ML092180973

E-mail Regarding Placement of Document in ADAMS for Millstone, Unit 3, Issuance of Amendment No. 59

ML092180973
Person / Time
Site:	Millstone
Issue date:	08/03/2009
From:	Nelson R Division of Operating Reactor Licensing
To:	Mark King NRC/RGN-III
References
TAC 76066, FOIA/PA-2011-0115
Download: ML092180973 (29)

Text

Sanders, Carleen From: Nelson, Robert Sent: Monday, August 03, 2009 1:38 PM To: King, Mike Cc: Sanders, Carleen

Subject:

Action: Placement of Document in ADAMS The letter from David Jaffe, NRC, to Edward J. Mroczka, Connecticut Yankee Atomic Power Co., dated January 25, 1991, Docket 50-423, "Issuance of Amendment (TAC No. 76066)" is considered to be the original and should be entered in ADAMS.

Carleen Sanders of our staff will deliver the document to you shortly.

Thanks, Robert A. Nelson Deputy Director Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Phone: (301) 415-1453 Fax: (301) 415-2102 I.

ASCII Text UNITED STATES NUCLEAR REGULATORY COMMISSION WASH1NGTON, D. C. 205151 January 25, 1991 Docket No. 50-423 RECEIVED Mr. Edward J, Mroczka Senior Vice President JAN 2 8 1991 Nuclear Engineering and Operations Connecticut Yankee Atomic Power Compay SENIOR VICE PRE$r:u-NT Northeast Nuclear Energy Company Nudoa Engir iNg &ODrdral.;s Post Office Box 270 Hartford, Connecticut 06141-0270

Dear Mr. Mroczka:

SUIJJECT: ISSUANCE OF AMENDMENT (TAC NO. 76066)

The Commission has issued the enclosed Amendmernt No. 59 to Facility Operating License No. NPF-49 for Millstone Nuclear Power Station, Unit No. 3, in response to your application dated February 26, 1990, as supplemented April 30, December 6 and 19, 1990.

The amendment modifies the Technical Specifications to allow an increase in the normal containment pressure range. The revised containment pressure range is 10.6 psia to 14.0 psia.

A copy of the related Safety Evaluation Is also enclosed. Also enclosed Is the Notice of Issuance which has been forwarded to the Office of the Federal Registe for publication.

Sincerely, i d H. ffe roject Manager Project Directorate 1-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Enclosures:

1. Amendment No, 59 to NPF-49

2. Safety Evaluation

3. Notice cc w/enclusures:

See next page

I ASCII TextI Fir. E. J. Mroczka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. 3 cc:

Gerald Garfield, Esquire R. K. Kacich, Manager Day, Berry and lioward Generation Facilities Licensing Counselors at Law Northeast Utilities Service Company City Place Post Office Box 270 Nartford, Connecticut 06103-3499 Hartford, Connecticut 06141-0270 V. 1. Romberg, Vice President 0. 0. Hordquist Nuclear Operations Director of Quality Services Northeast Utilities Service Company Northeast Utilities Service Company Post Office Box 270 Post Office Box 270 Hartford, Connecticut 06141-0270 Hartford, Connecticut 06141-0270 Kevin ¶cCarthy, Director Regional Administrator Radiation Control Unit Region I Department of Environmental Protectior U. S. Nuclear Regulatory Commission State Office Building 476 Allendale Road Hartford, Connecticut 06106 King of Prussia, Pennsylvania 19406 Bradford S. Chase, Under Secretary First Selectmen Energy Division Town of Waterford Office of Policy and Management Flail of Recores V-0 Washington Street 200 Boston Post Road Hartford, Connecticut 06106 Waterford, Connecticut 06385 S. E. Scace, Nuclear Station Director W. 3. Raymond, Resident Inspector Millstone Nuclear Power Station Millstone Nuclear Power Station N*ortheast Nuclear Energy Company c/o U. S. Nuclear Regulatory Commission Post Office Box 1?8 Post Office Box 811 1-aterford, Connecticut 06385 Nlantic, Connecticut 06357 C. H. Clement, Nuclear Unit Director M. R. Scully, Executive Director Millstone Unit 1lo. 3 Connecticut Municipal Electric Northeast Nuclear Energy Company Energy Cooperative Post Office Box 128 30 Stott Avenue Waterford, Connecticut 06385 Norwich, Convecticut 06360 Ms. Jane Spector Mr. Alan Menard, Manager federal Energy Regulatory Conmission Technical Services 825 N. Capitol Street, NIE. Massachusetts MunicipBl Wholesale Room 8608C Electric Company Washington, D.C. 20426 Post Office Box 426 L[dlow, Massachusetts 01056 Burlingtor Electric Department c/o Robert t, fletcher, Esq.

