, if the i it is exce ed, the radio iodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the gross specific tj ty level and radioiodine level in the reactor coolant were at their limits, the radioiodine con *on would be approximately I%. In a release of reactor coolant with a typical mixture of ioactivi the actual radioiodine contribution would probably be about 20%. The exc -ion of radionu 'des with half-lives less than 10 minutes from these determinations h een made for severn ons. The first consideration is the difficulty to identify short-liv radionuclides in a sample th equires a significant time to collect, transport, and, analyz he second consideration is the pre table delay time between the postulated release of r io-activity from the reactor coolant to its rele e to the environment and transport to the SI BOUNDARY, which is relatable to at least 30 min es decay time. The choice of 10 nt for the half-l ife cutoff was made because of the nuclear ch cteristics of the typical reactor olAant radioactivity. The radionuclides in the typical reactor cool have half-lives of less an 4 minutes or half-lives of greater than 14 minutes, which allows a di ction betwee 2e radionuelides above and below a half-life of 10 minutes. For these reasonW radi clides that are excluded from consideration are expected to decay to very low levels ore they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.

MILLSTONE - UNIT 3 B 3/4 4-6 Amendment No.

A-.. ,.feJgJ

.. bylM -etter IaRed U*=175i0

LDDCPGA4@. 05W-3-06 REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

Reuig ýv t 00'F prevents the releaseofsulastmgertr tube rupture since the saturation press~re0ant is beo the lift pressure of the atmospheric steam relief valves- Thez!:S ;;itzrai'tTac'e R it rvde adequate assuran .ce that excessive, specific activi in the reactor coolant will be e n sufficient time to take corrective acio pe uction if justified by in frequency the data obtained. of isotopic analyses following ages maybe JN J

'i 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM (EXCEPT THE PRESSURIZER).

BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

Figures 3.4-2 and 3.4-3 contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.

Each P/T limit curve defines an acceptable region for normal operatin. The usual use of the curves is operational requirements during heatup or cooldown maneuYering, when pressure and temperature indications are monitored and compared to the applicable curve to.determine that operation is within the allowable region. A heatup or cooldown is .defined as a temperature increase or decrease of greater than or equal to 10F in any one hour period. This definition of heatup and cooldown is based upon the ASME definition of isothermal conditions described in ASME,Section XI, Appendix E.

MILLSTONE -.UNIT 3 B 3/4 4-7. Amendment No. 1-57-,4W, A- 1,-.. e-4 Y, L A'711 "Al aU a

@KVve-07 L

v t e Letter Sequence Request
TAC:MD8492, Deletion of E BAR Definition and Revision to RCS Specific Activity Tech Spec (Approved, Closed)
Initiation Request Acceptance Administration Questions 25 September 2008 Answers 30 September 2008 Results Approval Other:
MONTHYEAR2008-09-25 [Table View]

ML080920781

Text

1.0 INTRODUCTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

8.0 REFERENCES

1.0 INTRODUCTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

8.0 REFERENCES

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