ML20204E741

Amend 228 to License DPR-65,allowing Implementation of Revised Main Steamline Break Analysis & Revised Radiological Consequences

ML20204E741
Person / Time
Site:	Millstone
Issue date:	03/10/1999
From:	Adensam E NRC (Affiliation Not Assigned)
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ML20204E747	List: ML20204E741 ML20204E758
References
NUDOCS 9903250118
Download: ML20204E741 (21)

Text

,

en acco y-q UNITED STATES g

g NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 2055fA001 4

i NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 228 License No. DPR-65

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

A. The applications for amendment by Northeast Nuclear Energy Company, et al. (the licensee) dated August 12,1998, as supplemented by letter dated October 30,1998, and September 28,1998, as supplemented by letters dated January 7 and 20,1999, comply 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 CFR Chapter I; B. The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) 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 Commission'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.

9903250118 990310 PDR ADOCK 05000336 p

PDR

2. Accordingly, changes to the Final Safety Analysis Report are authorized to reflect the revised main steamline break analysis and radiological consequences analyses as set forth in the licensee's application dated September 28,1998, as supplemented by letter dated -

January 20,1999. Additionally, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-65 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 228, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

{

Elinor G. Adensam, Director Project Directorate 1-2 Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: March 10, 1999

i ATTACHMENT TO LICENSE AMENDMENT NO. 228 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert 3/4 3-11 3/4 3-11 3/4 4-9 3/44-9 3/4 4-13 3/4 4-13 3/4 4-15 3/4 4-15 3/4 6-12 3/4 6-12 3/4 6-26 3/4 6-26 3/4 7-16 3/4 7-16 3/4 7-17 3/4 7-17 3/4 7-17a 3/4 7-17a 3/4 7-18 3/4 7-18 3/4 9-17 3/4 9-17 B 3/4 4-3 B 3/4 4-3 B 3/4 6-3 B 3/4 6-3 B 3/4 7-4 8 3/4 7-4 6-18 6-18 6-18a 6-18a 6-19 6-19 1

i a

INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued) i 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the "Tottl No. of Channels" Column of Table 3.3-3.

4.3.2.1.4 The trip value shall be such that the containment purge effluent shall not result in calculated ' concentrations of radioactivity offsite in excess of 10 CFR Part 20, Appendix B, Table 116 purposes of calculating this trip value, a x/Q = 5.8 x 10- sec/m shall beusedwhenthesystgmisa}ignedtopurgethroughthebuildingvent and a X/Q - 7.5 x 10-o sec/m shall be used when the system is aligned to purge through the Unit 1 stack, the gaseous and particulate (Half

. Lives greater than 8 days) radioactivity shall be assumed to be Xe-133 resp However, the setpoints shall be no greater and Cs-137$ cpm.ectively.

than 5 x10 MILLSTONE - UNIT 2 3/4 3-11 Amendment No. Q, 228 0409

o REACTOR COOLANT SYSTEN REACTOR COOLANT SYSTEN LEAKAGE LINITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE B0UNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

0.035 GPM primary-to-secondary leakage through any one steam generator, and d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System.

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

ACTION:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

b.

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, 1

reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

SURVEILLANCE REQUIRENENTS 4.4.6.2.1 Reactor Coolant System IDENTIFIED LEAKAGE and UNIDENTIFIED LEAKAGE shall be demonstrated to be within limits by performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation except when operating in the shutdown cooling mode.

4.4.6.2.2 Primary to secondary leakage shall be demonstrated to be within the above limits by performance of a primary to secondary leak rate determination at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The provisions of Specification 4.0.4 are not applicable for entry into MODE 4.

NILLSTONE - UNIT 2 3/4 4-9 Amendment No. 77, 77, pg, SJ.

osso 191,111,179,119, 228

REACTOR COOLANT SYSTEN j

SPECIFIC ACTIVITY LINITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a.

s 1.0 pCi/ gram DOSE EQUIVALENT I-131, and b.

s 100/E pCi/ gram of gross specific activity.

l APPLICABILITY: MODES 1, 2, 3, 4, and 5.

6_Gil0H:

MODES 1, 2, and 3*:

With the specific activity of the primary coolant > 1.0 pCi/ gram a.

DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Specification 3.0.4 is not applicable.

b.

With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T,y, < 515'F within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, c.

With the specific activity of the primary coolant > 100/E pCi/ gram of gross specific activity, be in HOT STANDBY with T,y, < 515'F l within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

MODES 1, 2, 3, 4 and 5:

d.