271 South Union Street Burlington, Vermont 05402

ASCII Text I NUCLEAR UNITED STAATES NUCLEAR REGULATORY COMMISSION WASHIPIOTON. DC 05 NORTHEAST NUCLEAR ENERGY COMPANY,_ET AL.

DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION,1 UNIT 1NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 59 License No. NPF-49

1. The Nuclear Regulatory Commission (the COmission) has found that:

A. The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee) dated February 26, 1990, as supplemented April 30, December 6 atd 19, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CPR Chapter 1; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commilssion's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is In accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

ASCII Text

-2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated In the attachment to this license amendrnent, and paragraph 2.C.(2) of Facility Operating License No. RPF-49 is hereby amended to read as follows:

(2) Technical Specifications The Technical Speciffications contained In Appendix A, as revised through Amendment No. 59 , and the Environmental Protection Plan contained in Appendix $, both of which are attached hereto are fiereby incorporated in the license. The licensee shall operate the facility In accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendnrent Is effective as of the date of its issuance, to be implemented within 30 days of issuance, FOR THE NUCLEAR REGUVITORY COMMISSION JojF. Stolz, Direc tr (Py ject Directorate 1-4

'-liision of Reactor Projects T/11 Office of Nuclear Reactor Regulatiun

Attachment:

Changes to the Technical Specificailons Date of Issuance: January 25, 1991

ASCII Text ATTACHMENT TO LICENSE AMENDMENT NO._._

FACILTIY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines icdicating the areas of change.

Remove Insert viii viii fx ix 3/4 6-1 3/4 6-1 3/4 6-? 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-4 3/4 6-4 3/4 6-5 3/4 6-5 3/4 6-6 3/4 6-6 3/4 6-7 .3/4 6-7 3/4 6-8 3/4 6-8 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 314 6-2

SII Text NIMITING CONDITIONS FOR OPERATION AND, SURVEILLANE,..REQUIREMENTS SECTION P.G..

FIGURE 3,4-1 DDSE EQUIVALENT 1131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

• P*Ci/gram DOSE EQUIVALENT 1131 ............... 3/4 4-30 TABLE 4,4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM, ................................ 3/4 4-31 3/4.4.9 PRESSURE/TEMPERATURE LIHITS Reactor Coolant System ................................... 3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -

APPLICABLE UP TO 10 EFPY ............... 3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS -

APPLICABLE UP TO io EFPY ................................. 3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITHDRAWAL SCHEDULE "........................... 3/4 4-36 Pressurizer "................. .......... 3/4 4-37 Overpressure Protection Systems ................. ........ 3/4 4-38 FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (FOUR LOOP OPERATION) ................. 3/4 4-40 FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE POR SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION) ............... 3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY ..................................... 3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS ............................. 3/4 4-43 3/4.5 EMERGENCY_ CORE CO.OLING, SYSTEMS 3/4.5.1 ACCUMULATORS ............................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tar9 GREATER THAN OR EQUAL TO 350.... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350 ................... 3/4 5-7 3/4-5.4 REFUELING WATER STORAGE TANK............................ 3/4 5-9 314.5 CONTAINMENT SYSTEM$

3/4.6.1 PRIMARY CONTAINMENT Containment Integrity .... ,...., ...... 3/4 6-1 Containment Leakage ................. 3/4 6-2 TABLE 3.6-1 ENCLOSURE BUILDING BYPASS LEAKAGE PATHS............... 3/4 6-4 Containment Air Locks ................

j

....... 3/4 6-5 Containment Pressure,... ...... i.................. 3/4 6-7

'ASCII Text INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE9UIREMENTS SECTION PAGE Air Temperature ................................... 3/4 6-9 Containment Structural ntegr ................. /6-10 Containment Ventilation System...,............, 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Quench Spray System .................... 3/4 6-12 Recirculation Spray System ......................... 3/4 6-13 Spray Additive System ............................ 3/4-6-14 3/4.6.3 CONTAINMENT ISOLATION VALVES ....................... 3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors .................................. 3/4 6-35 Electric Hydrogen Recombiners ...................... 3/4 6-36 FIGURE 3.6-2 HYDROGEN RECOMBINER ACCEPTANCE CRITERIA FLOW VS.

CONTAINMENi PRESSURE.............. 3/46-36a 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector .............................. 3/4 6-37 3/4.6.6 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System ... 3/4 6-38 Enclosure Building Integrity ....................... 3/4 6-40 Enclosure Building Structural Integrity ............ 3/4 6-41 3/4,7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves .... .. ,..................................... 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION;........ ........... 3/4 7-2 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES THREE LOOP OPERATION.................. ........... 3/4 7-2

ASCII Text 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENH CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3,6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

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

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within I hour or be in at least HOT STANDBY within the next 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 />.

