Text

Forwards Page Changes to Final Draft Tech Specs. Certification That Revised Tech Specs Consistent W/Fsar,Ser & Ssers Requested by 860630
	ML20199K047
Person / Time
Site:	Seabrook
Issue date:	06/26/1986
From:	Novak T; Office of Nuclear Reactor Regulation
To:	Harrison R; PUBLIC SERVICE CO. OF NEW HAMPSHIRE
References
	NUDOCS 8607090023
	Download: ML20199K047 (120)
	v • d • e

(d Docket No.: 50-443 Mr. Robert J. Harrison President and Chief Executive Officer JM 2 6 E Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105'

Dear Mr. Harrison:

SUBJECT:

CHANGES TO THE SEABROOK FINAL DRAFT TECHNICAL SPECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit 1 Final Draft Technical Specifications (TS) were transmitted to you for certification under oath and affinnation that the Final Draft TS are consistent with the Final Safety Analysis Report (FSAR), the Safety Evaluation Report (SER), and the .

as-built facility.

In a Public Service of New Hampshire letter dated June 20, 1986, it was noted that certification was not possible until certain inconsistencies were eliminated. After a number of discussions between our staffs to eliminate these inconsistencies a number of page changes to the Final Draft TS were made.

Enclosed are these page changes to the Final Draft Seabrook Station, Unit 1 TS. These changes are the results of the discussions between our staffs, and they are being forwarded to you for incorporation into the Final Draft TS transmitted to you on June 18, 1986.

You are requested to certify, under oath and affirmation, that the revised Final Draft TS are consistent with the FSAR, the SER and its supplements and the as-built facility. Your certification is requested by June 30, 1986.

For further information or clarification, please contact the Licensing Project Manager, Victor Nerses at (301) 492-8535.

Thomas M. Novak, Acting Director Division of PWR Licensing-A

Enclosure:

As stated cc: See next page 8607090023 e60626 PDR ADOCK 05000443 A PDR DISTRIBUTION:

Docket Files T. Novak J. Partlow V. Nerses NRC PDR PD#5 R/F E. Jordan M. Rushbrook Local PDR OELD B. Grimes ACRS (10)

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JLM(261986 Docket No.: 50-443 Mr. Robert J. Harrison President and Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105

Dear Mr. Harrison:

SUBJECT:

CHANGES TO THE SEABROOK FINAL DRAFT TECHNICAL SPECIFICATIONS In a letter dated June 18, 1986, the Seabrook Unit 1 Final Draft Technical Specifications (TS) were transmitted to you for certification, under oath and affirmation that the Final Draft TS are consistent with the Final Safety Analysis Report (FSAR), the Safety Evaluation Report (SER), and the as-built facility.

In a Public Service of New Hampshire letter dated June 20, 1986, it was noted that certification was not possible until certain inconsistencies were eliminated. After a number of discussions between our staffs to eliminate these inconsistencies a number of page changes to the Final Draft TS were made.

Enclosed are these page changes to the Final Draft Seabrook Station, Unit 1 TS. These changes are the results of the discussions between our staffs, and they are being forwarded to you for incorporation into the Final Draft TS transmitted to you on June 18, 1986.

You are requested to certify, under oath and affirmation, that the revised Final Draft TS are consistent with the FSAR, the SER and its supplements and the as-built facility. Your certification is requested by June 30, 1986.

For further information or clarification, please contact thp Licensing Pro,iect Manager, Victor Nerses at (301) 492- 35 f

.b NNo . Director Division of PWR L n ng-A

Enclosure:

As stated cc: See next page

k e Mr. Robert J. Harrison Public Service Company of New Hampshire Seabrook Nuclear Power Station cc:

Thomas Dignan, Esq. E. Tupper Kinder, Esq.

John A. Ritscher, Esq. G. Dana Bisbee, Esq.

Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Hosue Annex Concord, New Hampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Sumer Street Mr. John DeVincentis, Director Haverhill, Massachusetts 01839 Engineering and Licensing Yankee Atomic Electric Company Robert A. Backus, Esq. 1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 Lowell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manager United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8223 Harmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW Suite 430 Washington, D.C. 20009 Mr. Philip Ahrens, Esq.

Assistant Attorney General State House, Station #6 Augusta, Maine 04333 Jo Ann Shotwell, Esq.

Office of the Assistant Attorney General Environmental Protection Division Mr. Warren Hall One Ashburton Place Public Service Company of Boston, Massachusetts 02108 New Hampshire ,

Post Office Box 330 D. Pierre G. Cameron, Jr., Esq.

Seabrook, New Hampshire 03874 General Counsel Public Service Company of New Hampshire Seacoast Anti-Pollution League Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 5 Market Street Portsmouth, New Hampshire 03801 Regional Administrator, Region I U.S. Nuclear Regulatory Commission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street King of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 Richard Hampe, Esq.

New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301 L

i .

Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:

Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent, City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. Letty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C. 20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin, Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Town of Durham Hampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq.

Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue -

Post Office Box 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chichester, Chaiman Rye Nuclear Intervention Mr. Angie Machiras, Chairman Committee Board of Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road Newbury, Massachusetts 01950 Rye, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Commission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C. 20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter 10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874

f INDEX , BNALHAFT BASES SECTION PAGE 3/4.10.4 REACTOR COOLANT L00PS..................................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN..................... B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS........................................ B 3/4 11-1 3/4.11.2 GASE0US EFFLUENTS....................................... B 3/4 11-3 3/4.11.3 SOLID RADI0 ACTIVE WASTES................................ B 3/4 11-6 3/4.11.4 TOTAL 00SE.............................................. B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING ,

3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE CENSUS......................................... B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARIS0N PR0 GRAM...................... B 3/4 12-2 5.0 DESIGN FEATURES 5.1 SITE 5.1.1 EXCLUSION AREA.............................................. 5-1 5.1.2 LOW POPULATION Z0NE......................................... 5-1 5.1.3 MAPS DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASE0US AND LIQUID EFFLUENTS.................... 5-1 FIGURE 5.1-1 EXCLUSION AREA....................................... 5-2 FIGURE 5.1-2 LOW POPULATION Z0NE.................................. 5-3 FIGURE 5.1-3 SITE BOUNDARY FOR LIQUID AND GASE0US EFFLUENTS....... 5-4 5.2 CONTAINMENT 5.2.1 CONFIGURATION............................................... 5-1 5.2.2 DESIGN PRESSURE AND TEMPERATURE............................. 5-1 5.3 REACTOR CORE 5.3.1 FUEL ASSEMBLIES............................................. 5-5 5.3.2 CONTROL R0D ASSEMBLIES...................................... 5-5 5.4 REACTOR COOLANT SYSTEM 5.4.1 DESIGN PRESSURE AND TEMPERATURE............................. 5-5 5.4.2 V0LUME...................................................... 5-5 5.5 METEOROLOGIC AL TOWER LOCATI0N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-5 SEABROOK - UNIT 1 xii M 2 51986

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.

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 Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

. DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in NRC Regulatory Guide 1.109, Revision 1, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample with half-lives greater than 10 minutes.

SEABROOK - UNIT 1 1-2 O 2 5 1986

DEFINITIONS ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint l 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.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.14 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 B0UNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.15 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MEMBER (S) 0F THE PUBLIC 1.16 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Alsc excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recre-ational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain in Part A the radiological effluent sampling and analysis program and radiological environ-mental monitoring program. Part B of the ODCM shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

SEABROOK - UNIT 1 1-3 JUN 25 96

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DEFINITIONS FINAL DRAF QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.

CONTAINMENT ENCLOSURE BUILDING INTEGRITY 1.28 CONTAINMENT ENCLOSURE BUILDING INTEGRITY shall exist when:

a. Each door in each access opening is closed except when the access opening is being used for normal transit entry and exit,

b. The Containment Enclosure Filtration System is OPERABLE, and

c. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

SEABROOK - UNIT 1 1-5 l JilN 2 5 W

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS FINAL DSFT 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure 2.1-1 for four-loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.6.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.6.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.6.

SEABROOK - UNIT 1 2-1 JUN 2 51986 A

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REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SEABROOK - UNIT 1 2-2 EN 2 f 1986

a TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 SENSOR E TOTAL ERROR e

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE E 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

~ 2. Power Range, Neutron Flux

a. High Setpoint 7.5 4.56 0 $109% of RTP* $111.1% of RTP*

b. Low Setpoint 8.3 '6

, 0 $25% of RTP* $27.1% of RTP*

3. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with High Positive Rate i time constant i time constant 12 seconds 12 seconds

4. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with

, High Negative Rate i time constant i time constant 12 seconds 32 seconds

5. Intermediate Range, 17.0 8.41 0 $25% of RTP* $31.1% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.6 x 105 cps _

7. Overtemperature AT 6.5 3.31 1.04** See Note 1 See Note 2

+0.47**

8. Overpower AT 4.8 1.43 0.12 See Note 3 See Note 4 '"T1 imumme

9. Pressurizer Pressure - Low 3.1 0.71 1.69 11945 psig 11,935 psig Z

10. Pressurizer Pressure - High 3.1 0.71 1.69 52385 psig $2,395 psig My c

RTP = RATED THERMAL POWER i ("

- The sensor error for T avg is 1.04 and the sensor error for Pressurizer Pressure is 0.47. "As measured"

$ sensor errors may be used in lieu of either or both of these values, which then must be summed to deter-mine S overtemperature AT total channel value for S.

1

o a LIMITING SAFETY SYSTEM SETTINGS aa BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip, thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure that could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power, the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, the Pressurizer High Water Level trip is automatically reinstated by P-7.

Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 50% of RATED THERMAL POWER), an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely, on decreasing power between P-8 and the P-7, an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specified Setpoint provides allowances for starting delays of the Emergency Feedwater System.

SEABROOK - UNIT 1 B 2-6 JUN 2 51986

, e 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T avg GREATER THAN 200 F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 3.8% Ak/k for four-loop operation.

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

ACTION:

With the SHUTDOWN MARGIN less than 3.8% Ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 3.8% Ak/k:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b. When in MODE 1 or MODE 2 with k,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6;

c. When in MODE 2 with k,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6;

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

SEABROOK - UNIT 1 3/4 1-1 JUN 2 51986

REACTIVITY CONTROL SYSTEMS B0 RATION CONTROL SHUTDOWN MARGIN - T 3yg LESS THAN OR EQUAL TO 200 F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.2% Ak/k.

Additionally, the Reactor Coolant System boron concentration shall be greater than or equal to 2000 ppm boron when the reactor coolant loops are in a drained condition.

APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MAkGIN less than 1.2%* Ak/k or the Reactor Coolant System boron concentration less than 2000 ppm boron, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN and boron concentration are restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN and boron concentration shall be determined to be greater than or equal to 1.2%* Ak/k:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

SEABROOK - UNIT 1 3/4 1-3 JUN 2 51986

i e REACTIVITY CONTROL SYSTEMS B0 RATION SYSTEMS CHARGING PUMP - SHUTOOWN LIMITING CONDITION FOR OPERATION i

3.1.2.3 One charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 4, 5, and 6.

ACTION:

With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across the pump of greater than or equal to 2480 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.3.2 All charging pumps, excluding the above required OPERABLE pump, shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold legs decreas-ing below 325 F, whichever comes first, and at least once per 31 days there-after, except when the reactor vessel head is removed.

An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

SEABROOK - UNIT 1 3/4 1-9 JtJN 2 5 96 1

t

REACTIVITY CONTROL SYSTEMS BORATION SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.*

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1.3% Ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, on recirculation flow, that a differential pressure across each pump of greater than or equal to 2480 psid is developed when tested pursuant to Specification 4.0.5.

The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the centrifugal charging pump declared inoperable pursuant to Specification 4.1.2.3.2 provided that the centrifugal charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first.

SEABROOK - UNIT 1 3/4 1-10 JUN 25 nas 1

' S C i' ,ru REACTIVITY CONTROL SYSTEMS ,

Ilt BORON SYSTEMS ISOLATION OF UNB0 RATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.3.2.7 Previsions to isolate the Reactor Coolant System from unborated water sources shall be OPERABLE with:

a. The Boron Thermal Regeneration System (BTRS) isolated from the Reactor Coolant System, and

b. The Reactor Makeup Systems inoperable except for the capability of delivering up to the capacity of one Reactor Makeup Water pump to e the Reactor Coolant System.

APPLICABILITY: MODES 4, 5, and 6 ACTION:

With the requirements of the above specification not satisfied immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and, if within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the required SHUTDOWN MARGIN is not verified, initiate and continue boration at grester than or equal to 30 gpm of a solu-tion containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored and the isolation provisions are restored to OPERABLE.

SURVEILLANCE REQUIREMENTS 4.1.2.7 The provisions to isolate the Reactor Coolant System from unborated water sources shall be determined to be OPERABLE at least once per 31 days by:

a. Verifying that at least the BTRS outlet valve, CS-V-302, or the BTRS moderating heat exchanger outlet valve, CS-V-305, is closed and locked closed, and

b. Verifying that power is removed from at least one of the Reactor Makeup Water pumps, RMW-P-16A or RMW-P-168.

