ML20155G880

Forwards Addl Comments on Draft Proof & Review Tech Specs Transmitted by NRC .Justification for Changes Included W/Comments

ML20155G880
Person / Time
Site:	Seabrook
Issue date:	05/02/1986
From:	George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	Noonan V Office of Nuclear Reactor Regulation
References
SBN-1026, NUDOCS 8605070179
Download: ML20155G880 (12)

Text

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George S. Thomas vice Prescent Nx!*cr Pfcductio^

Pub 5c Service of New HampeNro N:w Hampshire Yankee Division May 2, 1986 SBN-1026 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, DC 20555 Attention:

Mr. Vincent S. Noonan, Project Director PWR Project Directorate No. 5

References:

(a) Construction Permit CPPR-135 and CPPR-136, Docket Nos.

50-443 and 50-444.

(b) USNRC Letter dated March 13, 1986, "Seabrook Technical Specifications," V. S. Noonan to R. J. Harrison Subj ect : Seabrook Station Proof and Review Technical Specifications

Dear Sir:

Enclosed please find additional comments on the Seabrook Station Proof and Review Technical Specifications provided by the Staff in Reference (b).

Justifications for these changes are included with the comments.

Should you have any questions, please contact Mr. Warren J. Hall at (603) 474-9574, extension 4046.

Very truly yours, C

Georg Thomas GST/cj b Enclosure cc: ASLB Service List 8605070179 860502 PDR ADOCK 05000443 A

PDR

)\\

P.O. Box 300 Seabrook,NH03874 Telephone (603)474-9521 s

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REACTIVITY CONTROL SYSTEMS

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FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a.

The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS), and b.

Two flow paths from the refueling water storage tank via charging pumps to the RCS.

APPLICABILITY: MODES 1, 2 and 3" ACTION:

With only one of the above required boren injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least ak/k at 200'F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow path to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

l3?o SURVEILLANCE REOUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; Q 4.--- At-least once-.per-18-monthrduring-shutdown-by-verifying-that-each-.

-automatic valve. in the-flow-path. actuates to-its-correct-position-on-

_a_. safety injection-test signal;-and c.

At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

• 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

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SEABROOK - UNIT 1 3/4 1-8 MAY O21986 Li MAR:131886

o REACTIVITY CONTROL 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.*

t' ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUT 00WN MARGIN equivalent to at least %% ak/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within tt e next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1

I37, SURVEILLANCE REOUIREMENTS 4.1.2.4.1 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 2,495 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or prior to the temperature of one or more of the RCS cold legs exceeding 375 F, whichever comes first.

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SEABROOK -

1 3/4 1-10 i

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MAY O2 uan liiUR

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-6 REACTIVITY CONTROL SYSTEMS

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BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 20,200 gallons, 2

2)

A minimum baron concentration of 7000 ppm, and 3)

A minimum solution temperature of 65*F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 479,000 gallons, 2)

A minimum baron concentration of 2000 ppm, 3)

A minimum solution temperature of 50*F, and 4)

A maximum solution temperature of 86*F.

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

ACTION:

a.

With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restor'e the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least' Ak/k at 200*F; restore the Boric Acid Storage System to OPERABL status within the next 7 days or be in COLD SHUTDOWN within the ext 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

/ 7c, b.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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e SEABROOK - UNIT 1 3/4 1-12 MAY 021986 MAR 131986

d POWER DISTRIBUTI6N LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued,,),

With the indicated AFD outside of the above required target band for c.

more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER.satii-4he=inetceted M0 i: MtM.- th; ete.e.

i ed 0 wt i: d.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

Monitoring the indicated AFD for each OPERABLE excore channel at least a.

once per 7 days when the AFD Monitor Alarm is OPERABLE, and b.

Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

One minute penalty deviation for each 1 minute of POWER OPERATION a.

outside of the target band at THERMAL POWER levels equal to or above l

50% of RATED THERMAL POWER, and I'

b.

One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value andr05-et-the end# N cycle !Me. The provi-sions of Specification 4.0.4 are n W app uca M1).un d M g

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.n Illit SEABROOK - UNIT 1 3/4 2-2 l

MAY 021986 M O l886

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHAN.