SIJRVE ILLANC L8EOUIREMENTS...

4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations* not capable of being closed by OPERABLE containment automatic isolation valves or operator action during periods when containment isolation valves are opened under administrative control,** and required to be closed during accident conditions are closed by valves, bMind flanges, or deactivated automatic valves secured in their positions.

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressure not less than P , 53.27 psia (38.57 psig), and verifying that when the measureb leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1,2d. for all other Type B I

and C penetrations, the combined leakage rate is less than 0.60 L8 -

• Except valves, blind flanges, and deactivated automatic valves which are located Inside the containment and are locked , sealed, or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

• • The following manual valves may be opened on an intermittent basis under administrative control. 3FPW-V66I, 3FPW-666, 3SPSVI3, 3SSP-VI44 3HCS-V2, 3HCS-V3, 3HCS-V9, 3HCS-VIO, 3HCS-V6, 3NCS-V13, 3SAS-Y875, 3SAS-VSO, 3CHS-V371, 3CCP-V8861 3CCP-V887, 3CVS-V13.

ASCII Text CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITIN.G CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or equal to L 0.65% by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa j

53.27 psla (38.57 psig);

b. A combined leakage rate of less than 0.60 L for all penetrations and valves subject to Type B aid C tests, 'Aen pressurized to Pa, and

c. A combined leakage rate of less than or equal to 0.042 L for all penetrations identified In Table 3.6-1 as Enclosure NuildiO leakage paths when pressurized to Pa" g bypass I APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

W1ith the measured overall integrated containment leakage rate exceeding 0.75 1h, or the measured combined leakage rate for all penetrations and valves sfibject to Type B and C tests exceeding 0.60 L , or the combined bypass leakage rate exceeding 0.042 L , restore the overa1l integrated leakage rate to less than 0.75 L , the combined leakage rate for all penetrations subject to Type B and C test to less than 0,60 L , and the combined bypass leakage rate to less than 0.042 La prior to incAaslng the Reactor Coolant System temperature above 200"F.

S.U.RYEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using methods and provisions of ANSI N45.4-1972 (Total Time Method) and/or ANSI/ANS 55.8-1931 (Mass Point Method):

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 +/- 1Omonth intervals during shutdown at a pressure not less than P#*53,27 psia (38.57 psig) during each tO-year service period, he third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection;

b. If any periodic Type A test falls to meet 0.75 L_, the test schedule for subsequent Type A tests shall be reviewed nd approved *by the Commission. If two consecutive:Type A tests fail to meet 0.75 L., a Type A test shall be performed at least every 18 months untllatwo consecutive Type A tests meet 0.75 La at which time the above test schedule may be resumed;

j ASCII exte CONTAINMENT SYSIEMS SURVEILLANCEREQUIREMENTS (Continued)

c. The accuracy of each Type A test shall be verified by a supplemental test which:

1) Confirms the accuracy of the test by verifying that the supple-mental test results, L , minus the sum of the Type A and the superimposed leak, L0, 9 s equal to or less than 0.25 La;

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and

3) Requires that the rate at which gas is injected Into the containment or bled from the containment during the supplemental test is between 0.75 La and 1.25 La.

d. Type B and C tests shall be conducted with gas at P 53.27 psia (38.57 involving:

tests psig), .at intervals no greater than 24 montl except for

1) Air locks

e. The combined bypass leakage rate shall be determined to be less than or equal to 0.042 L by applicable Type B and C tests at least once per 24 months except for penetrations which are not Individually testable; penetrations not individually testable shall be determined to have no detectable leakage when tested with soap bubbles while the containment is pressurized to P., 53.27 pslg (38.57 psig), during each Type A test;

f. Air locks shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.3;

g. Purge supply and exhaust isolation valves shall be demonstrated OPERABLE by the requirements of Specifications 4.6.3.2.c and 4.9.9.

h. The provisions of.Speclflcatfon 4.0.2 are rot applicable.

Amendment No. 59

ASCII Text ENCLOSURE BUILDING BYPASS LEAKAGE PATHS PENETRATI' ON SDESCRJPTWIN RELEASE LOCATION 14 X2 to Safety Injection Tanks Ground Release 15 Primary Water to Pressurizer Ground Release Relief Tanks 35 Vacuum Pump Suction Plant Vent 36 Vacuum Pump Suction Plant Vent 37 Air Ejector Suction Plant Vent 38 Chilled Water Supply Plant Vent 45 Chilled Water Return Plant Vent 52 Service Air Turbine Building Roof Exhaust 54 Instrument Air Turbine Building Roof Exhaust 55 Fire Protection Ground Release

59. Fuel Pool Purification Ground Release 60 Fuel Pool Purification Ground Release 70 Demineralized Water Ground Release 72 Chilled Water Supply Plant Vent a5 Containment Purge Ground Release 86 Containment Purge Plant Vent 116 Chilled Water Return Plant Vent 124 Nitrogen to Containment Plant Vent

EASCIJ Text CONTAIINM SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION__

3.6.1.3 The containment air lock shall be OPERABLE with;

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.05 L at Pd1 53,27 psha (38.57 psig).