SEABROOK - UNIT 1 3/4 1-14 JUN 2 5 M6

FINAL DRAFT (0.30,228) (0.844,228) 228

/ /

2,, / BANK B ,/

2 3 /

[ /

/

$ / /

@ /(0.0,164) /

g160 7 3

u, 1

/ (1.0,146)-

J e / BANK C /

U) 120 '

~

,/ s/

5

_ / /

b / /

o ** / /

~

/ /

E / /

de.e,49) / BANK D 40 -

x /

/

/ (03100)

/ I i l l 0.0 0.2 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR-LOOP OPERATION SEABROOK - UNIT 1 3/4 1-23 JUN 2 51986

I J 0 FINAL DRAFT 120

$<re eB ese N

100 U,NACCEPIABLE OPERATION UNACCEPTABLE OPERATION

! I-11,'9 0) (11,'90) 5 / \

3 / h\ j

'Oag i

/ \

d / \

hW

[ ACCEPTABLE OPERATION \

1 l

/ )'

~ 60

/ \

o / )

w i t h

m

-31,50) (31,5 0) g 40 a 1 N

20 ,

0

-50 -40 -30 -20 -10 0 10 20 30 40 50 FLUX DIFFERENCE ( AI)%

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER SEABROOK - UNIT 1 3/4 2-3 JUN 2 5 W

FINAL DRA 1.2 (6.0,1.0) 0 0.9.EM4)

T

\

_ e.e \

U L E \

o T y e.6 uz.e.e ss)

<I I

z i e.4 S

x e.2 i

s.e e 2 4 6 8 le 12 CORE HEIGHT (FT)

FIGURE 3.2-2 K(Z) - NORMALIZED F q(Z) AS A FUNCTION OF CORE HEIGHT SEABROOK - UNIT 1 3/4 2-5 JUN 2 5115

a TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION 8

g MINIMUM

, TOTAL NO. CHANNELS CHANNELS APPLICABLE e FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i'i H 1. Manual Reactor Trip 2 1 2 1, 2 1 H 2 1 2 3*, 4*, 5* 10

2. Power Range, Neutron Flux

a. High Setpoint 4 2 3 1, 2 2#

b. Low Setpoint 4 2 3 1###, 2 2#

3. Power Range, Neutron Flux 4 2 3 1, 2 2#

High Positive Rate

4. Power Range, Neutron Flux, 4 2 3 1, 2 2#

High Negative Rate

5. Intermediate Range, Neutron Flux 2 1 2 1###, 2 3 ca E 6. Source Range, Neutron Flux

a. Startup 2 1 2 2## 4

b. Shutdown 2 0 1 3,4,5 5

c. Shutdown 2 1 2 3*, 4*, 5* 10

7. Overtemperature AT 4 2 3 1, 2 6#

8. Overpower AT 4 2 3 1, 2 6#

9. Pressurizer Pressure--Low 4 2 3 1** 6# (1) 2:

10. Pressurizer Pressure--High 4 2 3 1, 2 6# (1) P

11. Pressurizer Water Level--High 3 2 2 1** 7# Q

.w m

$ M

l <

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

' 8 o TRIP 7 ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH E CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

1. Manual Reactor Trip N.A. N.A. N.A. R(13) N.A. 1, 2, 3* , 4* , @

2. Power Range, Neutron Flux

a. High Setpoint S D(2,4), Q(16) N.A. N.A. 1, 2 M(3, 4),

Q(4, 6),

R(4,5)

b. Low Setpoint S R(4) S/U(1) N.A. N.A. 1***, 2

3. Power Range, Neutron Flux, N.A. R(4) Q(16) N.A. N.A. 1, 2 g High Positive Rate s

y 4. Power Range, Neutron Flux, N.A. R(4) Q(16) N.A. N.A. 1, 2 m High Negative Rate

5. Intermediate Range, S R(4,5) S/U(1) N.A. N.A. 1***, 2 Neutron Flux

6. Source Range, Neutron Flux S R(4, 5) S/U(1),Q(9,16) N.A. N.A. 2**, 3, 4, 5

7. Overtemperature AT S R(12) Q(16) N.A. N.A. 1, 2 ==T"1 ammum

8. Overpower AT S R Q(16) N.A. N.A. 1, 2 s%

9. Pressurizer Pressure--Low S R Q(16,17) N.A. N.A. 1 y

10. Pressurizer Pressure--High S R Q(16,17) N.A. N.A. 1, 2 g

11. Pressurizer Water Level--High S R Q(16) N.A. N.A. 1

$ 12. Reactor Coolant Flow--Low S R Q(16) N.A. N.A. 1 ="T"1

$ M

a TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5

g TRIP ANALOG ACTUATING MODES FOR c CHANNEL DEVICE WHICH H

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST s IS REQUIRED Reactor Trip System Interlocks (Continued)

e. Power Range Neutron Flux, P-10 N.A. R(4) R N.A. N.A. 1, 2

f. Turbine Impulse Chamber Pressure, P-13 N.A. R R N.A. N.A. 1

19. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, 2, 3* ,

4*, 5*

20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) 1, 2, 3*,

y Logic 4*, 5*

U

21. Reactor Trip Bypass Breaker N.A. N. A. N.A. M(14),R(15) N.A. 1, 2, 3* ,

4*, 5*

==

c r-n en w

< l TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E

O MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

--e w 1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, and CBA Emergency Fan /

Filter Actuation).

a. Manual Initiation 2 1 2 1,2,3,4 17 w

0

b. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays

c. Containment 3 2 2 1,2,3 14*

Pressure--Hi-1

d. Pressurizer 4 2 3 1, 2, 3# 18*

Pressure--Low T

e. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 14*

Pressure--Low any steam line M P

c E

= ,

M

6 .

TABLE 3.3-3 (Continued)

M

@; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8

R

, MINIMUM c- TOTAL NO. CHANNELS CHANNELS APPLICABLE g FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

2. Containment Spray

a. Manual Initiation 2 1 with 2 1,2,3,4 17 2 coincident switches

b. Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays

c. Containment Pressure-- 4 2 3 1,2,3 15 R Hi-3

=

Y 3. Containment Isolation 0

a. Phase "A" Isolation

1) Manual Initiation 2 1 2 1,2,3,4 17

2) Automatic Actuation 2 1 2 1,2,3,4 13 Logic and Actuation Relays q

3) Safety Injection See Item 1. above for all Safety Injection initiating functions and Z requirements. y

b. Phase "B" Isolation

1) Manual Initiation 2 1 with 2 1,2,3,4 17 O ro 2 coincident N switches y

t>

=

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 7c TOTAL NO. CHANNELS MINIMUM CHANNELS APPLICABLE

' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

4. Steam Line Isolation (continued)

b. Automatic Actuation 2 1 2 1,2,3 20 Logic and Actuation Relays

c. Containment Pressure-- 3 2 2 1,2,3 14*

. Hi-2

d. Steam Line 3/ steam line 2/ steam line 2/ steam line 1, 2, 3# 14*

Pressure-Low any steam line Y e. Steam Generator

[ Pressiere - Negative 3/ steam line 2nteam line 2/ steam line 3** 14*

4 Rate--High any steam e line l S. Turbine Trip

a. Automatic Actuation 2 1 2 1, 2 22 Logic and Actuation Relays

b. Steam Generator 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 18" Water Level-- q High-High (P-14) -

c._ E c 6. Feedwater Isolation y

[ a. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1, 2 18*

Level--High-High (P-14)

R c'

b. 4 2 3 1, 2 18

Low RCS T,yg Coincident with Reactor Trip q M

o TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION x

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

' FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION c-h c. Safety Injection See Item 1. above for all Safety Injection initiating functions y and requirements.

7. Emergency Feedwater

a. Manual Initiation (1) Motor driven pump 1 1 1 1,2,3 21 (2) Turbine driven pump 2 1 2 1,2,3 21

b. Automatic Actuation Logic 2 1 2 1,2,3 20 and Actuation Relays

, c. Stm. Gen. Water Level--

g low-Low Z Start Motor-Driven Pump 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 18*

o and Start Turbine -

Driven Pump

d. Safety Injection Start Motor-Driven Pump See Item 1. above for all Safety Injection initiating functions and and Turbine-Driven Pump requirements.

e. Loss of-Offsite Power M Start Motor-Driven Pump and Turbine- See Item 9 for Loss-of-Offsite Power initiating functions and Driven Pump requirements. P h

g 8. Automatic Switchover to P Containment Sump w

sn @

g a. Automatic Actuation 2 1 2 1,2,3,4 13 E Logic and Actuation Relays

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 8

o MINIMUM 7 TOTAL NO. CHANNELS CHANNELS APPLICABLE c FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION z

b. RWST Level--Low-Low 4 2 3 1,2,3,4 18*

Coincident With:

Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

9. Loss of Power (Start Emergency Feedwater)

a. 4.16 kV Bus E5 and E6- 2/ bus 2/ bus 1/ bus 1, 2, 3, 4 18*

Loss of Voltage

b. 4.16 kV Bus E5 and E6-

, Degraded Voltage 2/ bus 2/ bus 1/ bus 1, 2, 3, 4 18*

N Coincident with SI

[ See Item 1. above for all Safety Injection initiating functions g, and requirements.

10. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, 3 2 2 1,2,3 19 P-11

b. Reactor Trip, P-4 2 2 2 1,2,3 21

c. Steam Generator Water 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 18* ummy g Leve1, P-14 ======

2 3:=

h .

e.,

i !!

a

4 O

i W TABLE 3.3-4 M

g; ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

R

, SENSOR c TOTAL ERROR 3

-e FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

- 1. Safety Injection (Reactor Trip, Feedwater Isolation, Start Diesel Generators, Phase "A" Isolation, Containment Ventilation Isolation, and Emergency Feedwater, Service Water to Secondary Component Cooling Water Isolation, and CBA Emergency Fan / Filter Actuation).

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

y c. Containment Pressure--Hi-1 4.2 0.71 1.67 1 4.3 psig 1 5.3 psig

d. Pressurizer Pressure--Low 13.1 10.71 1.69 1 1850 psig 1 1840 psig

e. Steam Line Pressure--Low 13.1 10.71 1.63 > 585 psig > 568 psig* -

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. N.A. N.A. p

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays 7

c. Containment Pressure--Hi-3 3.0 0.71 1.67 < 18.0 psig < 18.7 psig Gst e

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E!

o SENSOR 7 TOTAL ERROR c FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT z ALLOWABLE VALUE U 7. Emergency Feedwater s

a. Manual Initiation (1) Motor driven pump N.A. N.A. N.A. N.A. N.A.

(2) Turbine driven pung N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Steam Generator Water 17.0 15.28 1.76 > 17.0% of > 15.9% of narrow Level--Low-Low iiarrow range range instrument

, Start Motor-Driven Pump instrument span.

g and Start Turbine-Driven span.

w Pump

$ d. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints end Start Motor-Driven Pump Allowable Values.

and Turbine-Driven Pump

e. Loss-of-Offsite Power See Item 9. for loss-of-Offsite Power Setpoints and Allowable Values.

Start Motor-Driven Pump and Turbine-Driven Pump

8. Automatic Switchover to Containment Sump T

a. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A. >

P

b. RWST Level--Low-Low 2.75 1.8 h Coincident With

1. 0 >122,525 gals.

~>121,609 gals. g 2 Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and

" Allowable Values.

9 E. H

a -

TABLE 3.3-4 (Continued)

N g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 8

g SENSOR

, TOTAL ERROR c FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 5

9. Loss of Power (Start w Emergency Feedwater)

a. 4.16 kV Bus E5 and E6 N.A. N.A. N.A. > 2975 > 2908 volts Loss of Voltage volts with with a < 1.315 a < 1.20 second time second time delay.

delay. -

b. 4.16 kV Bus E5 and E6 N.A. N.A. N.A. > 3933 volts > 3902 volts Degraded Voltage Gith a < 10 sith a < 10.96 second time second time delay. delay.

w Coincident with:

D Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and w Allowable Values.

10. Engineered Safety Features Actuation System Interlocks

a. Pressurizer Pressure, P-11 N.A. N.A. N.A. $ 1950 psig 5 1960 psig

b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.

c. Steam Generator Water Level, See Item 5. above for all Steam Generator Water Level Trip P-14 Setpoints and Allowable Values. ,

Z M

P a

Q

t o

o TABLE 4.3-2 (Continued) l4 s ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP c ANALOG ACTUATING H0 DES 5

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY w RELAY SURVEILLANC FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

7. Emergency Feedwater (Continued)

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) 1,2,3 Q

and Actuation Relays

c. Steam Generator Water S R M N.A. N.A. N.A N.A 1,2,3 Level-Low-Low, Start Motor-Driven Pump and Turbine-Driven Pump

$ d. Safety Injection, Start See Item 1. above for all Safety Injection Surveillance Requirements.

Motor-Driven Pump and J, Turbine-Driven Pump a

e. Loss-of-Offsite Power See Item 9. for all Loss-of-Offsite Power Surveillance Requirements.

Start Motor-Driven Pump and Turbine-Driven Pump

8. Automatic Switchover to Containment Sump TE

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) -

M(1) Q 1, M Logic and Actuation Relays F"""

h g

b. RWST Level-tow-Low Coincident With S R M N.A. N.A. N.A. N.A 1, 2@

en c; Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

H

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS s!