With the specific activity of the primary coolant > 1.0 pCi/ gram DOSE EQUIVALENT I-131 or > 100/E Ci/ gram of gross specific activity, perform the sampling and analysis requirements of item 4 a) of Table 4.4-2 until the specific activity of the primary coolant is restored to within its limits.

• With T,,, > 515'F.

NILLSTONE - UNIT 2 3/4 4-13 Amendment No. 7. ///, J/J, JJJ, osts 199, 228

IABLE 4.4-2 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS l

AND ANALYSIS FRE0VENCY 1.

Gross Activity Determination 3 times per 7 days with a maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples 2.

Isotopic Analysis for DOSE 1 per 14 days 1

EQUIVALENT I-131 Concentration 3.

Radiochemical Analysis for 1 per 6 months

• l l

E Determination 4.

Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Including I-131, I-133, and I-135.

whenever the specific activity exceeds 1.0 l

pCi/ gram, DOSE EQUIVALENT l

I-131, or 100/E pCi/ gram of gross specific I

activity, and b) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

• Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer. The provisions of Specification 4.0.4 are not applicable.

MILLSTONE - UNIT 2 3/4 4-15 Amendment No. 228 0412

' CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS l

1 CONTAINMENT SPRAY AND COOLING SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.2.1 Two containment spray trains and two containment cooling trains, with each cooling train consisting of two containment air recirculation and cooling units, shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3*.

ACTION:

Inoperable Equipment Required Action a.

One containment a.1 Restore the inoperable containment spray spray train train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or l

be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

One containment b.1 Restore the inoperable containment cooling cooling train train to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c.

One containment c.1 Restore the inoperable containment spray spray train train or the inoperable containment cooling AND train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or One containment be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

cooling train d.

Two containment d.1 Restore at least one inoperable containment cooling trains cooling train to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

e.

All other e.1 Enter LC0 3.0.3 immediately.

combinations SURVEILLANCE REQUIREMENTS 4.6.2.1.1 Each containment spray train shall be demonstrated OPERABLE:

At least once per 31 days on a STAGGERED TEST BASIS by:

a.

1.

Starting each spray pump from the control room, 2.

Verifying, that on recirculation flow, each spray pump develops a discharge pressure of 2 254 psig,

• The Containment Spray System is not required to be OPERABLE in MODE 3 if pressurizer pressure is < 1750 psia.

MILLSTONE - UNIT 2 3/4 6-12 Amendment No. 7/J,228 0413

x CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a.

C.5.c and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train flow rate is 9000 cfm 10%.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

3.

Verifying a train flow rate of 9000 cfm 10% during train operation when tested in accordance with ANSI N510-1975.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying c.

within 31 days after removal that a laboratory analysis of a representa-tive carbon sample obtained in accordance with-Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the labcratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is s 2.6 inches Water Gauge while l

operating the train at a flow rate of 9000 cfm 10%.

2.

Verifying that the train starts on an Enclosure Building Filtra-tion Actuation Signal (EBFAS).

After each complete or partial replacement of a HEPA filter bank by e.

verifying that the HEPA filter banks remove greater than or equal to 99% of the D0P when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm 10%.

ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52.

The laboratory test of charcoal should be conducted at a temperature of 30*C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

Additionally, the charcoal sample shall have a removal efficiency of ;t 95%.

MIgLSTONE-UNIT 2 3/4 6-26 Amendment No. /Ji, //, RJ, 7#,228

l 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 Two independent Control Room Emergency Ventilation Trains shall be l

OPERABLE.

APPLICABILITY: ALL MODES ACTION:

Modes 1, 2, 3, and 4:

With one Control Room Emergency Ventilation Train inoperable, restore the inoperable train to OPERABLE status within 7 days 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 />.

MODES 5 and 6*

With one Control Room Emergency Ventilation Train inoperable, a.

restore the inoperable train to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Emergency Ventilation Train in the recirculation mode.

l b.

With both Control Room Emergency Ventilation Trains inoperable, or with the OPERABLE Control Room Emergency Ventilation Train required to be in the recirculation mode by ACTION (a.

powered by an OPERABLE normal and emergency po)wer source, suspendno all operations involving CORE ALTERATIONS or positive reactivity changes.

In Modes 5 and 6, when a Control Room Emergency Ventilation Train is l

determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of 3.7.6.1 Limiting Condition for Operation, provided: (1) its corresponding normal Dr emergency power source is OPERABLE; and (2) all of its redundan.t system (s), subsystem (s), train (s), component (s) and device (s) are OPERABLE, or likewise satisfy the requirements of the specification. Unless both conditions (1) and (2) are satisfied within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then Limiting Condition for Operation (LCO) 3.7.6.1.a or 3.7.6.1.b shall be invoked as applicable.