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

1 ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed* and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERABLE air lock door closed,

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days,

3. Otherwise, be in at least HOT STANDBY within the next 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 />, and

4. The provisions of Specification 3.0.4 are not applicable.

b. With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 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 />.

• Except during entry to repair an inoperable Inner door, for a cumulative time not to exceed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per year.

ASCII Text CONTAINMENT SYSTEMS SURVEILtANC.E.REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. 1) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,. by verifying no detectable seal leakage by pressure decay when the volume between the door seals is pressurized to greater than or equal to Pa, 53.27 psla (3B.57 psig), for at least 15 minutes; I

or

2) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying that the seal leakage is less than 0.01 L as determined by precision flow measurements when measured f8r at least 30 seconds with the volumQ between the seals at a constant pressure of greater than or equal to Pa' 53.27 psia (38.57 psig);

or

3) Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by completing an'overall air lock leakage test per 4.6.1.3,b.

b. By .conducting overall 'air lock leakage tests at not less than P 53.27 psia (38.57 pslg), and verifying the overall air lock leakaie rate is within its limit:

1) At least once per 6 months,* and

?) Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.**

c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

• The provisions of Specification 4.0.2 are not applicable.

• • Ihis represents an exemption to Appendix 0, paragraph III.D.2.(b)(ii), of 10 CFR Part 50.

ASCII Text CONTAINMENI-SYSTEMS CONTAINMENT PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment pressure shall be maintained greater than or equal to 10.6 psla and less than or equal to 14.0 psla.

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

With the containment pressure less than 10.6 psia or greater than 14.0 psia, restore the containment pressure to within the limits within I hour or be in I

at least HOT STANDBY within the next 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 4.6.1.4 The primary containment pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. I

ASCII Text This Page Intentionally Left Blank

LASCII Text 3/4-6 CONTAI MENT SYSIENS BASES ...

3/4.6.1 PRIMARY CONTA1INMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses, This restric-tion, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guidelines of 10 CFR Part 100 during accident conditions and the control room operators dose to within the guidelines of GDC 19.

3/4,6.1.2 CONTAINMENT LEAKAG.

The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix J of 10 CFR Part 50.

3/4.6..t C.O.NTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock-leakage tests.

3/4.6.1A- and 3/4.6.1.5 AIR PRESSURE and AIR TEMPERATURE The limitations on containment pressure and average air -temperature ensure that: (1) the containment structure is prevented from exceeding its design negative pressure of 8 psia, and (2) the containment peak pressure does not exceed the design pressure of 60 pska during LOCA conditions. Measure-inents shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature. The limits on the pressure and average air temperature are consistent with the assumptions of the safety analysis. The minimum total containment pressure of 10.6 psia is determined by summing the minimum permissible air partial pressure of 8,9 psia and the maximum expected vapor pressure of 1.7 psia (occurring at the maximum permissible containment initial temperature of 1204F).

ASCII Text CONTAINMENT SYSTEMS BASES

$/4.,,I.6 CONTAINMELT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility, Structural integrity is required to ensure that the containment will withstand the maximum pressure of 60 psia in the event of a LOCA. A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.7 -CONAINMENTVENTILATION SYSTEM The 42-inch containment purge supply and exhaust isolation valves are required to be locked closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inadvertently opened, the valves are locked closed In accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents powar from being supplied to the valve operator.

The Type C testing frequency required by 4.6,1.2d is acceptable, provided that the resilient seats of these valves are replaced every other refueling outage.

3/4.6.2 DEPRESSURIZATION AND COOLINGKSYSTEMS

/4..6.2.1 and 3/4.6,2.2 CONTAINMENT _UENCH SPRAY SYSTEM and kECIRCULATION SPRAY SYSTEM The OPERABILITY of the Containment Spray Systems ensures that containment depressurizatlon and iodine removal will occur in the event of a LOCA. The pressure reduction, iodine removal capabilities and resultant containment leakage are consistent with the assumptions used in the safety analyses.

3/4.6,2.3 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficlept NaDH is added to the containment spray in the event of a LOCA. The limits on faOH volume and concentration ensure a pH value of between 7.0 and 7.35 for the solution recirculated within containment after a LOCA. This pH band minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes an a*lowance for water not usable because of tank discharge line location or other physical charac-teristics.