, TRIP c ANALOG ACTUATING MODES 3

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIOi!AL ACTUATION RELAY RELAY SURVEILLANC w FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

9. Loss of Power (Start Emergency Feedwater)

a. 4.16 kV Bus E5 and N.A. R N.A M N.A. N.A. N.A. 1,2,3,4 E6 Loss of Voltage

b. 4.16 kV Bus E5 and N.A. R. N.A. M N.A. N.A. N.A. 1,2,3,4 E6 Degraded Voltage Coincident With

, Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements s

[ 10. Engineered Safety i

Features Actuation U System Interlocks

a. Pressurizer N.A. R M H.A. N.A. N.A. N.A. 1,2,3 Pressure, P-11

b. Reactor Trip, P-4 N.A. N.A N.A. N.A. R N.A. N.A. 1, 2, 3

c. Steam Generator S R M N.A. M(1) M(1) Q 1, Water Level, P-14 -

2 TABLE NOTATION P

(1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

b

= (2) A DIGITAL CHANNEL OPERATIONAL TEST will be performed on this instrumentation.

to c.=,

b e H

TABLE 3.3-6

, RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 9

o MINIMUM CHANNELS CHANNELS APPLICABLE ALARM / TRIP 7 FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION g 1. Containment U a. Containment - Post LOCA - 1 2 All < 10 R/h 27 H High-Area Monitor

b. RCS Leakage Detection

1) Particulate Radioactivity N.A. 1 1,2,3,4 N.A. 26

2) Gaseous Radioactivity N.A. 1 1,2,3,4 N.A. 26

2. Containment Ventilation Isolation

a. On Line Purge Monitor 1 2 1,2,3,4

23 y b. Manipulator Crane Area Monitor 1 2 5, 6 ** 23

[* 3. Main Steam Line 1/ steam line 1/ steam 1, 2, 3, 4 N.A. 27 line U 4. Fuel Storage Pool Areas

a. Fuel Storage Building "T1 Exhaust Monitor N.A. 1 *** **** 25 E5 6

5. Control Room Isolation p

a. Air Intake-Radiation Level P

1) East Air Intake 1/ intake 2/ intake All 24

2) West Air Intake 1/ intake 2/ intake All **** 24 @

6. Primary Component Cooling Water

a. Loop A 1 1 All <2x 28 Hackground

b. Loop B 1 1 All <2x 28 g Background

[ TABLE NOTATIONS

Two times background; purge rate will be verified to ensure compliance with Specification 3.11.2.1 requirements <

g **Twotimesbackgr{ndor15mR/hr,whicheverisgreater.

- - With irradiated fuel in the fuel storage pool areas.

- - - Two times background or 100 CPM, whichever is greater.

TABLE 3.3-6 (Continued)

ACTION STATEMENTS ACTION 23 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment ventilation isolation valves are maintained closed.

ACTION 24 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the Co;. trol Room Emergency Ventilation System in the recirculation mode of operation.

ACTION 25 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all ,

operations involving fuel movement in the fuel storage pool areas.

ACTION 26 - Must satisfy the ACTION requirement for Specification 3.4.6.1.

ACTION 27 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel (s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.8.2 within 14 days following the event outlining the actions taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 28 - With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirement, collect grab samples daily from the Primary Component Cooling Water System and the Service Water System and analyze the radioactivity until the inoperable Channel (s) is restored to OPERABLE status.

SEABROOK - UNIT 1 3/4 3-38 M P 5 sq

FINAL DRAFT TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerographs

a. 1-SM-XT-6700 Free Field East Cont. i Ig 1*

Room Air Intake

b. 1-SM-XT-6701 Containment Foundation i lg 1*

c. 1-SM-XT-6710 Cont. Opr. Floor i lg 1*

2. Triaxial Peak Accelerographs

a. 1-SM-XR-6702 Reactor Vessel Support 0-20 Hz. 1

b. 1-SM-XR-6703 Reactor Cool. Piping 0-20 Hz. I

c. 1-SM-XR-6704 PCCW Piping 0-20 Hz. 1

3. Triaxial Seismic Switches

a. 1-SM-XS-6700 Free Field N.A. 1*

b. 1-SM-XS-6701 Containment Foundation N.A. 1*

c. 1-SM-XS-6709 Containment Foundation 0.025g to 0.25g 1*

d. 1-SM-XS-6710 Cont. Opr. Floor N.A. 1*

4. Triaxial Response-Spectrum Recorders

a. 1-SM-XR-6705 Containment Foundation 1-30 Hz. 1*

b. 1-SM-XR-6706 SG llB Support 1-30 Hz. 1

c. 1-SM-XR-6707 Prim. Aux. Bldg. 1-30 Hz. 1

d. 1-SM-XR-6708 Service Water Pump House 1-30 Hz. 1 ;

I

With reactor control room indication SEABROOK - UNIT 1 3/4 3-42 slUN 2y p--

l

TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

1. Triaxial Time-History Accelerographs

a. 1-SM-XT-6700 Free Field East Cont. M* R SA Room Air Intake

b. 1-SM-XT-6701 Containment Foundation M* R N.A.

c. 1-SM-XT-6710 Cont. Opr. Floor M* R N.A.

2. Triaxial Peak Accelerographs

a. 1-SM-XR-6702 Reactor Vessel Support N.A. R N.A.

b. 1-SM-XR-6703 Reactor Cool. Piping N.A. R N.A.

c. 1-SM-XR-6704 PCCW Piping N.A. R N.A.

3. Triaxial Seismic Switches

a. 1-SM-XS-6700 Free Field ** M R SA

b. 1-SM-XS-6701 Containment Foundation ** M R N.A.

c. 1-SM-XS-6709 Containment Foundation ** M R N.A.

d. 1-SM-XS-6710 Cont. Opr. Floor ** M R N.A.

4. Triaxial Response-Spectrum Recorders

a. 1-SM-XR-6705 Containment Foundation ** M# R N.A.

b. 1-SM-XR-6706 SG llB Support N.A. R N.A.

c. 1-SM-XR-6707 Prim. Aux. Bldg. N.A. R N.A.

d. 1-SM-XR-6708 Service Water Pump House N.A. R N.A.

l

Except seismic trigger

- With reactor control room indications.

CHANNEL CHECK to consist of turning the test / reset switch and verify all lamps illuminate on 1-SM-XR-6705.

s SEABROOK - UNIT 1 3/4 3-43 JUN 2 51566

TABLE 3.3-8 MAL DiWT METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE

1. Wind Speed

a. Lower Level Nominal Elev. 43 ft 1

b. Upper Level Nominal Elev. 209 ft 1

2. Wind Direction

a. Lower Level Nominal Elev. 43 ft 1

b. Upper Level Nominal Elev. 209 ft 1

3. Air Temperature - AT

a. Lower Level Between Elev. 43 ft and 150 ft 1

b. Upper Level Between Elev. 43 ft and 209 ft 1 SEABROOK - UNIT 1 3/4 3-45 JUN ni 1g

FINAI. DR!1T INSTRUMENTATION MONITORING INSTRUMENTATION REMOTE SHUT 00WN SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.5 The Remote Shutdown System transfer switches, power, controls and monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With the number of OPERABLE remote shutdown monitoring channels less than the Total Number of Channels as required by Table 3.3-9, within 60 days restore the inoperable channel (s) to OPERABLE status or, pursuant to Specification 6.8.2, submit a Special Report that defines the corrective action to be taken.

c. With one or more Remote Shutdown System transfer switches, power, or control circuits inoperable, restore the inoperable switch (s)/

circuit (s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.5.1 Each remote shutdown monitoring instrumentation channel in Table 3.3-9 shall be demonstrated OPERABLE:

a. Every 31 days by performance of a CHANNEL CHECK, and

b. Every 18 months by performance of a CHANNEL CALIBRATION.

4.3.3.5.2 Each Remote Shutdown System transfer switch, power and control circuit listed in Table 3.3-9, including the actuated components, shall be demonstrated OPERABLE at least once per 18 months.

SEABROOK - UNIT 1 3/4 3-46 3Ri 2 51986

TABLE 3.3-9 g REMOTE SHUIDOWN SYSTEM

$ TOTAL NO. MINIMUM g READOUT OF CHANNELS Pc INSTRUMENT LOCATION CHANNELS OPERABLE e

g 1. Intermediate Range Neutron Flux CP-108 A and B 2 1 p 2. Source Range Neutron Flux CP-108 A and B 2 1 g 3. Reacter Coolant Temperature -

Average-Wide Range for Loops 1 and 4

a. T CP-108 A 2 2 c

b. T CP-108 B 2 2 H

4. Pressurizer Pressure CP-108 A and B 2 2

5. Pressurizer Level CP-108 A and B 2 2

6. Steam Generator Pressure CP-108 A and B 1/stm. gen. 1/stm. gen.

7. Steam Generator Water Level CP-108 A and B 1/stm. gen. 1/stm. gen.

m 8. Steam Generator-Emergency Feedwater

} Flow Rate CP-108 A and B 1/stm. gen. 1/stm. gen.

, 9. Boric Acid Tank Level CP-108 A and B 1/ tank 1/ tank

$ TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION

1. Emergency Feedwater Pump Steam Supply Valves MS-V-393/127 CP-108 A

2. Emergency Feedwater Pump Steam Supply Valves MS-V-394/128 CP-108 8

3. Emergency Feedwater Pump Steam Supply Valves MS-V-395 CP-108 A and B

4. Emergency Feedwater Pump FW-P-378 BUS 6 SWGR

5. Emergency Feedwater Recirculation Valve FW-V-346 CP-108 A

6. Emergency Feedwater Recirculation Valve FW-V-347 CP-108 B

7. SG A EFW Control Valve FW-FV-4214 A CP-108 A

8. SG A EFW Control Valve FW-FV-4214 B CP-108 B

9. SG B EFW Control Valve FW-FV-4224 A CP-108 A q

10. SG B EFW Control Valve FW-FV-4224 B

11. SG C EFW Control Valve FW-FV-4234 A CP-108 B CP-108 A g

12. SG C EFW Control Valve FW-FV-4234 B CP-108 B c 13. SG D EFW Control Valve FW-FV-4244 A CP-108 A P E 14. SG D EFW Control Valve FW-FV-4244 B CP-108 8 eo m

15. SG A Atmospheric Relief Valve MS-PV-3001

16. SG B Atmospheric Relief Valve MS-PV-3002 CP-108 A E CP-108 B g 17. SG C Atmospheric Relief Valve MS-PV-3003 CP-108 A

TABI'

3-9 g REMOTE SHuiDOWN SYSTEM E TOTAL NO. MINIMUM 7c READ 0UT OF CHANNELS TRANSFER SWITCHES / CONTROL CIRCUITS LOCATION CHANNELS OPERABLE o

c- 18. SG D Atmospheric Relief Valve MS-PV-3004 5 CP-108 B

19. MS Isolation Valves MS-V-86/88/90/92 CP-108 A

[ 20. MS Isolation Valves MS-V-86/88/90/92 CP-108 B

21. Pressurizer Heaters, Group A CP-108 A

22. Pressurizer Heaters, Group B CP-108 B

23. Charging Pump CS-P-2A BUS 5 SWGR

24. Charging Pump CS-P-2B BUS 6 SWGR

25. Charging Pump Suction from RWST CS-LCV-112D CP-108 A

26. Charging Pump Suction from RWST CS-LCV-112E CP-108 8

27. Pressurizer Relief Valve (PORV) RC-PCV-456A CP-108 A

28. Pressurizer Relief Valve (PORV) RC-PCV-456B CP-108 B

29. PORV Block Valve RC-V-122 CP-108 A

30. PORV Block Valve RC-V-124 CP-108 B M

31. High Pressure Injection SI-V-138 CP-108 A

32. High Pressure Injection SI-V-139 CP-108 8 T 33. VCT Discharge Isolation Valve CS-LCV-112B CP-108 A

$ 34. VCT Discharge Isolation Valve CS-LCV-112C CP-108 8 3>

N P O Cl3 iiil N

TABLE 3.3-10

@l; ACCIDENT MONITORING INSTRUMENTATION

o y TOTAL MINIMUM

, NO. OF CHANNELS INSTRUMENT CHANNELS OPERABLE U 1. Containment Pressure

a. Normal Range 2 1

b. Extended Range 2 1

2. Reactor Coolant Outlet Temperature - TH0T (Wide Range) 4 2 Temperature

3. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 4 2

4. Reactor Coolant Pressure - Wide Range 2 1 w

A 5. Pressurizer Water Level 2 1

6. Steam Generator Pressure 2/ steam generator 1/ steam generator

7. Steam Generator Water Level - Narrow Range 1/ steam generator 1/ steam generator

8. Steam Generator Water Level - Wide Range 1/ steam generator 1/ steam generator

9. Refueling Water Storage Tank Water Level 2 1

10. Reactor Coolant System Subcooling Margin Monitor 2 1 M

11. Containment Building Water Level 2 1 E

12. Core Exit Thermocouples 4/ core quadrant 2/ core quadrant P g 13. Containment Post-LOCA Area Monitor 2 1 Q f,

a --4

4 -

TABLE 3.3-10 (Continued)

ACCIDENT MONITORING INSTRUMENTATION

o 8 TOTAL MINIMUM 7 NO. OF CHANNELS c INSTRUMENT CHANNELS OPERABLE z

14. Intermediate Range Neutron Flux 2 1

15. Intermediate Range Neutron Flux Rate 2 1

16. Containment Isolation Valve Position

2/ Penetration 1/ Penetration

17. Containment Enclosure Negative Pressure 2 1

18. Condensate Storage Tank Water Level ** 2 1

19. Reactor Vessel Level Indication System 2 1

20. Containment Hydrogen Concentration 2 1 Y

Applies to penetrations with 2 active valves in series. These valves are moved to the closed position by cutomatic signals.