7CTOR LIMITIN'G CONDITION FOR OPERATION 3.2.3 F shall be less than 1.49 [1.0 + 0.2 (1-P)].

H gem po ne M h er APPLICABILITY: MODE 1.

ACTION:

N With F e ceeding its limit.

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> reduce the THERMAL POWER to the level where the LIMITING CONDITION FOR OPERATION is satisfied.

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the limit required by ACTION a.,

above; THERMAL POWER may then be increased provided F is demonstrated through incore mapping to be within its limit.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

Fhshallbedemonstratedtobewithinitslimitpriortooperation 4.2.3.2 above 75% RATED THERMAL POWER after each fuel loading and at least once per 31 EFPD thereafter by:

a.

Using the moveable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% RATED THERMAL POWER.

b.

Using the measured valve of F which does not include an allowance H

for measurement uncertainty.

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SEABROOK - UNIT 1 3/4 2-8 p.

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6-MAY O 21986 C Q R11 f t9 81l -

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TABLE 4.3-1 T,M REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ANALOG ACTUATING MODES FOR

_,e CilANNEL DEVICE WillCH

$E CilANNEL CllANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE

-i FUNCTIONAL UNIT CifECK CALIBRATION TEST 1EST LOGIC TEST IS REQUIRED w

1.

Manual Reactor Trip N.A.

N.A.

N.A.

R(14)

N.A.

1, 2, 3 *, 4 *, 5

• 2.

Power Range, Neutron Flux a.

liigh Setpoint S

D(2, 4),

M N.A.

N.A.

1, 2 M(3, 4),

Q(4, 6),n x %

R(4, 5) b.

Low Setpoint S

R(4)

H N.A.

N.A.

1***,

2

{

3.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 liigh Positive Rate U

4.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 liigh Negative Rate S.

Intermediate Ran0e, S

R(4, 5)

S/U(1)

N.A.

N.A.

1***,

2 Neutron Flux 6.

Source Range, Neutron Flux 5

R(4, 5)

S/U(1),Q(9,17)

N.A.

N.A.

2**,

3, 4, 5 7.

Overtemperature AT S

R(12)

M N.A.

N.A.

1, 2

,-/

8.

Overpower AT S

R H

N.A.

N.A.

1, 2 9.

Pressurizer Pressure--Low 5

R H

N.A.

N.A.

I

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e" 10.

Pressurizer Pressure--liigh S

R H

N.A.

N.A.

1, 2 hI'"lJ.

11.

Pressurizer Water Level--liigh S

R H

N.A.

N.A.

I h

_._a 12.

Reactor Coolant Flow--Low 5

R H

N.A.

N.A.

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-o TABLE 4.3-1 (Continued)

TABLE NOTATIONS

• 0nly if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.

• • Below P-6 (Intermediate Range Neutron F1ux Interlock) Setpoint.

• • • Eelow P-10 (Low Set oint Power Range Neutron Flux Interlock) Setpoint.

-nfrhy, 16 %

tulu km. deme L

1. e 9 2 EFPD AM M (1)

f. performed in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX D'IFFERENCE above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Initial plateau curves shall be measured for each detector. Subsequent plateau curves shall be obtained, evaluated and compared to the initial curves. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

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

(8) With power greater than or equal to the Interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive annun-ciator window.

(9) Surveillance in MODES 3*,

4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip attachments of the Reactor M Trip Breakers.

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EMERGENCY CORE COOLING SYSTEMS

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SURVEILLANCE REQUIREMENTS (Continued) d.

At least once per 18 months by:

1)

Verifying automatic isolation and interlock action of the RHR 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 prevent the valves from being opened, and

_mb With_ a simulated or actual Reactor Coolant System pressure ar en,tu-signaN4sse 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.

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

e.

1)

Verifying tha't 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.ested pursuant to Specification 4.0.5:

1)

Centrifugal charging pump 1 2480 psid, 2)

Safety Injection pump 1 1445 psid, and 3)

RHR pump

> 183 psid.

g.

