ML20148U463

Amends 173 & 177 to Licenses DPR-24 & DPR-27,respectively, Revising Listed TSs & Definitions for Both Units to Incorporate Listed Changes to Facility Into TS & FSAR

ML20148U463
Person / Time
Site:	Point Beach
Issue date:	07/01/1997
From:	Gundrum L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20148U467	List: ML20148U463 ML20148U481
References
NUDOCS 9707100187
Download: ML20148U463 (26)

Text

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i UNITED STATES j

NUCLEAR REGULATORY COMMISSION 2

WA&MINGTON, D.C. 20eeNmo1 o

%7*.@..i}

WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 173 License No. DPR-24 1.

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

A.

The applications for amendment (TSCR 188 and 189) by Wisconsin Electric Power Company (the licensee) dated June 4.1996, as supplemented August S. September 26. October 21. November 13.

November 20. and December 2. 1996, and January 16. March 20. and April 2.1997, comply with the standards and requirements of the Atomic Energy Act of 1954. as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I:

B.

The facility will operate in conformity with the applications. the provisions of the Act, and the rules and regulations of the Commission:

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9707100187 970701 DR ADOCK 05000266 PDR m

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2.

Accordingly. the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.

I and paragraph 3.B of Facility Operating License No. DPR-24 is hereby amended to read as follows:

B.

Technical SDecifications l

The Technical Specifications contained in A3pendices A and B. as revised through Amendment No. 173 are here)y incorporated in the license. The licensee shall operate the facility.in accordance with the Technical Specifications.

3.

This license amendment is effective immediately u)on issuance. The Technical Specifications shall be implemented wit 11n 45 days from the l

l date of issuance and the Final Safety Analysis Report changes shall be 1mplemented by June 30, 1998.

Implementation of this amendment includes incorporation of accident analyses submitted in support of this l

amendment into the Final Safety Analysis Report in sufficient detail to l

support future evaluations performed in accordance with 10 CFR 50.59 and as described in the licensee's applications dated June 4.1996 as supplemented on August 5. September 26. October 21. November 13.

November 20. and December 2. 1996 and January 16. March 20. and April 2. 1997 and evaluated in the staff's safety evaluation dated July 1. 1997.

J FOR THE NUCLEAR REGULATORY COMMISSION l

l u$t

>1d t i e s w a

Linda L. Gundrum. Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation I

Attachment:

Changes to the Technical Specifications Date of issuance:

July 1. 1997 l

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't UNITED STATES 4

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NUCLEAR REGULATORY COMMISSION 2

WASHINGTON. D.C. 20066 4001

%,n @T /

+....

l WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT. UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 177 License No. DPR-27 1.

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

A.

'The applications for amendment (TSCR 188 and 189) by Wisconsin Electric Power Company (the licensee) dated June 4. 1996, as supplemented August 5. September 26. October 21. November 13.

November 20. and December 2. 1996, and January 16. March 20. and 1

April 2.1997, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I:

B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission:

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's regulations:

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public:

and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly. the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.

and paragraph 3.B of Facility Operating License No. DPR-27 is hereby l

amended to read as follows:

B.

Technical Soecifications l

The. Technical Specifications contained in A)pendices A and B. as l

revised through Amendment No. 177, are here)y incorporated in the i

license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective immediately u)on issuance. The Technical Specifications shall be implemented witlin 45 days from the date of issuance and the Final Safety Analysis Report changes shall be implemented by June 30. 1998.

Implementation of this amendment includes.

incorporation of accident analyses submitted in support of this amendment into the Final Safety Analysis Report in sufficient detail to support future evaluations performed in accordance with 10 CFR 50.59 and as described in the licensee *s applications dated June 4. 1996, as i

supplemented on August 5. September 26. October 21. November 13.

November 20. and December 2. 1996, and January 16. March 20. and April 2.1997, and evaluated in the staff's safety evaluation dated July 1. 1997.