- Calculated on basis of pressure sensed at suction to the Emergency Feedwater Pumps.

M r- ,

$E 2

O P3 (P h r

U

TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 8

E MINIMUM c CHANNELS w

INSTRUMENT OPERABLE ACTION w 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Test Tank Discharge 1 29

b. Steam Generator Blowdown Flash Tank Drain 1* 30

c. Turbine Building Sumps Effluent Line 1 30

2. Flow Rate Measurement Devices g a. Liquid Radwaste Test Tank Discharge 1 31 w b. Steam Generator Blowdown Flash Tank Drain 1* 31

c. Circulating Water Discharge 1** N.A.

3. Radioactivity Monitors Providing Alarm but Not Termination of Release

a. Primary Component Cooling Water System (In lieu of 1 32 service water monitors) mmmim

4. Rate of Change Monitor 3

a. Primary Component Cooling Water System Head Tank (In lieu of 1 33 e service water monitors) F tc ta U

0nly applicable when steam generator blowdown is directed to the discharge transition structure.

h ** Pump performance curves generated in place should be used to estimate flow.

TABLE 4.3-5 N

g RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS B

E

, CHANNEL

c. CHANNEL SOURCE CHANNEL OPERATIONAL 2

INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Test Tank Discharge D P R(2) P(1)

b. Steam Generator Blowdown Flash Tank Drain D M R(2) Q(1)

c. Turbine Building Sumps Effluent Line D M R(2) Q(1) g 2. Flow Rate Measurement Devices y a. Liquid Radwaste Test Tank Discharge

D(3) N.A. R N.A.

E

b. Steam Generator Blowdown Flash Tank Drain D(3) N.A. R N.A.

c. Circulating Water Discharge ** N.A. N.A. N.A.

3. Radioactivity Monitor Providing Alarm But Not Termination of Release

a. Primary Component Cooling Water System D M R(2) Q(1)

(In lieu of service water monitors)

4. Rate of Change Monitor g====

a. Primary Component Cooling Water System (In lieu of service water monitors)

D(4) N.A. R N.A. g e

^ Isolation of the flow path is accomplished by the Waste Test Tank Discharge Pump Trip Circuitry. "T"

- Pump curves may be used to estimate flow. 9

FINAL DRAFT !

TABLE 4.3-5 (Continued) ,

l TABLE NOTATIONS (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm / Trip Setpoint.

(2) The initial Ch.WlEL CALIBRATION shall be performed using one or more of the reference 2candards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

(4) CHANNEL CHECK shall consist of verifying indication of tank level during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SEABROOK - UNIT 1 3/4 3-59 Jt#i251986

~

TABLE 3.3-13 (Continued) r N

@" RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION n

, MINIMUM CHANNELS c: INSTRUMENT OPERABLE APPLICABILITY ACTION 5

-d 4. TURBINE GLAND SEAL CONDENSER EXHAUST w

a. Iodine Sampler 1 ***

35

b. Particulate Sampler 1 ***

35

c. Sampler Flow Rate Indicator 1 ***

32

5. RADI0 ACTIVE GAS WASTE SYSTEM CUBICLES EXPLOSIVE GAS MONITORING SYSTEM Hydrogen Monitors 1/ cubicle

36 R.

Y W

--=

r- ,

$? .a

w TABLE 4.3-6 I RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8

o

' CHANNEL MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE E INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED w

s 1. RADI0 ACTIVE GAS WASTE SYSTEM EXPLOSIVE GAS MONITORING SYSTEM 0xygen Monitor D N.A. Q(5) M **

(Process)

2. PLANT VENT-WIDE RANGE GAS MONITOR

a. Noble Gas Activity Monitor D M R(3) Q(2) *

{ b. Iodine Sampler W N.A. N.A. N.A. *

c. Particulate Sampler W N.A. N.A. N.A. *

d. Flow Rate Monitor D N.A. R Q**** *

e. Sampler Flow Rate Monitor D N.A. R Q i M r-Cll3 5 ::Xlll m

a

TABLE 4.3-6 (Continued) I RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

o 7 CHANNEL MODES FOR WHICH c CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE z INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED

- 3. GASEOUS WASTE PROCESSING SYSTEM (Providing Alarm and Automatic Termination of Release)

a. Noble Gas Activity Monitor D N.A. R(6) Q(1)

4. TURBINE GLAND SEAL CONDENSER EXHAUST

a. Iodine Sampler W N.A. N.A. N.A ***

$ b. Particulate Sampler W N.A. N.A. N.A. ***

c. Sampler Flow Rate Indicator D N.A. N.A. N.A. *

5. RADI0 ACTIVE GAS WASTE SYSTEM CUBICLE EXPLOSIVE GAS MONITORING SYSTEM

a. Hydrogen Monitors D N.A. Q(4) M

r-I a p@

g N

M

9 wee !

b u .

y TABLE 4.3-6 (Continued) nl aL5G' TABLE NOTATIONS At all times. l During RADI0 ACTIVE WASTE GAS SYSTEM operation. !

When the gland seal exhauster is in operation. j

- - - The CHANNEL OPERATIONAL TEST for the flow rate monitor shall consist of a verification that the Radiation Data Management System (RDMS) indicated flow is consistent with the operational status of the plant.

Noble Gas Monitor for this release point is based on the main condenser air evacuation monitor.

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm / Trip Setpoint.

(2) The Digital CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm Setpoint.

(3) The iniital CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall per-mit calibrating the system over its intended range of energy and measure-ment range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

a. One volume percent hydrogen, balance nitrogen, and

b. Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

a. One volume percent oxygen, balance nitrogen, and

b. Four volume percent oxygen, balance nitrogen.

(6) The channel calibration shall be performed using sources of various activities covering the measurement range of the monitor to verify that the response is linear. Sources shall be used to verify the monitor response only for the intended energy range.

SEABROOK - UNIT 1 3/4 3-66 JLM 2 51986 F

- --- -- -y

TABLE 4.4-1 MIMIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION l l

No. of Steam Generators per Unit Four Preservice Inspection Four First Inservice Inspection Two Second & Subsequent Inservice Inspections One (1)

TABLE NOTATION (1) The third and fourth steam generators that were not inspected during the first inservice inspection shall be inspected during the second and third inspections, respectively. For the fourth and subsequent inspections, the inservice inspection may be limited to one steam generator on a rotating schedule encompassing 12% of the tubes if the results of the previous in-spections of the four steam generators indicate that all steam generators are performing in a like manner. Note that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances, the sample sequence shall be modified to inspect the most severe conditions.

SEABROOK - UNIT 1 3/4 4-18 JUN 2 51906

Sb REACTOR COOLANT SYSTEM uu lim puM REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE LIMITING 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. 1 gpm total reactor-to-secondary leakage through all steam ger.erators and 500 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System,

~

e. 40 gpm CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 psig i 20 psig, and

f. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

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

ACTION:

a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE B0UNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY ,

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following l 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by l use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l l

l SEABROOK - UNIT 1 3/4 4-21 JUN 2 5 6

REACTOR COOLANT SYSTEM RNAL DRUT i

REACTOR COOLANT SYSTEM LEAKAGE OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

b. Monitoring the containment drainage sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; I

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving steady-state operation ** and at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter during steady-state operation, except that not more than 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months shall elapse between any two successive inventory balances; and

e. Monitoring the feactor H9ad Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Changes in a throttle valve position to increase flow to an RCP shall be made only within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after completing this surveillance. This sur-veillance shall be completed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after any change of throttle valve position.

- T,yg being changed by less than 5 F/ hour l

l l

l l SEABROOK - UNIT 1 3/4 4-22 JWI 2 51986

, , 13,, f ill h L U n. . ;

D u

~

\

t \

250

> h t \

R \

s X u 200 L

u. \

U \

UNACCEPTABLE g \ OPERATTON E150 \

e, \

8 \

>- \

5 \

b 200 \

E ! \

0 l ACCEPTABLE l \

1T

.L OPERATION

,_ \

b 50 a

R S

w 0 0 0 0 20 40 60 80 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131 SEABROOK - UNIT 1 3/4 4-28 M 2 5 1ges

FINAL DRAF i 2800 2600 LEAK 1EST_ _

f r j 2400 LIMIT j j j 2200 l (I l 2000 f

/ ;

- 1800 !

e5 / I

~

E5 1600

%;? [ }7 E 1400- ! !

Sb HEATUP / /

$$12e8 CURVE N/ /

g:s / / I c / - CRITICALITY LIMIT

/ ,

1000 i 8ASED ON INSERVICE de. /' '

/

ES HYDRDSTATIC TEST

/

~

800 TEMPERATURE (255 F) ---

/ FOR THE SERVICE PERIO[L__

UP TO 16 EFPY 600

___ - ' /

. 400 200 0

100 200 300 358 400 500 RCS ' TEMPERATURE (*F)

(10 F PER DIVISION)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY SEABROOK - UNIT 1 3/4 4-31 JUN 25 W

FINAL DRAF 2800 2600 2400 2200 C00L DOWN LIM:TS---s !

\ /

s 2000 I

}

- 1800 f sa E5 1600 !

%t ]

w o '(

me 1400 o /

$wA

,/

wo 1200 EE /

m' 1000 /

~

800 '

A 600

&Bk:8W i lM@:$/

400 3,t.

200 l

B l

100 200 300 35e 400 RCS

TEMPERATURE (*F)

(10 F PER OIVISION)

FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY SEABROOK - UNIT 1 3/4 4-32

.illN 2 5 g l

v 2 2500 i E

o .. . - - . _ ..__ . . _ . _ . . ..__ _ .. ._ . . _ ___ __ . . ._. .

'[

o VAUD FOR THE FIRST 16 ErPY. SETPOINT f

' [

i

_ CONTAINS WARGIN OF 50 F FOR c- TRANSIENT EFFECTS. T

J Z /

g 2000 G Y h . P=54 Psic, Tst40 F !

~ P=412.7 + 9.597 e.01420ST; T>t40 F )

E ,

O 1500 ,

t I i

/

C _ . . ._. - -. . _ _. . _ . ._. -_. . _ ... .-.

Y y

~~

g ,

[il /

t m

=

Il

/

i '>

3 1000 -

/

0 -

i f

/

l s /

\ a f 500 ,

I l l t

{

200 so too iso 200 250 soo ano %

j RCS TEMPER ATURE ( F)

FIGURE 3.4-4 RCS COLD OVERPRESSURE PROTECTION SETPOINTS h O M

I ~

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolant System vent path consisting of one vent valve and one block valve powered from emergency busses shall be OPERABLE and closed

at each of the following locations:

a. Reactor vessel head, and

b. Pressurizer steam space.

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

ACTION:

a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves and block valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both Reactor Coolant System vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the vent valves and block valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or 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 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.11.1 Each Reactor Coolant System vent path block valve not required to be closed by ACTION a. or b., above, shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel from the control room.

4.4.11.2 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by;

a. Verifying all manual isolation valves in each vent path are locked in the open position,

For an OPERABLE vent path using a power-operated relief valve (PORV) as the vent path, the PORV block valve is not required to be closed.

SEABROOK - UNIT 1 3/4 4-38 JLR 2 51986

3/4.5 EMERGENCY CORE COOLING SYSTEMS , a sE vsIL lJ gg 3/4.5.1 ACCUMULATORS ,

l HOT STANDBY, STARTUP, AND POWER OPERATION I I

LIMITING CONDITION FOR OPERATION 3.5.1.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a. The isolation valve open and power removed,

b. A contained borated water volume of between 6121 and 6596 gallons,

c. A boron concentration of between 1900 and 2100 ppm, and

d. A nitrogen cover pressure of between 585 and 664 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of a closed isola-tion valve, restore the inoperable accumulator to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and

2) Verifying that each accumulator isolation valve is open.

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the accumulator solution; and

Pressurizer pressure above 1000 psig.

SEABROOK - UNIT 1 3/4 5-1 3912 5 BBC,

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350 F SURVEILLANCE REQUIREMENTS 4.5.2 (Continued)

d. At least once per 18 months by:

1) Verifying automatic isolation and interlock action of the RHR i system from the Reactor Coolant System by ensuring that:

a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 365 psig, the interlocks nrevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 660 psig, the interlocks will cause the valves to automatically close.

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. At least once per 18 months, during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on (Safety Injection actuation and Automatic Switchover to Containment Sump) test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal charging pump, b) Safety Injection pump, and c) RHR pump.

f. By verifying that each of the following pumps develops the indicated

differential pressure on recirculation flow when tested pursuant to l Specification 4.0.5

l

1) Centrifugal charging pump, 1 2480 psid;

2) Safety Injection pump, 2 1445 psid; and

3) RHR pump, 2 176 psid.