AmendmenL No. 59

ASCII Text

• 'UNIT6D STATES NUCLEAR REGULATORY COMMISSION WASA4HINGTON, 0. C, 20665 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 59 TO FACILITY OPERATING LICENSE NO. NPF-49 NORTHEAST NUCLEAR ENERGY COMPANYF,_ET_AL.

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 3 DOCKET NO. 50-423

1.0 INTRODUCTION

By application for lfcerise amendment dated February 26, 1990, as supplemented April 30, December 6 and 19, 1990, Northeast Nuclear Energy Company, et al.

(the licensee), requested changes to Millstone Unit 3 Technical Specifications

{TS) regarding normal containment operating pressure. The current IS require that the containment pressure be maintained subatmospheric and be greater than 8.9 psia but less than or equal to I? psia during operation Modes I through 4.

The licensee proposed to change the cDntainment operating pressure and associated IS to a new range between 10.6 psia and 14.0 psla.

2.0 DISCUSSION Millstone Unit 3 is a dual-containment plant. The containment is comprised of a primary containment structure and a secondary containment enclosure building aýnd an associated supplementary leak collection and release system (SLCRS).

Containment entries are required for inspecting unidentified reactor coolant system leakage, investigating boron precipitation, and plant sta6rt-up surveillances or inspections. The risk of injury to plant personnel performing such physical labor in the subatomospheric containment has been found significant due to crossing the pressure boundary and also due to oxygen deficiency. Personnel are required to wear self-contained respirator (Rexnord "Bio-Packs") to supply supplemental oxygen but the environment of low pressure and high temperature in the containment causes significant potential for personnel Injury during containment entries. The licensee stated that 38 personnel medical incidents had occurred due to containment entries during the past 4 years since the plant was licensed. In addition , the use of Bio-Packs cause personnel working in the containment to become less efficient.

In order to allow containment entry wifth a minimal pressure change and eliminate the need to carry heavy, awkward supplemental oxygen units (Bio-Packs), the licensee proposes to increase the containment operating pressure. In support of the TS change, the licensee performed safety aralyses

ASCII Text to assess the impact on the accidents evaluated as the design basis, the potential for creation of a 6ew unanalyzed eyeit, and the impact cn the margin of safety. The staff's evaluation of the licensee's submittals Is described below.

3,0 EVALUATION The current containment parameters and the licensee proposed changes are listed in Table 1. The licensee's revised safety analyses are based on the proposed parameters.

Table I Containment Parameter Current Proposed Change Normal Operating Pressure 9.8 psia 14.0 psia Design Pressure 45 psi9 45 psig

.Peak Pressure (Pa) 36.1 pslg 38.57 psig Containment Leak Rate (La) 2912.6B SCFII 2206.33 SCFH (0.9 wt% per day) (0.65 wt% per day)

Secondary Containment Bypass Leakage Fraction 0.O-La O,042La (0.009 wt% per day) (0.028 wt% per day)

Service Water Temperature 750 F 75OF 3.1 Containment Integrity Analysis 3.1.1 Containment Pressure and Temperature Responses Two loss-of-coolant-accident (LOCA) cases for containment pressure/temperature responses were reanalyzed by the licensee using the same methods and computer models as described in Section 6.2.1 of the Final Safety Analysis Report (FSAR) except the initial containment pressure was increased to 14.2 psig.

The licensee reanalyzed thu hot leg double-ended rupture (DER) and the pump suction DER with failure of one engineering safety features (ESF) train. The limiting accident for peak containment pressure was found to be the hot leg DER at 3B.57 psig which was below the containment design pressure of 45 pslg.

Since the staff has previously reviewed and approved the methodology and analytical model, the staff concludes that the licensee's LOCA analysis is acceptable.

The pump suction DER with failure of one ESF train was found tu be the limiting accident for the long term containment pressure transient. The current analysis showed that the containment pressure depressurized to atmospheric pressure in 41.33 minutes after a LOCA and then the containment

ASCII Text

'3 -

pressure returned to subatmnopheric. The licensee recalculated this pressure transient and the result showed that the containment pressure remains above atmospheric pressure for the duration of the accident. The staff's review found that containment pressure remaiing above atmospheric would cause continued leakage from the containment. This will be further discussed in Section 3.3 of this evaluation.

3.1.2 Main Steam Line Break Analysis The licensee rtcalculated the containment pressure response for a vinin steam line break (NSLB) for full DER at hot standby (zero power). The peak containment pressure based on a new containment operating pressure of 14.2 psia was calculated to be 34.5 psig which was below the peak containment pressure following a LOCA. The staff concludes that the MSLB reanalysis has a minor effect on the containment pressure responses.