FOR'THE NUCLEAR REGULATORY COMMISSION hn:t'n k

, ul u > ~

t Linda L. Gundrum. Project Manager Project Directorate 111-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical

{

Specifications Date of issuance: July 1. 1997 L

ATTACHMENT TO LICENSE AMENDMENT NOS. 173 AND 177 TO FACILITY OPERATING LICENSE NOS. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT 15.1-6 15.1-6 15.2.1-1 15.2.1-1 Figure 15.2.1-1 Figure 15.2.1-1 Figure 15.2.1-2

. Figure 15.2.1-2 15.2.2-1 15.2.2-1 15.2.3-1 15.2.3-1 15.2.3-2 15.2.3-2 15.2.3-3 15.2.3-3 15.2.3-3a 15.2.3-3a-15.2.3-6 15.2.3-6 15.3.1-9 15.3.1-9 15.3.1-10 15.3.1-10 15.3.1 15.3.1-19 Figure 15.3.1-5 Figure 15.3.1-5 15.3.4-2 15.3.4-2 15.3.4-3 15.3.4-3 Table 15.3.5-1 Table 15.3.5-1 (Page 1 of 2)

'(Page 1 of 2)

Table 15.4.1-2 Table 15.4.1-2 (Page 1 of 5)

(Page 1 of 5)

(Page 2 of 5)

(Page 2 of 5) 15.5.3-2A 15.5.3-3 15.6.9-3 15.6.9-3 I

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o.

Dose Equivalent I-131 Dose Equivalent 1-131 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, 1-133. I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listeG in Table 2.1 of Federal Guidance Report No. 11. " Limiting Values of Radionuclide Intake and Air Concentration and I

Dose Conversion Factors for Inhalation. Submersion, and Ingestion,"

i September 1988.

p.

E-AverageDisintegrationEnergy l

I E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

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8 Unit 1 - Amendment No. M. W.173 Unit 2 - Amendment No. M. W.177 15.1-6

15.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT. REACTOR CORE Acolicability:

Applies to the limiting combinations of thermal power. reactor coolant system pressure, and coolant temperature during operation.

Obiective:

To maintain the integrity of the fuel cladding.

Soecification:

1.

The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 for Units 1 and 2r The safety limit is exceeded if the point defined by the combi'.n on of reactor coolant system average temperature and power leval is at any time above the appropriate pressure line.

Basis:

The restrictions of this safety limit prevent overheating of the fuel and pos.eible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excess cladding temperature because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not 6 directly measurable parameter during operation and therefore thermal power and Reactor Coolant temperature and pressure have been related to DNB.

Figure 15.2.1-1 applies :,n Unit 2 following U2R22 and to Unit 1 following U1R24.

Prior to U1R24. Figure 15.2.1-2 applies to Unit 1.

Unit 1 - Amendment No. 443, 173 Unit 2 - Amendment No. 446. 177 15.2.1-1

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Figure 15.2.1-1 POINT BEACH NUCLEAR PLANT UNITS 1 AND 2*

REACTOR CORE SAFETY LIMITS 670 660.-

650

,425 psia 640 n50 piia q

88 630 G

3 ca

$620 d

2000 psia j

e 3 610 5

600 m5p

\\

590 j

580 570 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

1.1 1.2 Core Power (fraction of 1518.5 MWt)

This figure applies to Unit 2 following U2R22 and to Unit 1 following U1R24.

Prior to U1R24. Figure 15.2.1-2 applies to Unit 1.

Unit 1 - Amendment No.

86. 440. 442, 173 Unit 2 - Amendment No. 446, 460. 469. 177

Figure 15.2.1-2*

REACTOR CORE SAFETY LIMITS POINT BEACH UNIT 1 i

.i s

556" -

0 SSE' 2400 P$!A i

1 j

i gag.

2250 PSIA i

t t

552-4 2000 PS!A

,$28-e 4

w i

1775 PSIA 512<

1 i

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lat.

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589<

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1

=

e.

.I

.2

.5 4

.5

.6 7

.t 9

1 1.1 1.2 P0vtt irrectica or acaiaell This figure applies to Unit 1 prior to U1R24.