SEABROOK - UNIT 1 3/4 5-6 Jim 2 51986

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T avg EQUAL TO OR LESS THAN 200 F LIMITING CONDITION FOR OPERATION 3.5.3.2 All Safety Injection pumps shall be inoperable.

APPLICABILITY: MODE 5 and MODE 6 with the reactor vessel head on.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS

, 4.5.3.2 All Safety Injection pumps shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position at least once per 31 days.

"An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

SEABROOK - UNIT 1 3/4 5-10 10N 2 5 W

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6.1.4 Primary containment internal pressure shall be maintained between 14.6 and 16.2 psia.

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

ACTION:

With the containment internal pressure outside of the limits above, restore the internal pressure to within the limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.4 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SEABROOK - UNIT 1 3/4 6-9 dlNf 2 51%6

CONTAINMENT SYSTEMS RVAl. D PRIMARY CONTAINMENT CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

a. Each 36-inch containment shutdown purge supply and exhaust isolation valve shall be closed and locked closed, and

b. The 8-inch containment purge supply and exhaust isolation valve (s) shall be sealed closed except when open for purge system operation for pressure control; for ALARA, respirable, and air quality consider-ations to facilitate personnel entry; and for surveillance tests that require the valve (s) to be open.

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

ACTION:

a. With a 36-inch containment purge supply or exhaust isolation valve open or not locked closed, close and lock close that valve or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one or more of the 8-inch containment purge supply or exhaust isolation valves open for reasons other than given in Specifica-tion 3.6.1.7.6 above, close the open 8-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one or more containment purge supply or exhaust isolation valves having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3, restore the inoperable valve (s) to OPERABLE status or isolate the affected penetration (s) so that the measured leakage rate does not exceed the limits of Specifications 4.6.1.7.2 or 4.6.1.7.3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and close the purge supply if the affected penetration is the exhaust penetration, otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The 8-inch containment purge supply and exhaust isolation valves may not be opened while in MODE 1 or MODE 2 until installations of the narrow range con-tainment pressure instrument channels and alarms are completed.

SEABROOK - UNIT 1 3/4 6-12 JUN 2 5 W

CONTAINMENT SYSTEMS 3/4.6.5 CONTAINMENT ENCLOSURE BUILDING l CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION 3.6.5.1 Two independent Containment Enclosure Emergency Air Cleanup Systems shall be OPERABLE. l APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Enclosure Emergency Air Cleanup System inoperable, re-store the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.5.1 Each Containment Enclosure Emergency Air Cleanup System shall be demonstrated OPERABLE: I

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

b. 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-cating with the system by:

1) Verifying that the cleanup system satisfies the in place pene-tration leakage testing acceptance criteria of less than 0.05%

and uses the test procedure guidance in Regulatory Posi-tions C.S.a, C.5.c, and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 2100 cfm i 10%;

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

ANSI H510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Rev. 2, March 1978.

SEABROOK - UNIT 1 3/4 6-21 JtNt 15 W

CONTAINMENT SYSTEMS '

FINAL DRAFT CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE EMERGENCY AIR CLEANUP SYSTEM SURVEILLANCE REQUIREMENTS 4.6.5.lb.2 (Continued) of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978, by showing a methyl iodide penetration of less than 2.14% when tested at a temperature of 30 C and at a relative humidity of 95% in accordance with ASTM-D3803; and

3) Verifying a system flow rate of 2100 cfm i 12% during system e operation when tested in accordance with ANSI N510-1980.

. c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 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, by showing a methyl iodide penetration of less than 2.14% when tested at a tem-perature of 30 C and at a relative humidity of 95% in accordance with ASTM-D3803.

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 2100 cfm i 10%,

2) Verifying that the system starts on a Safety Injection test signal,

3) Verifying that the filter cross connect valves can be manually opened,

4) Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch Water Gauge in the annulus within 4 minutes after a start signal, and

e. After each complete or partial replacement of a high efficiency particulate air (HEPA) filter bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a dioctyl phthalate (DOP) test aerosol while operating the system at a flow rate of 2100 cfm i 10%; and SEABROOK - UNIT 1 3/4 6-22 M 2 51gg

CONTAINMENT SYSTEMS FlNAl. DRAR CONTAINMENT ENCLOSURE BUILDING CONTAINMENT ENCLOSURE BUILDING INTEGRITY SURVEILLANCE REQUIREMENTS 4.6.5.1 (continued)

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in place penetration leakage testing acceptance criteria of less than 0.05%

in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 2100 cfm i 10%.

l I

l SEABROOK - UNIT 1 3/4 6-23 Jim 2 5 ISIE

PLANT SYSTEMS B TURBINE CYCLE 1

AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. One motor-driven emergency feedwater pump, and one startup feedwater pump capable of being powered from an emergency bus and capable of being aligned to the dedicated water volume in the condensate storage tank, and

b. One steam turbine-driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two emergency feedwater pumps inoperable, restore at least one emergency feedwater pump to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and restore both emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one emergency feedwater pump and the startup feedwater pump inoperable, restore both emergency feedwater pumps to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and all three pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With three auxiliary feedwater pumps .qrpkc4. 5, immediately initiate corrective action to restore at least one auxniary feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:

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

1) Verifying that the motor-driven emergency feedwater pump develops a discharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpm; SEABROOK - UNIT 1 3/4 7-3 M 2 51986

PLANT SYSTEMS i

TURBINE CYCLE AUXILIARY FEEDWATER SYSTEM SURVEILLANCE REQUIREMENTS 4.7.1.2.la. (Continued)

2) Verifying that the steam turbine-driven pump develops a discharge pressure of greater than or equal to 1460 psig at a flow of greater than or equal to 270 gpm when the secondary steam supply pressure is greater than 500 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;

3) Verifying that the startup feedwater pump develops a discharge pressure of greater than or equal to 1375 psig at a flow of greater than or equal to 425 gpm.

4) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position;

5) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER; and

6) Verifying that valves FW-156 and FW-163 are operable for alignment of the startup feedwater pump to the emergency feedwater header.

b. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Emergency Feedwater

, System Actuation test signal;

2) Verifying that each emergency feedwater pump starts as designed automatically upon receipt of an Emergency Feedwater Actuation System test signal;

3) Verifying that with all manual actions, including power source and valve alignment, the startup feedwater pump starts within the required elapsed time; and

4) Verifying that each emergency feedwater control valve closes on receipt of a high flow test signal.

SEABROOK - UNIT 1 3/4 7-4 JUh 2 51336

PLANT SYSTEMS FINAL DRAT TURBINE CYCLE CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) system shall be OPERABLE with

a. A volume of 212,000 gallons of water contained in the condensate storage tank, and

b. A concrete CST enclosure that is capable of retaining 212,000 gal-lons of water.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With the CST or the CST enclosure inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months restore the CST and the CST enclosure to OPERABLE status or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.3 The CST and the CST enclosure shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume in the CST is within its limits and the CST enclosure integrity is maintained.

i

)

1 I

SEABROOK - UNIT 1 3/4 7-6 M 2 5196ti

FINAEDER PLANT SYSTEMS TURBINE CYCLE ATMOSPHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.7.1.6 The atmospheric relief valves shall be OPERABLE.

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

ACTION:

With less than one atmospheric relief valve per steam generator OPERABLE, re-store the required atmospheric relief valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.6 Each atmospheric relief valve shall be demonstrated OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that the nitrogen accumulator tank is at a pressure greater than or equal to 500 psig.

b. Prior to startup following any refueling shutdown or cold shutdown of 30 days or longer, verify that all valves will open and close fully.

When steam generators are being used for decay heat removal.

SEABROOK - UNIT 1 3/4 7-10 JUN 2 51986

PLANT SYSTEMS

' l 3/4.7.3 PRIMARY COMPONENT COOLING WATER SYSTEM l i

LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent primary component cooling water loops shall be OPERABLE, including two OPERABLE pumps in each loop.

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

a. With one primary component cooling water (PCCW) pump inoperable, restore the required primary component cooling water pumps to OPERABLE status within 7 days or be in at least HOT STANDBY within the next ,

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months

b. With two primary component cooling water pumps inoperable, restore at least one of the inoperable primary component cooling water pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following )

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With two primary component cooling water pumps within one loop inoperable, restore at least one of the inoperable pumps to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.3 At least two primary component cooling water loops shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At least once per 18 months during shutdown, by verifying that:

1) Each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Fea-ture actuation signal, and

2) Each of the four primary Component Cooling Water System pump starts automatically upon loss of or failure to start of its redundant pump within the loop.

SEABROOK - UNIT 1 3/4 7-12 M 25 k

PLANT SYSTEMS flNAL DRAFT 3/4.7.4 SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE, in-cluding three OPERABLE pumps in each loop.

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

ACTION:

a. With one service water pump inoperable, restore the required service water pumps to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two service water pumps inoperable, restore at least one of the inoperable service water pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With two service water pumps within one loop inoperable, restore at least one of the inoperable pumps to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With one cooling tower service water pump inoperable, restore the required cooling tower service water pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.4 At least two Station Service Water loops shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At least once per 18 months during shutdown, by verifying that:

1) Each automatic valve servicing safety-related equipment actuates to its correct position on its associated Engineered Safety Fea-ture actuation test signal, and

2) Each of the four Station pumps aligned to the ocean ultimate heat sink (UHS) starts automatically. upon loss of or failure to start of the redundant pump within the loop and each of the two pumps aligned to the cooling tower UHS starts on a cooling tower actuation (TA) signal.

SEABROOK - UNIT 1 3/4 7-13 JJ. ; t *. ,

O

PLANT SYSTEMS Hnatum..i 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink (UHS) shall be OPERABLE with:

a. A service water pumphouse water level at or above minus 37'-0" Mean Sea Level, USGS datum, and

b. A mechanical draft cooling tower comprised of one cooling tower cell with one OPERABLE fan and a second cell with two OPERABLE fans, and a contained basin water level of equal to or greater than 35.9* feet at a bulk average water temperature of less than or equal to 67.3 F, and

c. A portable tower makeup pump system stored to be OPERABLE for 30 days fcllowing a Safe Shutdown Earthquake.

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

ACTION:

a. With the service water pumphouse inoperable, restore the service water pumphouse to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the mechanical draft cooling tower inoperable, restore the cooling tower to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the portable tower makeup pump system inoperable, continue operation and notify the NRC within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in accordance with the procedure of 10 CFR 50.72 of actions or contingencies to ensure an adequate supply of makeup water to the mechanical draft cooling tower for a minimum of 30 days.

l l

With the cooling tower in operation with valves aligned for tunnel heat treat- 1 ment, the tower basin level shall be maintained at greater than or equal to '

34.3 feet and the bulk average temperature shall be maintained at less than or equal to 71 F.

SEABROOK - UNIT 1 3/4 7-14 JUN 2 51935

PLANT SYSTEMS f g DRAFT U" d ULTIMATE HEAT SINK SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by:

1) Verifying the water level in the service water pumphouse to be at or above minus 37'-0" Mean Sea Level, and

2) Verifying the water in the mechanical draft cooling tower basin to be greater than or equal to a level of 35.9 feet.

b. At least once per week by verifying that the water in the mechanical draft cooling tower basin to be at a bulk average temperature of 67.3 F.

c. At least once per 31 days by:

1) Starting from the control room each UHS cooling tower fan that is required to be OPERABLE and operating each of those fans for at least 15 minutes, and

2) Verifying that the portable tower makeup pump system is stored in its design operational readiness state.

d. At least once per 18 months by verifying automatic actuation of each cooling tower fan on a Tower Actuation test signal.

SEABROOK - UNIT 1 3/4 7-15 JUN 2 51986

. e b

3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 Two independent Control Room Area Ventilation Systems shall be OPERABLE.

APPLICABILITY: All MODES.

ACTION:

MODES 1, 2, 3, and 4:

With one Control Room Area Ventilation System inoperable, N. store the inoperable system to OPERABLE status within 7 days or be ir it least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6:

a. With one Control Room Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE Control Room Area Ventilation System in the recirculation mode.

b. With both Control Room Area Ventilation Systems inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.7.6 Each Control Room Area Ventilation System shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the Control Room Area Ventilation System is maintaining the temperature of equipment and instrumentation in the control room area below its limiting equipment qualification temperature.

b. At least once per 18 months or after any significant modification to the Control Room Area Ventilation Systems by verifying a system flow rate of 25,700 cfm i 10% through the air conditioner unit (3A and 3B) and a flow rate of at least 500 cfm makeup from each intake to the emergency filtration unit with a discharge of 2000 cfm i 10% from the filtration unit.

4 l

SEABROOK - UNIT 1 3/4 7-16 JUN 2 5 IE

PLANT SYSTEMS CONTROL ROOM AREA VENTILATION SYSTEM SURVEILLANCE REQUIREMENTS 4.7.6 (Continued)

c. At least once per 18 months by:

1) Verifying that on a high radiation signal from the control room makeup air intake, the subsystem automatically switches to the emergency recirculation mode of operation and the isolation dampers close within 5 seconds.