,MIgLSTONE-UNIT 2 3/4 7-16 Amendment No. 77, J/p, Jff 228

^

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS l

=

4.7.6.1 Each Control Room Emergency Ventilation Train shall be demonstrated l OPERABLE:

a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the control room air temperature is s 100*F.

b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating from the control room, flow through the HEPA filters and charcoal at>sorber train and verifying that the train operates for at leastl 15 minutes.

c.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any ventilation zone communi-l cating with the train by:

l 1.

Verifying that the cleanup train satisfies the in-place l testing acceptance criteria and uses the test procedures of Regulatory Positions C.5.a, C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the train flow rate is 2500l cfm 10%.

2.

Verifying within 31 days after removal that a

laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978.*

The carbon sample shall have a removal efficiency of 2 95 percent.

3.

Verifying a train flow rate of 2500 cfm i 10% during train l operation when tested in accordance with ANSI N510-1975.

d.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52.

The laboratory test of charcoal should be conducted at a temperature of 30*C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

MILLSTONE - UNIT 2 3/4 7-17 Amendment No. EJ, /g, Jpp, JJJ, ws ill 199, Ill,228

e e

PLANT '7 STEMS SURVEILLANCE REQUIREMENTS (Continued)

At least once per 18 months by:

e.

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 3.4 inches Water Gauge while operating the train at a flow rate of 2500 cfm 10%.

2.

Verifying that on a recirculation signal, with the Control Room Emergency Ventilation Train operating in the normal mode and the smoke purge mode, the train automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks.

I NILLSTONE - UNIT 2 3/4 7-17a Amendment No. 7J, 77, Jpp, JJp, sus l1F, 199 171.

,,g x

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 3.

Verifying that control room air in-leakage is less than 130 SCFM with the Control Room Emergency Ventilation System 1 operating in the recirculation / filtration mode.

f.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal +

99% of the D0P when they are tested in-place iri accordance with ANi!

N510-1975 while operating the train at a flow rate of 2500 cfm 10%.

I g.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 2500 cfm 10%.

1 I

NIgLSTONE-UNIT 2 3/4 7-18 Amendment No. 7/, J#, J#,228

f REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1.

Verifying that the cleanup train satisfies the in-place testing acceptance criteria and uses the test procedures of Regulatory Positions C.S.a, C.S.c and C.5.d of Regulatory Guide 1.52, Revi-sion 2, March 1978, and the train flow rate is 9000 cfm 10%.

2.

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

3.

Verifying a train flow rate of 9000 cfm 10% during train operation when tested in accordance with ANSI N510-1975.

c.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratoiy analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978.*

d.

At least once per 18 months by:

1.

Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is 12.6 inches Water Gauge l

while operating the train at a flow rate of 9000 cfm i 10%.

2.

Verifying that on a Spent Fuel Storage Pool Area high radiation signal, the train automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks.

e.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the D0P when they are tested in-place in accordance with ANSI N510-1975 while operating the train at a flow rate of 9000 cfm i 10%.

1 j

ASTM D3803-89 shall be used in place of ANSI N509-1976 as referenced in table 2 of Regulatory Guide 1.52.

The laboratory test of charcoal should i

be conducted at a temperature of 30*C and a relative humidity of 95% within the tolerances specified by ASTM D3803-89.

Additionally, the charcoal sample shall have a removal efficiency of 2 95%.

MILLSTONE - UNIT 2 3/4 9-17 Amendment No. 77, 77), 7 #,

228

REACTOR COOLANT SYSIEj!

BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE

{

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary leakage Detection Systems."

3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation prwides allowance for a limited amount of leakage from known sources whose preserce will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakige detection systems.

The steam generator tube leakage limit of 0.035 GPM per steam generator ensures that the dosage contribution from the tube leakage will be less than the limits of General Design Criteria 19 of 10CFR50 Appendix A in the event of either a steam generator tube rupture or steam line break.

The 0.035 GPM limit is consistent with the assumptions used in the analysis of these accidents.

PRESSURE B0UNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Therefore, the preaence of any PRESSURE B0UNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

MI LSTONE - UNIT 2 B 3/4 4-3 Anendment Nos. JJJ, J77,228

I a

CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY AND COOLING SYSTEMS The OPERABILITY of the containment spray system ensures that contain-ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses. The leak rate surveillance requirements assure that the leakage assumed for the system outside containment during the recircula-tion phase will not be exceeded.

The OPERABILITY of the containment cooling system ensures that

1) the containment air temperature will be maintained within limits during normal operation, and 2) adequate heat removal capacity is available when i

operated in conjunction with the containment spray system during post-LOCA conditions.