3.1.3 Subcompdrtmnent Pressurization Analysis The initial atmospheric conditions within the subcompartmett which can maximize the differential pressure across the walls are the maximum allowable temperature, minimum absolute pressure, and zero percent relative humidity.

Increasing initial pressure will increase air mass in the compartment and reduce pressure difference across the walls. Therefore, the staff concludes that the proposed change has no effect on current contaiment subcompartment analysis.

3.1.4 Combustible Gas Concentration The increased conttdnment operating pressure will result in lower hydrogen concentration In the cuntainment because the rate of hydrogen generation is utchanged but the mass of air in the containment is increased. Therefore, the staff concludes that the proposed change has no effect on current evaluation of hydrogen control.

3.2 Safety Systems Evaluation 3.2.1 Quench Spray System/Containment Recircutatlon System The Quench S pray System (QSS) and the Containment Recirculation System (CRS) had previously been reviewed and approved by the NRC staff for their containment pressure reduction and core cooling roles, respectively, The licensee now proposes to credit the QSS and CRS for removal of post-LOCA fission products inside containment.

The ERC staff has reviewed the QSS and CkS against the criteria of Standard Review Plan (SRP) 6.5.2, Revision 2, "Contalrient Spray as a Fission Product Cleanup System." In a letter dated December 6, 1990, the licensee addressed the criteria of SRP 6.5.2, Revision 2 regarding the QSS and CRS.

ASCII Text 4 -

1he staff concludes that the conitainment spray system as a fission product cleaiup system is acceptable and meets the relevant requirements of General Design Criterion 41, "Containment Atmosphere Cleanup," general Design Criterion 42, "Inspection of Containment Atmosphere Cleanup Systems," and General Design Criteriori 43, "Testlng of Containment Atmosphere Cleanup Systems." This conclusion is based on the following.

The concept upon which the proposed system is based has been demonstrated to be effective for iodine absorption and retention under post-accident cojiditions. The proposed system design is an acceptable application of this concept. The system provides suitable redundancy in components and features such that its safety function can be accomplished assuming a single failure.

The staff concludes that the system meets the requirements of Geceral Design Criterion 41.

The proposed pre-operational tests, post-operational testing and surveillance, and proposed limiting conditions of operation for the spray system provide adequate assurance that the iodine scrubbing function of the containment spray system will meet or exceed the effectiveness assumed in the accident evaluation and, therefore, meets the requirements of General Design Criteria 42 and 43.

3.2.2 Containment Air Recirculation System The containment air recirculation (CAR) system is not designed to operate ptst-LOCA and is automatically shut down by a containment depressurization actuation signal. Therefore, the proposed change has no effect on the consequences of a DBA due to the CAR system performance.

3.2.3 Contairnment Vacuum System The containment vacuum system reduces the containment pressure from atmospheric to subatmospheric using a vacuum ejector. -The proposed change will result in less frequent operation of the vacuum pump in order to mointain the new subatnvospheric pressure. The system is not safety related.

Therefore, the staff concludes that the proposed change has no effect on the consequences of a DBA due to the containment vacuum system performance.

3.2.4 Containment Pressure Monitors At the present time, there are two narrow range containment pressure transmitters (3LNS&P143A and B) that provide indication in the control room for a containment pressure range of 8.5 to 13,5 psla during normal operation.

These transmitters and associated instrumentatlon/displays will be modified prior to implementing the proposed changes to the TS to achieve a range of 8,5 to 14.5 psi as indicated Ui the licensee's letter dated December 19, 1990. Ve find this cog=mitment to be acceptable.

ASCII Text 3.3 Containment Leakage Evaluation The current containment integrity analysis assumed that the containment pressure would drop to approximately 4 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after a LOCA and then the containment would be uaintained subatmospheric for 30 days. TUe current containment integrated leak rate was set at La, or 0.9% by weight of the containment air per day (0.9 wt%/day), for the first hour of a LOCA and zero leakage after the containment returned to subatmopheric. The proposed change in contairnent operating pressure will result in containment pressure remaining above atmospheric for the duration of the accident and, therefore, continued containment leakage is assumed.

To compensate for the increased time inleakage release, the licensee proposed to reduce the TS allowable leak rate from 0.9 wt%/day to 0.65 wt%/day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.326 wt%/day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until 30 days. The licensee stated that the proposed limit of 0.65 wt%/day represents the maximuzn containmeiit allowable leakage in compliance with 10 CFR Part 100 requirements. The licensee provided containment integrated leak rate test (CILRT) results for the second refueling outage. The as-left containment leakage rate was 0,2919 wt%/day or 641 SCFH. The current acceptable leakage for the CILRT is 0.75Ua(0.9), or 0.575 wt%/day, which corresponds to an allowable leakage rate of 1428 SCFH.