2) Verifying that on an S signal the emergency filtration fans start.

3) Verifying that the system maintains the control room area at a positive pressure of greater than or equal to a pressurization 1/8-inch Water Gauge relative to adjacent areas during system operation at less than or equal to a pressurization flow of 550 cfm.

SEABROOK - UNIT 1 3/4 7-17 JUN 25 386

TABLE 3.7-3 t

AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F)

1. Control Room 75

2. Cable Spreading Room 104

3. Switchgear Room - Train A 104

4. Switchgear Room - Train B 104

5. Battery Rooms - Train A 97

6. Battery Rooms - Train B 97

7. ECCS Equipment Vault - Train A 104

8. ECCS Equipment Vault - Train B 104

9. Centrifugal Charging Pump Room - Train A 104

10. Centrifugal Charging Pump Room - Train B 104

11. ECCS Equipment Vault Stairwell - Train A 104 ,

12. ECCS Equipment Vault Stairwell - Train B 104

13. PCCW Pump Area 104

14. Cooling Tower Switchgear Room - Train A 104

15. Cooling Tower Switchgear Room - Train B 104

16. Cooling Tower SW Pump Area 127

17. SW Pumphouse Electrical Room - Train A 104

18. SW Pumphouse Electrical Room - Train B 104

19. SW Pump Area 104

20. Diesel Generator Room - Train A 120

21. Diesel Generator Room - Train B 120

22. EFW Pumphouse 104

23. Electrical Penetration Area - Train A 100

24. Electrical Penetration Area - Train B 85

25. Fuel Storage Building Spent Fuel Pool Cooling 104 Pump Area

26. Main Steam and Feedwater Pipe Chase - East 130

27. Main Steam and Feedwater Pipe Chase - West 130 SEABROOK - UNIT 1 3/4 7-23 p 2 5 ISS- ,

1

ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 (Continued)

ACTION:

diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of initial loss or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one diesel generator inoperable in addition to ACTION a. or

b. above, verify that:

1. All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and

2. When in MODE 1, 2, or 3, the steam-driven emergency feedwater pump is OPERABLE.

If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With two of the above required offsite A.C. circuits inoperable, demonstrate the OPERABILITY of two diesel generators by performing the requirements of Specification 4.8.1.1.2a.5) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generators are already operating; restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . With only one offsite source restored, restore at least two offsite circuits to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

e. With two of the above required diesel generators inoperable, demon-strate the OPERABILITY of two offsite A.C. circuits by performing the requirements of Specification 4.8.1.1.la. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore at least two diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SEABROOK - UNIT 1 3/4 8-2 EIN t 51986

o .

ELECTRICAL POWER SYSTEMS Q A.C. SOURCES e .

OPERATING i SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) c) Simulated loss of-offsite power in conjunction with an SI Actuation test signal, or d) An SI Actuation test signal by itself.

6) Verifying the generator is synchronized, loaded to greater than or equal to 6083 kW in less than or equal to 120 seconds *,

and operates with a load greater than or equal to 6083 kW for at least 60 minutes; and

7) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the day fuel tank;

c. At least once per 31 days by checking for and removing accumulated water from the fuel oil storage tanks;

d. By sampling new fuel oil in accordance with ASTM-D4057-81 prior to addition to storage tanks and:

1) By verifying in accordance with the tests specified in ASTM-D975-81 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degree at 60 F, or a specific gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate, or an absolute specific gravity at 60/60 F of greater than or equal to 0.81 but less than or equal to 0.89, or an API gravity of greater than or equal to 28 degrees but less than or equal to 42 degrees;

All diesel generator starts for the purpose of this surveillance test may be preceeded by an engine prelube period. Further, all surveillance tests and all other engine starts for the purpose of this surveillance tests, with the exception of once per 184 days, may also be preceded by warmup procedures (e.g.,

gradual acceleration and/or gradual loading greater than 60 seconds) as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized.

SEABROOK - UNIT 1 3/4 8-4 AN 2 5 506

_ --f 5 Eliflam ads h'13 ELECTRICAL POWER SYSTEMS A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) b) A kinematic viscosity at 40 C of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the supplier's certification; c) A flash point greater than or equal to 125'F; and d) A clear and bright appearance with proper color when tested in accordance with ASTM-D4176-82.

2) By verifying within 30 days of obtaining the sample that the other properties specified in Table 1 of ASTM-D975-81 are met when tested in accordance with ASTM-0975-81 except that the analysis for sulfur may be performed in accordance with ASTM-D1552-79 or ASTM-D2622-82.

e. At least once every 31 days

1) By obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTH-D2276-78, Method A, and

2) By visually inspecting the lagging in the area of the flanged joints on the silencer outlet of the diesel exhaust system for leakage (also after an extended operation of greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

f. At least once per 18 months, during shutdown, by:

1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service;

2) Verifying the generator capability to reject a load of greater than or equal to 671 kW while maintaining voltage at 4160 420 volts and frequency at 60 1 4.0 Hz;

3) Verifying the generator capability to reject a load of 6083 kW without tripping. The generator voltage shall not exceed 4784 volts during and following the load rejection;

4) Simulating a loss-of-offsite power by itself, and:

SEABROOK - UNIT 1 3/4 8-5 JUN 2 51966

o PT ELECTRICAL POWER SYSTEMS EE H L 5# l A.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued) a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected shutdown loads through the emergency power sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After ener-gization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 i 420 volts and 60 1 1.2 Hz during this test.

5) Verifying that on an SI actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 4160 420 volts and 60 1 1.2 Hz within 10 seconds after the auto-start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test;

6) Simulating a loss-of-offsite power in conjunction with an SI actuation test signal, and tower actuation test signal (TA);

and a) Verifying deenergization of the emergency busses and load

, shedding from the emergency busses; b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 12 seconds, energizes the auto-connected emergency (accident) loads through the emergency power sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 1 420 volts and 60 1 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engine overspeed, low lube oil pressure, 4160-volt bus fault, and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concur-rent with a Safety Injection actuation signal.

SEABROOK - UNIT 1 3/4 8-6 JUN 2 51986

ELECTRICAL POWER SYSTEMS A.C. SOURCES l OPERATING SURVEILLANCE REQUIREMENTS 4.8.1.1.2 (Continued)

13) Verifying that the following diesel generator lockout features prevent diesel generator starting:

a) Barring device engaged, or b) Differential lockout relay.

14) Simulating a Tower Actuation (TA) signal while the diesel generator is loaded with the permanently connected loads and auto-connected emergency (accident) loads, and verifying that the service water pump automatically trips, the cooling tower pump and fan automatically starts. After energization the steady state voltage and frequency of the emergency buses shall be maintained at 4160 1 420 volts and 60 1 1.2 Hz; and

15) While diesel generator 1A is loaded with the permanently connected loads and auto-connected emergency (accident) loads, manually connect the 1500 hp startup feedwater pump to 4160-volt bus E5. After energization the steady-state voltage and frequency of the emergency bus shall be maintained at 4160 1 420 volts and 60 1 1.2 Hz.

g. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least 514 rpm in less than or equal to 10 seconds; and

h. At least once per 10 years by:

1) Draining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite solution, or equivalent, and

2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code at a test pressure equal to 110% of the system design pressure.

SEABROOK - UNIT 1 3/4 8-8 O25g

FINAL dim TABLE 4.8-1 8 DIESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN NUMBER OF FAILURES IN LAST 20 VALID TESTS

LAST 100 VALID TESTS

TEST FREQUENCY

<1 <5 At least once per 31 days

[2 [6 At least once per 7 days **

r

Criteria for determining the number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.

For the purpose of determining the required test frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like-new condition is completed, provided that the overhaul, including appropriate post-maintenance operation and testing, is specifically approved by the manufacturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests in a single series. Ten of these tests shall be in accordance with the routine Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 and four tests in accordance with the 184-day testing requirement of Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6. If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to transvalue the failure count to zero requires NRC approval.

- This test frequency shall be maintained until seven consecutive failure-free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one or less.

SEABROOK - UNIT 1 3/4 8-10 JUN 2 51986

rm o '

WEN u ELECTRICAL POWER SYSTEMS 3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE and energized:

a. Train A

1) 125-volt Battery Banks 1A and 1C,

2) One full-capacity battery charger on Bus #11A, and

3) One full-capacity battery charger on Bus #11C.

b. Train B

1) 125-volt Battery Banks 1B and ID,

2) One full-capacity battery charger on Bus #11B, and

3) One full-capacity battery charger on Bus #110.

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

ACTION:

a. With one of the required battery banks in one train inoperable, close the bus tie to connect the remaining operable battery bank to the D.C.

bus supplied by the inoperable battery bank within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; restore the inoperable battery bank to OPERABLE status within 30 days

or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one of the full-capacity chargers inoperable, demonstrate the operability of its associated battery bank by performing Surveillance Requirement 4.8.2.1.a.1) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and at least once per hour thereafter. If any Category A limit in Table 4.8-2 is not met, declare the battery inoperable.

SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the Category A limits, and

2) The total battery terminal voltage is greater than or equal to 128 volts on float charge.

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:

No more than one battery at a time may be taken out of service for more than 30 days.

SEABROOK - UNIT 1 3/4 8-12 JUN 2 5 $86 1

ELECTRICAL POWER SYSTEMS D.C. SOURCES OPERATING SURVEILLANCE REQUIREMENTS 4.8.2.lb (Continued)

1) The parameters in Table 4.8-2 meet the Category B limits,

2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 8 ohm," and

3) The average electrolyte temperature of 16 connected cells (4 cells per row) is above 65 F.

c. At least once per 18 months by verifying that:

1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,

2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,

3) The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 8 ohm,* and

4) Each battery charger will supply at least 150 amperes at a minimum of 132 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test;

e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and

f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

0btained by subtracting the normal resistance of: (1) the cross room rack connector (210 x 10 8 ohm, typical) and (2) the bi-level rack connector (35 x 10 8 ohm, typical) from the measured cell-to-cell conection resistance.

SEABROOK - UNIT 1 3/4 8-13 JUN 2 51966

' 1" ELECTRICAL POWER SYSTEMS D.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two 125-volt battery banks in one D.C. Train and the associated full-capacity chargers shall be OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

a. With one of the required battery banks inoperable, immediately close the bus tie to the alternate OPERABLE battery.

b. With both required battery banks and/or full-capacity chargers inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required battery banks and full-capacity chargers to OPERABLE status as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System through a 1.58-square-inch vent.

SURVEILLANCE REQUIREMENTS 4.8.2.2 The above-required 125 volt battery banks and full-capacity chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.

SEABROOK - UNIT 1 3/4 8-15 JUN 2 51986

e , TIML Uihd l' ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the specified manner:

a. Train A, A.C. Emergency Busses consisting of:

1) 4160-volt Emergency Bus #ES,

2) 480-volt Emergency Bus #E51,** and

3) 480-volt Emergency Bus #E52.**

b. Train B, A.C. Emergency Busses consisting of:

1) 4160-volt Emergency Bus #E6,

2) 480-volt Emergency Bus #E61,**

3) 480-volt Emergency Bus #E62,** and

4) 480-volt Emergency Bus #E64.

c. 120-volt A.C. Vital Panel #1A energized from its associated inverter connected to D.C. Bus #11A,*

d. 120-volt A.C. Vital Panel #18 energized from its associated inverter connected to D.C. Bus #11B,*

e. 120-volt A.C. Vital Panel #1C energized from its associated inverter connected to D.C. Bus #11C,*

f. 120-volt A.C. Vital Panel #1D energized from its associated inverter connected to D.C. Bus #110,*

g. 120-volt A.C. Vital Panel #1E energized from its associated inverter

, connected to D.C. Bus #11A,*

h. 120-volt A.C. Vital Panel #1F energized from its associated inverter connected to D.C. Bus #11B,*

Two inverters may be disconnected from their D.C. bus for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as necessary, for the purpose of performing an equalizing charge on their associated battery bank provided: (1) their vital busses are energized, and (2) the vital busses associated with the other battery bank are energized from their associated inverters and connected to their associated D.C. bus.

- These busses can be considered OPERABLE if the 480 volt bus ties are closed.

These bus ties will be under administrative control to ensure loading is within transformer rating.

SEABROOK - UNIT 1 3/4 8-16 JUN 1t 5 M

nW ELECTRICAL POWER SYSTEMS l ll1 plM~

ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 (Continued)

i. Train A,125-volt D.C. Busses consisting of:

1) 125-volt D.C. Bus #11A energized from Battery Bank 1A or IC, and

2) 125-volt D.C. Bus #11C energized from Battery Bank 1C or 1A.

j. Train B,125-volt D.C. Busses consisting of:

1) 125-volt D.C. Bus #11B energized from Battery Bank 1B or 10, and

2) 125-volt D.C. Bus #11D energized from Battery Bank 1D or IB.