]

1 To be OPERABLE, the two trains of the containment spray system shall be

)

capable of taking a suction from the refueling water storage tank on a containment spray actuation signal and automatically' transferring suction to the containment sump on a sump recirculation actuation signal. Each containment spray train flow path from the containment sump shall be via an OPERABLE' shutdown cooling heat exchanger.

The containment cooling system consists of two containment cooling trains.

Each containment cooling train has two containment air recirculation and cooling units.

For the purpose of applying the appropriate action l

statement, the loss of a single containment air recirculation and cooling unit will make the respective containment' cooling train inoperable.

Either the containment spray system or the containment cooling system has sufficient he.t removal capability to handle any design basis accident.

However, the cor.tainment spray system is more effective in dealing with the superheated steam from a main steam break inside containment.

In addition, the containmant spray system provides a mechanism for removing iodine from the containment atmosphere. Therefore, at least one train of containment spray is 1

required to be OPERABLE when pressurizer pressure is p_ 1750 psia, and the 3110wed outage time for one. train of containme'nt spray reflects the dual function of containment spray for heat removal and iodine removal.

3/4.6.3 CONTAINMENT ISOLATION VALVES The Technical Requirements Manual contains the list of containment isolation valves (except the containment air lock and equipment hatch). Any changes to this list will be reviewed under 10CFR50.59 and approved by the Plant Operations Review Committee (PORC).

The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmos-phere or pressurization of the containment. Containment isolation within MILLSTONE - UNIT 2 B 2/4 6-3 Amendment No. 7), 77, JJP, 77) ous 228

r

• PLANT SYSTEMS RARN 3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident con-ditions. The redundant moling capacity of this system, assuming a single failure, is com a t with the assumptions used in the accident analyses.

3/4.7.5 FLOOD LEVEL The service water pump motors are normally protected against water damage to an elevation of 22 feet.

If the water level is exceeding plant grade level or if a severe storm is approaching the plant site, one service water pump motor will be protected against flooding to a minimum elevation of 28 feet to ensure that this pump will continue to be capable of removing decay heat from the reactor.

In order to ensure operator acco:ssibility to the intake structure action to provide pump motor protection will be initiated when the water level reaches plant grade level.

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the Control Room Emergency Ventilation System l

ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions.

The OPERABILITY of this system in conjunction with control room design provisions is based on limitir.g the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.

This limitation is consistent wit the requirements of General Design Criteria 19 of Appendix "A",

10 CFR 50.

l

~

The control room radiological dose calculations use the conservative minimum acceptable flow of 2250 cfm based on the flowrate surveillance requirement of 2500 cfm i 10%.

Currently there are some situations where the CREV System may not automatically start on an accident signal, without operator action. Under mo;t situations, the emergency filtration fans will start and the CREV System will be in the accident lineup. However, a failure of a supply fan (F21A or B) or an exhaust fan (F31A or B), operator action will be required to return to a full train lineup. Also, if a single emergency bus does not power up for one train of the CREV System, the opposite train filter fan will automatically start, but the required supply and exhaust fans will not automatically start.

Therefore, operator action is required to establish the whole train lineup.

This action is specified in the Emergency Operating Procedures. The radiological dr.o calculations do not take credit for CREV System cleanup l

action until 10 minutes into the accident to allow for operator action.

When the CREV System is checked to shift to the recirculation mode of operation, this will be performed from the normal mode of operation, and from the smoke purge mode of operation.

l HILLSTONE - UNIT 2 B 3/4 7-4 Amendment No.228 0419

ADMINISTRATIVE CONTROLS d.

Documentation of all failures (inability to lift or reclose within the tolerances allowed by the design basis) and challenges to the i

pressurizer PORVs or safety valves.

ANNUAL RADI0 ACTIVE EFFLUENT REPORT 6.9.1.6 A routine Annual Radioactive Effluent Report covering the operation of the unit during the previous calendar year of operation shall be submitted by May 1 of each year.

The report shall include that information delineated in the REM 0DCM.

Any changes to the REM 0DCM shall be submitted in the Annual Radioactive Effluent Report.

MONTHLY OPERATING REPORT 6.9.1.7 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S.