The proposed containment leakage rate is 0.75La(0.65), or 0.488 wt%/day, which corresponds to an allDwable leakage rate of 1076 SCFH. The staff finds that the proposed containment leakage rate is equivalent to 0.52La which is less than 0.75La required by Appendix 3 to 10 CFR Part 50. Furthermore, the CItR~s were purformed at Pa of 39.4 psig which was higher than the proposed new test pressure of 38.6 psig. The CILRT result would be lower if the tests were performed with the new test pressure. Based on the licensee provided Information, the staff concludes that the proposed containment leakage rate is conservative and acceptable.

The licensee proposed to increase the secondary containment bypass leakage rate from O.OlLa to 0.042La or 04009 wt%/day to 0,028 wt%/day. The licensee performed a co4talnment radiological leakage analysis to provide the maximum value achievable for bypass leakage and found that the increased bypass leakage still meets the 10 CFR Part 100 dose limit. The staff concludes that the proposed bypass leakage rate is acceptable.

3.4 Electric Equipment Qualification for Service Conditions The current electric equipment qualification (EEQ) was based on a normal containment pressure range of 9.5 to 14,7 psia. The proposed containment operation pressure 14.2 psia falls within this range, and therefore, will not impact current EEQ. The licensee stated that the proposed increase in containment pressure would result in some increase in the radiation consequences following a DBA, but would not impact the existing accident radiation qualification of EEQ equipment. The staff confinmed the results of the radiation qualificatlon and found that the calculated maximum radiation

ASCII Text level was lower than the electric equipment tested values by more than 10%.

This provided an acceptable margin for the radiation qualification of EEQ equipment. Therefore, the staff concludes that the current EEQ is acceptable.

4.0 POST LOCA DOSE ASSESSMENT The original and current radiological consequence analyses were based on the sub-atmospheric design which terminates all primary containment leakage within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Consequently, the proposed chatge in the containment pressure in itself, without modifying any other requirements, would result in an increase In calculated offsite radiological consequences in an event of a LOCA.

Thtrefore, in order to compensate for the potential increase in the post-LOCA

.offsite doses, the licensee clalmed full credit for the iodine removal capabilities of the containment chemical spray in accordance with SRP Section 6.5.2, Revision 1. The licensee stated that such credit is not claited for the original and current LOCA analysis since the radiological consequences were acceptable without the spray. The staff found in the Millstone Unit 3 Safety Evaluation Report (NUREG-1031) dated July 1984 that the radiological consequences were also acceptable without the containment spray credit for iodine removal.

In addition, the licensee also proposed to change the allowable containment leak rates is follows:

Allowable Leak Rates (volume percent per d4z TS Sectioni-s 3.1.6.2 and 3.1.5.4j Primary Containment Leak Rate (La)

D'to 1.0 1 to 24 4 to 720 (hours) (hours) _(houjrs)

Current 0.9 0 0 Proposed 0.65 0.325 0.325 Bypa*s Leakage Current 0.009 0.009 0.009 Proposed 0,042 0.042 0.042 Uslin the.above proposed leak rates with a full credit allowed for iodine removal by the containment spray and the assumptions and parameters in Table 15.2 of Millstone Unit 3 SER, the staff computed the offslte doses for the

ASCII Text Millstone 3 Exclusiou Area (EAB) and Low Popaulaton Zone (LPZ) boundaries, The computed offslte doses are listed in Table 2, are within the acceptance Criteria given in Section 15.7.5 of the SRP and the exposure guidelines of 10 CFR Part 100 and are therefore acceptable.

TABLE 2 POST-OCA OFFSITE DOSES original() Revised(2) Limit(3)

Exclusion Area Boundary Thyroid 158 265 300 Whole Body 21 24 25 Low Population Zone Thyroid 8 180 300 Whole Body 1.1 5.6 25 I1i Table SER dated July 1984 Staff 15.1 of Millstone recalculated values3 10 CFR Part 100

5.0 PROPOSED CHANGE

S TO THE TS The licensee has proposed the following changes to the TS:

1. The peak calculated contaiwment pressure (P ) would be changed to 53.27 psia (38.57 psig) in Sections 4.6.1.1.c, 3,LI;2.a, 4.6.1.2.a, 4.6.1.2.,d 4.6.1.2.e, 3.6.1.3,b, 4.6.1.3.a.1 and a.2, 4.6.1.3.b.

2. The integrated liak rate at P containment leak rate (L. ) would be changed from 0.9 weight percent per day to 0,65 weight pgrcent per day in Section 3.6.1.2.a.