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

ACTION:

a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the train within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With one A.C. vital panel either not energized from its associated inverter, or with the inverter not connected to its associated D.C.

bus: (1) reenergize the A.C. Vital panel within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; and (2) reenergize the A.C. vital panel from its associated inverter connected to its associated D.C.

bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With one D.C. bus not energized from its associated battery bank, reenergize the D.C. bus from its associated battery bank or close the bus tie to the alternate OPERABLE battery of the same train within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses and panels shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

SEABROOK - UNIT 1 3/4 8-17 auN 2 5 Y

a 1 ELECTRICAL POWER SYSTEMS a st Wilk U e

ONSITE POWER DISTRIBUTION SHUTDOWN LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:

a. One train of A.C. emergency busses consisting of the 4160-volt and the 480-volt A.C. emergency busses listed in 3.8.3.la. and b.

(excluding 480-volt Emergency Bus #E64);

b. Two of the four 120-volt A.C. vital Panels 1A, 18, 1C, and 10 energized from their associated inverters connected to their 8 respective D.C. busses;

c. One of the two 120-volt A.C. Vital Panels 1E or IF energized from its associated inverter connected to the respective D.C. bus; and

d. Two 125-volt D.C. busses energized from their associated battery banks.

APPLICABILITY MODES 5 and 6.

ACTION:

With any of the above required electrical busses and panels not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses and panels in the specified manner as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the RCS through at least a 1.58-square-inch vent.

SURVEILLANCE REQUIREMENTS 4.8.3.2 The specified busses and panels shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

SEABROOK - UNIT 1 3/4 8-18

$N L L W1

a f ELECTRICAL POWER SYSTEMS h w b 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES A.C. CIRCUITS INSIDE PRIMARY CONTAINMENT LIMITING CONDITION FOR OPERATION 3.8.4.1 The circuit breakers feeding the following loads inside primary con-tainment shall be padlocked in the open position:

Loads Circuit Panel Refueling Canal Skimmer Pump 1-SF-P-272 1-ED-MCC-111 Polar Gantry Crane 1-MM-CR-3 1-ED-US-11 Distribution Panel 1-ED-PP-7A 1-ED-US-11 Distribution Panel 1-ED-PP-78 1-ED-US-23 Rod Control Cluster Change Fixture 1-FH-RE 12 1-ED-MCC-111 APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel (s) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

EXCEPTION:

If any of the above-mentioned loads are required for brief durations (not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) during plant operation, the pertinent circuit breaker can be unlocked and closed for the required duration provided this change in breaker position becomes part of the applicable operating procedure used for the work inside containment.

SURVEILLANCE REQUIREMENTS 4.8.4.1 Verify at least once per 31 days that the circuit breakers listed above are locked in the open position.

SEABROOK - UNIT 1 3/4 8-20 g

ELECTRICAL POWER SYSTEMS FINAL DMR l ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SURVEILLANCE REQUIREMENTS 4.8.4.2.a.1) (Continued) b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and c) For each circuit breaker found inoperable during these func-tional tests, one additional circuit breaker of the inoper-able type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

2) By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers and overload devices. Circuit breakers and overload devices selected for functional testing shall be selected on a rotating basis.

Testing of air circuit breakers shall consist of injecting a cur-rent with a value equal to 300% of the pickup of the long-time delay trip element and 150% of the pickup of the short-time de-lay trip element. The instantaneous element shall be tested by injecting a current equal to 120% of the pickup value of the element.

Testing of thermal magnetic molded-case circuit breakers shall consist of injecting a current with a value equal to 300% of the circuit breaker trip rating and -25% to +40% of the circuit breaker instantaneous trip range or setpoint.

Testing of combination starters (a magnetic only molded-case circuit breaker in series with a motor starter and integral overload device) shall consist of injecting a current with a value equal to -25% to +40% of the circuit breaker instantaneous trip setpoint, and 200% and 300% of the thermal overload device trip rating to the respective devices.

Circuit breakers and/or overload devices found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker and or overload de- .

vices found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers and or overload devices of the inoperable type shall also be func-tionally tested until no more failures are found or all circuit breakers and or overload devices of that type have been func-tionally tested. .

SEABROOK - UNIT 1 3/4 8-22 JUN 2 51986

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ELECTRICAL POWER SYSTEMS a EE Wi th ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES SURVEILLANCE REQUIREMENTS 4.8.4.2.a (Continued)

3) Corrective actions for any generic degradation of overcurrent protective devices, such as setpoint drift, manufacturing deficiencies, material defects, etc., shall be applicable to -

all (Class 1E and non-Class 1E) protective devices of identical design.

b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

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SEABROOK - UNIT 1 3/4 8-23 JUN 2 5 566

o 1 ELECTRICAL POWER SYSTEMS I lam E E ELECTRICAL EQUIPMENT PROTECTIVE DEVICES MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION LIMITING CONDITION FOR OPERATION 3.8.4.3 Each thermal overload protection for safety-related motor operated valves shall be OPERABLE.

APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE.

ACTION:

With the thermal overload protection for one or more of the above-required valves inoperable, bypass the inoperable thermal overload within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , restore the inoperable thermal overload to OPERABLE status within 30 days, or declare the affected valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected system (s).

SURVEILLANCE REQUIREMENTS 4.8.4.3 The thermal overload protection for the above required valves shall be demonstrated OPERABLE at least once per 18 months and following maintenance on the motor starter by selection of a representative sample of at least 25% of all thermal overloads for the above-required valves and replacing them with precalibrated devices that have been subjected to a CHANNEL CALIBRATION.

SEABROOK - UNIT 1 3/4 8-24 JUN 2 E $?i 9

REFUELING OPERATIONS FUEL STORAGE BUILDING EMERGENCY AIR CLEANING SYSTEM SURVEILLANCE REQUIREMENTS 4.9.12b (Continued)

1) Verifying that the cleanup system satisfies the in place I penetration and bypass leakage testing acceptance criteria l of less than 0.05% and uses the test procedure guidance in Regulatory Positions C.5.a, C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978,* and the system flow rate is 17,000 cfm i 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, Revision 2, March 1978, by showing a methyl iodide penetration of less than 1.0% when tested at a temperature of 30 C and at a relative hu-midity of 70% in accordance with ASTM-03803; and

3) Verifying a system flow rate of 17,000 cfm i 12% during system operation when tested in accordance with ANSI N510-1980.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation by 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,* by showing a methyl iodide penetration of less than 1.0% when tested at a tem-perature of 30 C and at a relative humidity of 95% in accordance with ASTM-03803. ,

, d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 17,000 cfm i 10%,

2) Verifying that the system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to 1/4 inch Water Gauge relative to the outside atmosphere during system operation.

ANSI H510-1980 shall be used in place of ANSI N510-1975 as referenced in Regulatory Guide 1.52, Rev. 2, March 1978.

SEABROOK - UNIT 1 3/4 9-14

' ' A> 2 7 G

O f 3/4.11 RADIOACTIVE EFFLUENTS ,

3/4.11.1 LIQUID EFFLUENTS 1 CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser (see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 4 microcurie /ml total activity.

APPLICABILITY: At all times. ,

AJIION:

With the concentration of radioactive material released in liquid effluents at the point of discharge from the multiport diffuser exceeding the above limits, restore the concentration to within the above limits within 15 minutes.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Part A of the ODCM.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1.

SEABROOK - UNIT 1 3/4 11-1 OUMt5 Q

o 9 RADIOACTIVE EFFLUENTS GASEOUS EFFLUENTS EXPLOSIVE GAS MIXTURE - SYSTEM LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 2% by volume.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM l greater than 2% by volume, reduce the oxygen concentration to the I above limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months unless the hydrogen concentration is verified to be less than 4% by volume.

~

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen or oxygen in the WASTE GAS HOLOUP SYSTEM shall be determined to be within the above limit by continuously monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen er oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.10-SEABROOK - UNIT 1 3/4 11-9

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O TABLE B 3/4.4-1 M

g REACTOR VESSEL TOUGHNESS 8

Si!

T RT Avg. Shelf Energy e Material Cu P NDT NDT NMWD* MWD" z Component Code No. Spec. No. @ @ (*F) (*F) (ft-lb) (ft-Lb)

H Closure Head Dome R1809-1 A533B,CL.1 0.15 0.011 -40 10 80.5 -

Closure Head Torus R1810-1 A533B,CL.1 0.08 0.012 50 0 104 -

Closure Head Flange R1802-1 A508,CL.2 -

0.013 10 10 105.5 -

Vessel Flange R1801-1 A508,CL.2 -

0.012 20 30 91 -

Inlet Nozzle R1804-1 A508,CL.2 0.10 0.011 0 0 125 -

Inlet Nozzle R1804-2 A508,CL.2 0.09 0.010 -20 -20 125 -

Inlet Nozzle R1804-3 A508,CL.2 0.08 0.010 -20 -20 131 -

Inlet Nozzle R1804-4 A508,CL.2 0.10 0.013 -20 -20 128 -

Outlet Nozzle R1805-1 A508,CL.2 -

0.003 -20 -10 115 -

, Outlet Nozzle R1805-2 A508,CL.2 -

0.004 -20 -20 132 -

, Outlet Nozzle R1805-3 A508,CL.2 -

0.009 -10 -10 128 -

g Outlet Nozzle R1805-4 A5@,CL.2 -

0.005 -10 -10 117 -

, Nozzle Shell R1807-1 A5d38,CL.1 0.08 0.011 -30 30 66 -

4 Nozzle Shell R1807-2 A533B,CL.1 0.09 0.012 -40 30 66.5 -

Nozzle Shell R1807-3 A533B,CL.1 0.06 0.010 -20 10 107 -

Inter. Shell R1806-1 A533B,CL.1 0.04 0.012 -30 40 82 139.5 Inter. Shell R1806-2 A533B,CL.1 0.05 0.007 -30 0 102 143.5 Inter. Shell R1806-3 A5338,CL.1 0.07 0.007 -40 10 115 138 Lower Shell R1808-1 A533B,CL.1 0.05 0.005 -30 40 78 120.5 Lower Shell R1808-2 A533B,CL.1 0.05 0.007 -20 10 77 127 Lower Shell R1808-3 A533B,CL.1 0.06 0.007 -20 40 78 130.5 Bottom Head Torus R1811-1 A5338,CL.1 0.15 0.010 -50 0 94.5 -

Bottom Head Dome R1812-1 A533B,CL.1 0.09 0.009 -30 0 97.5 - sumy' Inter. & Lower Shell -

Long Weld Seams G1.72 Sub Arc Weld 0.07 0.008 -50 -50 200 -

2 Inter. & Lower Shell Girth Weld Seam G1.72 Sub Arc Weld 0.07 0.008 -50 -50 200 p

p c

to en *NMWD - Normal to Major Working Direction E

g" ** MWD - Major Working Direction '

""'If' mm -

FINAL DRAFT 20 10 -

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4 10" 0 5 10 15 20 25 30 35 EFFECTIVE FULL POWER (YEARS)

FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE SEABROOK - UNIT 1 B 3/4 4-10 M 2 51986

REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued)

COLD OVERPRESSURE PROTECTION The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the COMS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) a maximum pres-sure overshoot beyond the PORV Setpoint which can occur as a result of time de-lays in signal processing and valve opening; (2) a 50 F heat transport effect made possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; (3) instrument uncertain-ties; and (4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lock-out of both Safety Injection pumps and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary coolant temperature is more than 50 F above reac-tor coolant temperature. Exceptions to these requirements are acceptable as described below.

Operation above 350 F but less than 375 F with only centrifugal charging pump OPERABLE and no Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . As shown by analysis, LOCAs occurring at low temperature, low pres-sure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator ;

injection. Given the short time duration and the condition of having only one I centrifugal charging pump OPERABLE and the probability of a LOCA occurring dur- I ing this time, the failure of the single centrifugal charging pump is not l assumed.

Operation below 350 F but greater than 325 F with all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . During low ,

pressure, low temperature operation all automatic Safety Injection actuation signals except Containment Pressure - High are blocked. In normal conditions, '

a single failure of the ESF actuation circuitry will result in the starting of atmostonetrainofSafetyInjection(onecentrifugalchargingpump,andone Safety Injection pump). For temperatures above 325 F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORVs without exceeding Appendix G limit. Given the short time duration that this condition is allowed and the low probability of a single failure of a PORV is not assumed. Initiation of both trains of Safety Injection during this 4-hour time frame due to operator error or a single fail- l ure occurring during testing of a redundant channel are not considered to be l credible accidents.

OperationwithallcentrifugalchargingpumpsandbothSafetyInjection pumps OPERABLE is acceptable when RCS temperature is greater than 350 F, a single PORV has sufficient capacity to relieve the combined flow rate of all SEABROOK - UNIT 1 B 3/4 4-15 g 2 5 $B6

9 ,

3/4.5 EMERGENCY CORE COOLING SYSTEMS 8 85 mrE Ul BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration, and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

The accumulator power-operated isolation valves are considered to be

" operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single-failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the ever.t of a LOCA assuming the loss of one subsystem through any single-failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of

' supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold-leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumps except the required OPERABLE charging pump to be in-operable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on pro-vides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or RHR suction relief valve.

SEABROOK - UNIT 1 B 3/4 5-1 2 85IMi w

l w- c CONTAINMENT SYSTEMS ,

BASES 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS E4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment 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 safety analyses.