By verifying the correct position of each electrical and/or mechanical position stop for the following ECCS throttle valves:

1)

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and

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SEABROOK - UNIT 1 3/4 5-6 l

MAY 021986 MAR ll1886 -

6

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Table 3.7-l3 Area Temperature Monitoring Area Temperature Limit (*F) 1.

Control Room 85 2.

Cable Spreading Room

• 99.5 e

3.

Switchgear Room - Train A*

99.5 4.

Switchgear Room - Train B*

99.5 5.

Battery Rooms - Train A 90.5 6.

Battery Rooms - Train B 90.5 7.

ECCS Equipment Vault - Train A 99.5 8.

ECCS Equipment Vault - Train B 99.5 9.

Centrifugal Charging Pump Room - Train A 99.5 10.

Centrifugal Charging Pump Room - Train B 99.5 11.

ECCS Equipment Vault Stairwell - Train A TT -[-

12.

ECCS Equip =ent Vault Stairwell - Train B 910 13.

PCCW Pump Area 99.5 14.

Cooling Tower Switchgear Room - Train A*

99.5 15.

Cooling Tower Switchgear Room - Train B*

99.5 16.

Cooling Tower SW Pump Area

• 122.5 17.

SW Pumphouse Electrical Room - Train A*

99.5 18.

SW Pumphouse Electrical Room - Train B*

99.5 19.

Sw Pu=p Area *

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99.5 20.

Diesel Generator Room - Train A*

115.5 21.

Diesel Generator Room - Train B*

115.5 22.

EFW Pumphouse 99.5 23.

Electrical Penetration Area - Train A 93.5 24.

Electrical Penetration Area - Train B 80.5 25.

Fuel Storage Building Spent Fuel Pool 99.5 Cooling Pump Area 26.

Main Steam and Feedwater Pipe Chase - East 125.5 27.

Main Stean and Feedwater Pipe Chase - West 125.5

• Mild Environment Area htAY 021986 waif) g 9 33

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4.

Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight 17f,8 h grams uranium.

The initial core loading shall have a maximum enrichment of 3.1 weight percent U-235.

Reload fue!.shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

CONTR00 ROD 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 nominal values of absorber material shall be 80%

silver, 15% indium, 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 For a temperature of 650 F, except for the pressurizer which is c.

680 F.

VOLUME 12o.2LS~

5.4.2 The total water and steam volume of the Reactor Coolant System is -12;350 of 588 F.

cubic feet at a nominal T,yg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

MAR 131986 MAY 021986 i

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SEABROOK - UNIT 1 5-5

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f JUSTIFICATIONS 1.

Specification 4.1.2.2.b - This surveillance requirement can be deleted since it is identical to surveillance requirement 4.5.2.e.1 on page 3/4 5-6.

2.

Specification 3.1.2.2, 3.1.2.4, 3.1.2.6 (Action) - The change from 1%

AK/K to 1.3% Ac/c is made to comply with the boron dilution event analysis for Seabrook Station.

3.

Specification 3.2.1 (Action C) - This deletion is made because it is in conflict with the LCO.

4.

Specification 4.2.1.4 - This change is made because not all plots will predict 0% at end of life cycle.

5.

Specification 3.2.3 - This change is made to show what values are used to determine the value "P".

6.

Design Features - Section 5.3.1 - The value for the weight of uranium was finalized as marked value of 1768.

The value of 5.0 weight percent U-235 is used since the final determination of maximum enrichment for Seabrook reload fuel has not yet been determined.

7.

Design Feature - Section 5.4.2 - This number conforms to the value used in the Westinghouse analysis for Seabrook Station.

8.

Table 4.3-1 (pages 3/4 3-10 and 3/4 3-13) - This change is made to clarify that quarterly testing will be done based upon 92 EFPD rather than 92 calendar days.

9.

Specification 4.5.2.d.l.b - This change is made to correct the wording so that the surveillance will be performed to assure proper functioning of the valves.

10.

Table 3.7 This change is made to add temperature limits for two areas inadvertently omitted from our previous submittal.