3. The combined bypass leakage rate would be changed from 0.01 La to 0.042 La in Sections 3.6.1.2 and 4.6.1.2.e.

ASCII Text

-B-

4. The operating containment pressure of 14.0 psia would be specified in Section 3,6.1.4. In addition, the maximum and minimum nlimit for the corntainmejit pressure would be specified as total containment pressure instead of air partial pressure.

5. Figure 3.6.1 would be deleted as the containment pressure will be read directly from the main control board indicators.

6. IS Table 3.6-1 would be changed as follows:

a. Penetrations 2-28 and Z-29 (aerated drains and gaseous vents) would be deleted,

b. Penetrations Z-59, Z-60, and Z-124 (fuel pool purification and nitrogen supply to containment) would be added.

c. Table 3.6.1 would be revised to include description for each penetration.

The proposed changes to the TS associated with the operating containment pressure and the associated peak calculated containment pressure (Pa),

coritainmunt leak rate (La) and bypass leakage rates are supported by the analysis presented in Section 3, herein. The results of the aralyses indicated that the potential post-LOCA off-site radiological consequences are within the limits of 10 CR Part 100. Accordingly, the proposed changes to the TS are acceptable.

With regard tD TS Table 3.6-I, "Enclosure Building Bypass Leakage Paths," the licensee has performed a review of the penetrations specified In this table whose combined leakage must be less than .01 La per TS 3.6.3.2. The licensee has determined that two penetrations, Nos. 28 and 29, do flat represent potential leakage paths. Since potential leakage would occur within the Auxiliary Duildings, for these penetrations, the liquid would be maintained within, the building while gaseous releases would be processed by the safety-grade ventilation systems. Accordingly, penetrations 28 and 29 should be deleted from TS Table 3.6-1. Conversely, the licensee has identified three punutrations, Nos. 59, 60 and 124, whose leakage could bypass the Enclosure Building and thus are appropriately added to TS Table 3.6-1. Finally, adding the proposed penetration descriptions to TS Table 3.6-1 does not effect either the associated Limiting Conditions for Operation or the Surveillance Hequirements and is,- thus, acceptable.

6,0 ENVIRONMENTAL CONSIDERATIONS Pursuant to 10 CFR 51.21 and 51.35, an environmental assessment and finding of no significant impact was prtpared and published in the Federal Register on December 20, 1990 (55 FR 52228). Accordingly, based upon the enviriimeiitaT assessment, we have determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.

ASCII Text

7.0 CONCLUSION

We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed mannerI and (2) such activities will be conducted ip compliance with the Commission s regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: January 25, 1991 Principal Contributors:

J. Guo D. Jaffe J. Lee

ASCII Textj 7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHEAST NUCLEAR EUERGY COMPANY DOCKET NO. 50-423 ROTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENJSE The U.S. Nuclear Regulatory Comnission (Commission) has issued Amendment No, 59 to Facility Operating License No. NPF-49 issued to Northeast Nuclear Energy Company, which revised the Technical Specifications for operation of the Mitlstone Nuclear Power Station, Unit No. 3 located in New London County, Connecticut. The amendment is effective as of the date of issuance.

The amendment modified the Technical Specifications to allow an increase in the normal containment pressure range. The revised containment pressure range is 10.6 psia to 14.0 psia.

The application for the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulatious. The Commission has made appropriate findings as required by the Act and the Commlssion's rules and regulations in 10 CFA Chapter 1, which are set forth in the liceuse amenduent.

Notice of Consideration of Issuance of Amendment and Opportunity for Hearing in connection with this action was published 4iithe FEDERAL REGISTER

ASCII Text on April 16& 1990 (55 FR 14149). No request for a hearing or petition for leave to intervene was filed following this notice, The Commission has prcpared an Environmental Assessment related to the action and has determined not to prepare an environmental impact statement.

Based upon the environmental assessment, the Commission has concluded that the Issuance of this amendment will not have a significant effect on the quality of the human environment.

Fur further details with respect to the action, see (1) the application for amendment dated February 26, 1990, as supplemented April 30, December 6 and 19, 1990, (2)Amendment No. 59 to License No. WPF-49, (3) the Commission's related Safety Evaluation, and (4) the Comission's Environmental Assessment.

All of these items are available for public inspection at the Commission's Public Document Room, the Gelman Building, 2120 L Street N.W., Washington, D.C.

and at the Learning Resources Center, Thames Valley State Technical College, 574 New London Turnpike, Norwich, Connecticut 06360. A copy of items (2), (3) and (4) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Director, Division of Reactor Projects - Ill.

Dated at Rockville, Maryland this 25th day of January 1991.

FOR TUE NUCLEAR REGULATORY COMMISSION Day il ff "p++djert Manager Project Directorate 1-4 Division of Reactor Projects - I/I1 Office of Nuclear Reactor Regulation