The two independent Containment Spray Systems provide post-accident cool-ing of the containment atmosphere. The Containment Spray Systems also provide a mechanism for removing iodine from the containment atmosphere, and, therefore, e the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained consistent with those assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient Na0H is added to the containment spray in the event of a LOCA. The limits on Na0H volume and concentration ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are con-sistent with the iodine removal efficiency assumed in the safety analyses.

3/4.6.3 CONTAINMENT ISOLATION VALVES 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 atmosphere or pressurization of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of Appendix A to 10 CFR Part 50.

Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of con-trolling the expected hydrogen generation associated with: (1) zirconium-water SEABROOK - UNIT 1 B 3/4 6-3 Jimt51m

3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1320 psia) of its design pressure of 1200 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accor-dance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is 1.839 x 107 lbs/hr which is 121% of the total secondary steam flow of 1.514 x 107 lbs/hr at 100% RATED THERMAL POWER. A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions are derived on the following bases:

For four loop operation:

3p = (X) - (Y)(V) x 109 X

Where:

SP = Reduced Reactor Trip Setpoint in percent of RATED THEP. MAL POWER, .

V = Maximum number of inoperable safety valves per steam line, i

109 = Power Range Neutron Flux-High Trip Setpoint for four loop operation, X = Total relieving capacity of all safety valves per steam line in lbs/hr, and Y = Maximum relieving capacity of any one safety valve in Ibs/hr SEABROOK - UNIT 1 B 3/4 7-1 M'E5 %

PLANT SYSTEMS BASES TURBINE CYCLE 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss-of-of fsite power.

The electric motor-driven emergency feedwater pump is capable of deliver-ing a total feedwater flow of 650 gpm at a pressure of 1185 psig to the en-trance of the steam generators. The steam-driven emergency feedwater pump is capable of delivering a total feedwater flow of 650 gpm at a pressure of 1460 psig to the entrance of the steam generators. The startup feedwater pump serves as the third auxiliary feedwater pump and can be manually aligned to be powered from an emergency bus (Bus 5). The startup feedwater pump is capable of taking suction on the dedicated emergency feedwater volume of water in the condensate storage tank and delivering a total feedwater flow of in excess of 650 gpm at a pressure of 1185 psia to the entrance of the steam generator via either the main feedwater header or with manual alignment to the emergency feedwater flow path. This capacity is sufficient to ensure that adequate feed-water flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the Residual Heat Removal System may be placed into operation.

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water vol-ume ensures that sufficient water is available to cool the RCS to a temperature of 350 F. The OPERABILITY of the concrete enclosure ensures this availability of water following rupture of the condensate storage tank by a tornado generated missile. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 1 gpm reactor-to-secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.

SEABROOK - UNIT 1 B 3/4 7-2 M 2 51986 .

, e

1 RADI0 ACTIVE EFFLUENTS BASES .

LIQUID EFFLUENTS 3/4.11.1.2 DOSE (Continued)

LCOs, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

]

3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM I The OPERABILITY of.the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate por _

tions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reason-ably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design ob-jective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits gcverning the use of appropriate portions of the Liquid Radwaste Treat-ment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR Part 50 for liquid effluents.

3/4.11.1.4 LIQUID HOLDUP TANKS The temporary tanks include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contair.ed in the speci-fied tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations would be less than the limits '

of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

3/4.11.2 GASE0US EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column I. These limits provide reasonable as-surance that radioactive material discharged in gaseous effluents will not re-sult in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE B0UNDARY, to annual average concentrations ex-ceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE B0UNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be l

l SEABROOK - UNIT 1 B 3/4 11-2 1

Jint 2 5 886 '

l l

RADI0 ACTIVE EFFLUENTS BASES GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE (Continued) sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin.

These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year.

3/4.11.2.2 DOSE - NOBLE GASES This specification is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I to 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section I.B of Appendix 1.

The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I at the SITE B0UNDARY that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcula-tional procedures based on models and data such that the actual exposure of a MEMBEP. OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The dose calculation methodology and parameters estab-lished in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the method-ology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Com-pliance with 10 CFR Part 50, Appendix I," Revision I, October 1977 and Regu-latory Guide 1.111, " Methods for Estimating Atmospheric Transport and Disper-sion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors,"

Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE B0UNDARY are based upon the historical average atmospheric. conditions.

3/4.11.2.3 00SE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to implement the requirements of Sections II.C, III. A and IV. A of Appendix I to 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure SEABROOK - UNIT 1 B 3/4 11-3 JUN t5 8

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flNatMFT SECTION 5.0 DESIGN FEATURES i

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DESIGN FEATURES 5.2. CONTAINMENT (Continued)

f. Nominal thickness of steel liner = 1/4, 3/8, and 1/2 inch for the floor, wall, and dome, respectively.

g. Net free volume = 2.704 X 108 cubic feet.

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 52.0 psig and a temperature of 296 F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core she.11 contain 193 fuel assemblies with each fuel assembly con-taining 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar in phy-sical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235.

CONTROL R0D ASSEMBLIES 5.3.2 The core shall contain 57 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The ncminal values of absorber material shall be 80% silver, 15% in-dium, and 5% cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650 F, except for the pressurizer which is 680 F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,265 cubic feet at a nominal T of 588.5 F.

avg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

SEABROOK - UNIT 1 5-5 g w,

DESIGN FEATURES rdd Os.kmMI 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Ak eff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance of 1.5% Ak/k for uncertainties as described in Section 4.3 of the FSAR, and

b. A nominal 10.35 inch center-to-center distance between fuel assemblies placed in the storage racks.

5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 14 feet 6 inches.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1236 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

1 SEABROOK - UNIT 1 5-6 Jt1N 2 51996

FINAL DR!R PSNH PESIDENT

& CHIEF EXECUTIVE OFFICDt f64 YAMGI PRESIDENT & CEO SENIOR VICE PRESIDENT VICE PRESIDENT VICE PESIDENT WM APO DIRECTOR OF PRODUCTION OUAllTY PROGRAMS STATICW HANAGER my GUALITY MANAGER D T E DIRECTOR OF DIRECTIR CF SERVICES MONT. CONTROL ENGDEERING FIGURE 6.2-1 0FFSITE ORGANIZATION SEABROOK - UNIT 1 6-2 JUN E 5 &

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k- r ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine station operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of station design and operating experience information, including units of similar design, which may indicate areas for improving station safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving station safety to the Executive Assistant to the Senior Vice President.

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of station activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

RECORDS 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar month to the Executive Assistant to the i Senior Vice President.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the control room commander in the areas of thermal hydraulics, reactor engi-neering, and plant analysis with regard to the safe operation of the station.

6.3 TRAINING 6.3.1 A retraining and replacement licensed training program for the station staff shall be maintained under the direction of the Training Center Manager -

and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix A of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the NRC letter dated March 28, 1980 to all licensees, and shall include familiarization with relevant industry operational experience.

Not responsible for sign-off function. -

l SEABROOK - UNIT 1 6-5 JUN 25 5

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RESPONSIBILITIES 6.4.1.7 (Continued) and the Station Manager however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS 6.4.1.8 The SORC shall maintain written minutes of each SORC meeting that, at a minimum, document the results of all 50RC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President-Nuclear Production and the NSARC.

6.4.2 NUCLEAR SAFETY AUDIT REVIEW COMMITTEE (NSARC)

FUNCTION 6.4.2.1 The NSARC shall function to provide independent review and audit of designated activities in the areas of:

a. Nucletr power plant operations,

b. Nuclear engineering,

c. Chemistry and radiochemistry,

d. Metallurgy,

e. Instrumentation and control,

f. Radiclogical safety,

g. Mechanical and electrical engineering, and

h. Quality assurance practices.

The NSARC shall report to and advise the Senior Vice President on those areas of responsibility specified in Specifications 6.4.2.7 and 6.4.2.8.

COMPOSITION 6.4.2.2 The NSARC shall be composed of at least five (5) individuals. The Chairman, Vice Chairman and members, including designated alternates, shall be appointed in_ writing by the Senior Vice President. Collectively, the individuals appointed to the NSARC should be competent to conduct reviews identified by Specification 6.4.2.1. Each member shall meet the qualifica-tions of ANSI 3.1-1978, Section 4.7.

ALTERNATES 6.4.2.3 All alternate members shall be appointed in writing by the Senior Vice President to serve on a temporary basis; however, no more than a minority shall participate as voting members in NSARC activities at any one time.

CONSULTANTS l 6.4.2.4 Consultants shall be utilized as determined by the NSARC to provide expert advice to the NSARC.

l l SEABROOK - UNIT 1 6-8 g 25 %

ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.4.2.5 The NSARC shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and thereafter at least once per 6 months i 6 weeks.

QUORUM 6.4.2.6 The quorum of the NSARC necessary for the performance of the NSARC review and audit functions of these Technical Specifications shall consist of the Chairman or Vice-Chairman and at least four NSARC members including alter-nates. No more than a minority of the quorum shall have line responsibility for operation of the unit. The Vice Chairman, or his designated alternate, can participate as an NSARC member when the Chairman is in attendance.

REVIEW 6.4.2.7 The NSARC shall be responsible for the review of:

a. The safety evaluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that such actions did not constitute an unreviewed safety question;

b. Proposed changes to procedures, equipment, or systems that involve an unreviewed safety question as defined in 10 CFR 50.59;

c. Proposed tests or experiments that an unreviewed safety question as defined in 10 CFR 50.59;

d. Proposed changes to Technical Specifications or this Operating License;

e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;

f. Significant operating abnormalities or deviations from normal and expected performance of station equipment that affect nuclear safety;

g. All REPORTABLE EVENTS; l h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and

i. Reports and meeting minutes of the SORC.

1 AUDITS 6.4.2.8 Audits of station activities shall be performed under the cognizance of the NSARC. The audits shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified interval SEABROOK - UNIT 1 6-9 JUN 2 5 kW

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ADMINISTRATIVE CONTROLS fc"" . g RECORDS 6.4.2.9 Records of NSARC activities shall be prepared and distributed as indicated below:

a. Minutes of each NSARC meeting shall be prepared and forwarded to the Senior Vice President within 14 days following each meeting;

b. Reports of reviews encompassed by Specification 6.4.2.7 shall be included in the minutes where applicable or forwarded under sepa-rate cover to the Senior Vice President within 14 days following completion of the review; and

c. Audit reports encompassed by Specification 6.4.2.8 shall be forwarded to the Senior Vice President and to the management positions respons-ible for the areas audited within 30 days after completion of the audit by the auditing organization.

6.5 REPORTABLE EVENT ACTION The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the 50RC and the results of this review shall be submitted to the NSARC and the Vice President-Nuclear Production.

6.6 SAFETY LIMIT VIOLATION The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Vice President-Nuclear Production and the NSARC shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the 50RC. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence;

c. The Safety Limit Violation Report shall be submitted to the Commission, the NSARC, and the Vice President-Nuclear Production within 14 days of the violation; and

d. Operation of the station'shall not be resumed until authorized by the Commission.

SEABROOK - UNIT 1 6-11 JUN 15 W

b b r ADMINISTRATIVE CONTROLS i l 6.7 PROCEDURES AND PROGRAMS 6.7.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;

b. The emergency operating procedures required to implement the require-ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33;

c. Security Plan implementation, '

d. Emergency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation; l

f. OFFSITE DOSE CALCULATION MANUAL implementation;

g. Quality Assurance Program for effluent and environmental monitoring;

h. Fire Protection Program implementation; and

i. Technical Specification Improvement Program implementation.

6.7.2 Each procedure of Specification 6.7.1, and changes thereto, shall be reviewed by the 50RC and shall be approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative procedures.

6.7.3 Temporary changes to procedures of Specification 6.7.1 may be made pro-vided:

a. The intent of the original procedure is not altered;

b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and

c. The change is documented, reviewed by the 50RC, and approved by the Station Manager within 14 days of implementation.

6.7.4 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the RHR and containment spray, Safety Injection, chemical and volume control. The program shall include the following:

SEABROOK - UNIT 1 6-12 JUN 2 51986

r l

ADMlNISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **

6.8.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the station during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Semiannual Radioactive Effluent Release Reports shall include a

, summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the station as outlined in Regulatory Guide 1.21,

" Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

For solid wastes, the format for Table 3 in Appendix B shall be supplemented a with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year **. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured),

or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-3) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological I conditions concurrent with the time of release of radicactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the 0FFSITE DOSE CALCULATION MANUAL (0DCM).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year "In lieu of submission with the Semiannual Radioactive Effluent' Release l Report, the licensee has tte option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

- The dose calculations may be reported in a supplement submitted 30 days later.

SEABROOK - UNIT 1 6-17 JW 2 515

r ADMINISTRATIVE CONTROLS =

PROCESS CONTROL PROGRAM (PCP) 6.12.2 (Continued)

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC.

b. Shall become effective upon review and acceptance by the 50RC.

6.13 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.13.1 The ODCM shall be approved by the Commission prior to implementation.

6.13.2 Licensee-initiated changes to the ODCM:

a. Changes to Part A shall be submitted to and approved by the NRC staff prior to implementation.

b. Changes to Part B shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain:

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s);

2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC. .

c. Changes to Part B shall become effective upon review and acceptance by the 50RC.

SEABROOK - UNIT 1 6-22 JUN 2 51986 m --

ML20199K047

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