Nuclear Reguhtory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector, no later than the 15th of each mon 3 following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT 6.9.1.8 a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

3/4.1.1.1 SHUTDOWN MARGIN - T

> 200*F 3/4.1.1.2 SHUTDOWN MARGIN - T " 1 200*F 3/4.1.1.4 ModeratorTemperatuEECoefficient 3/4.1.3.6 Regulating CEA Insertion Limits 3/4.2.1 Linear Heat Rate 3/4.2.3 TotalIntegratedRadialPeakingFactor-F}

3/4.2.6 DNB Margin b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the

)

NRC, specifically those described in the following documents:

j 1)

XN-75-27(A) and Supplements 1 through 5,

" Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors,"

Exxon Nuclear Company, Report and Supplement I dated April 1977, Supplement 2 dated Decembu 1980, Supplement 3 dated September 1981 (P), Supplement 4 dated December '986 (P), and Supplement 5 dated February 1987 (P).

l l

2)

ANF-84-73 Revision 5 Appendix B (P)(A), " Advanced Nuclear Fuels Methodology for Pressurized Water Reactors:

Analysis of Chapter 15 Events," Advanced Nuclear Fuels, July 1990.

1 MILLSTONE - UNIT 2 6-18 Amendment Nos. JJ, JJ, J/J JJJ, 0407

((f,//f,((%,

117,199,191,199, 228

)

4

'. ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (CONT.)

3)

XN-NF-82-21(P)(A) Revision 1, " Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company, September 1983.

4)

EMF-84-93(P) Revision 1 "Steamline Break Methodology for PWRs," Siemens Power Corporation, June 1998.

5)

XN-75-32(P)(A)

Supplements 1

through 4,

" Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company, October 1983.

6)

XN-NF-82-49(P)(A)

Revision 1,

" EXXON Nuclear Company Evaluation Model EXEM PWR Small Break Model," Advanced Nuclear Fuels Corporation, April 1989.

7)

XN-NF-82-49(P)(A) Revision 1 Supplement 1,

" Exxon Nuclear Company Evaluation Model Revised EXEM PWR Small Break Model,"

Siemens Power Corporation, December 1994.

8)

EXEM PWR Large Break LOCA Evaluation Model as defined by:

XN-NF-82-20(P)(A)

Revision 1

Supplement 2,

" Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company, February 1985.

XN-NF-82-20(P)(A) Revision 1 and Supplement 1, 3, and 4,

" Exxon Nuclear Compan / Evaluation Model EXEH/PWR ECCS Model Updates," Advanced Nuclear Fuels Corporation January 1990.

XN-NF-82-07(P)(A) Revision 1,

" Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company, November 1982.

XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.

ANF-81-58(P)(A) Revision 2 Supplements 3 and 4, "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model,"

Advanced Nuclear Fuels Corporation, June 1990.

XN-NF-85-16(P)(A) Volume 1 and Supplements 1, 2, and 3; Volume 2, Revision 1 and. Supplement 1, "PWR 17 x 17 Fuel Cooling Test Program,'

Advanced Nuclear Fuels Corporation, February 1990.

XN-NF-85-105(P)(A) and Supplement 1, " Scaling of FCTF Based Reflood Heat Transfer Correlation for Other Bundle Designs," Advanced Nuclear Fuels Corporation, January 1990.

MILLSTONE - UNIT 2 6-18a Amendment No. 79, 97, J/9, JJJ.

o*ol 119,119,179, 117, J99, 191 199, 228

ADMINISTRATIVE CONTROLS 1

1 l

CORE OPERATING LIMITS REPORT (CONT.)

l 9)

XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized water reactors," Exxon Nuclear Company, October 1983.

10) XN-NF-621(P)(A) Revision 1,

" Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company, September 1983.

I

11) XN-NF-82-06(P)(A) Revision 1 and Supplements 2, 4,

and 5,

" Qualification of Exxon Nuclear Fuel for Extended 8urnup,"

Exxon Nuclear Company, October 1986.

12) ANF-88-133(P)(A) and Supplement 1, " Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTV," Advanced Nuclear Fuels Corporation, December 1991.

13) XN-NF-85-M p)(A), " Exxon Nuclear Uranium Dioxide /Gadolinia Irradiation Examination and Thermal Conductivity Results,"

Exxon Nuclear Company, November 1988.

14) ANF-89-151(P)(A),

"ANF-RELAP Methodology for Pressurized Water Reactors:

Analysis of Non-LOCA Chapter 15 Events,"

Advanced Nuclear Fuels Corporation, May 1992.

15) XN-NF-507(P)(A)

Supplements and 2,

" ENC Setpoint Methodology for C.E.

Reactors:

Statistical Setpoint Methodology," Exxon Nuclear Company, September 1986.

c.

The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory l

Commission, Document Control Desk, Washington, D.C.

20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector within the time period specified for i

l each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

Inoperable Seismic Monitoring Instrumentation, Specification a.

3.3.3.3.

l MILLSTONE - UNIT 2 6-19 Amendment No. JJ7, 177,228 0407 l