Following U1R24. Figure 15.2.1-1 applies to Unit 1.

I Unit 1 - Amendment No.

86. MG. 443,173 2

Unit 2 - Amendment No. 446. MG. M9.177

15.2.2 SAFETY LIMIT. REACTOR COOLANT SYSTEM PRESSURE Acolicability l

Applies to the maximum limit on Reactor Coolant System Pressure.

Obiective To maintain the integrity of the Reactor Coolant System.

Soecification The Reactor Coolant System pressure shall not exceed 2735 psig with fuel assemblies installed in the reactor vessel.

BLill The Reactor Coolant System") serves as a barrier preventing radionuclides contained in the reactor coolant from reaching the atmosphere.

In the event of a fuel cladding failure the Reactor Coolant System is the primary barrier against the release of fission products.

By establishing a system pressure limit, the continued integrity of the Reactor Coolant System -is assured.

The maximum transient pressure allowable in the Reactor Coolant System pressure vessel under the ASME Code.Section III is 110% of design pressure. The maximum transient pressure allowable in the Reactor Coolant System piping. valves and fittings under USAS Section B31.1 is 120% of design pressure.

Thus, the safety limit of 2735 psig (110% of design pressure) has been established.<2)

The nominal settings of the power-operated relief valves.(2335 psig). the reactor high-pressure trip and the safety valves (2485 psig) have been l

established to assure never reaching the Reactor Coolant System pressure safety limit except for the hypothetical locked rotor and rod ejection accidents which use the faulted condition stress limit acceptance criterion of 3105 psig (3120 l

psia). The initial hydrostatic test was conducted at 3110 psig to assure the integrity of the Reactor Coolant System.

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Reference (1)

FSAR. Section 4 (2)

FSAR. Section 4.3 Unit 1 - Amendment No. 173 Unit 2 - Amendment No. 177 15.2.2-1

15.2.3.

LIMITING SAFETY SYSTEM SETTINGS. PROTECTIVE INSTRUMENTATION Acolicability:

Applies to trip settings for instruments monitoring reactor power and reactor coolant. pressure temperature flow. pressurizer level, and permissives related to reactor protection.

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Obiective:

To provide for automatic protective action in the event that the principal process variables approach a safety limit.

Soe:ification:

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1.

Protective instrumentation for reactor trip settings shall be as follows:

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A.

Startup protection (1)

High flux, source range - within span of source range l

instrumentation.

(2)

High flux, intermediate range - s40% of rated power.

l (3)

'iigh flux power range (low setpoint ) - s25% of rated power.

B.

Core limit protection (1)

High flux, power range (high setpoint) - s108% of rated power

-(2)

High pressurizer pressure * - 52385 psig for operation at 2250 psia primary system pressure 52210 psig for operation at 2000 psia primary system pressure.

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These values apply to Unit 2 following U2R22 and to Unit 1 following j

i U1R24.

Prior to U1R24 the high pressurizer pressure reactor trip setpoint for Unit 1 is s2385 psig i

Unit 1 - Amendment No. (Chg ').173 Un1*. 2 - Amendment No. 3. 177 15.2.3-1

(3)

Low pressurizer pressureo - a1905 psig for operation at 2250 psia primary system pressure i

21800 psig for operation at 2000 psia primary system pressure (4)

Overtemperature AT ( 1

.)

1+r3S s AT, (K -K (T( 1

)-T' )(1+rj)+K (P-P' )-f(AI))

i 2 3

1+r4S 1+r S 2

where (values are applicable to operation at both 2000 psia and 2250 psia unless otherwise indicated)

AT, indicated AT at rated power. F

=

T average temperature. F

=

T' s 572.9 F**

P pressurizer pressure, psig

=

P' 2235 psig (??50 gia operation only)

=

P' 1985 psig (2000 psia operation only)**

=

K s 1.19 (2250 psia operation only) 1 K

s 1.14 (2000 psia operation only)**

1 K

0.025 (2250 psia operation only)

=

2 K

0.022 (2000 psia operation only)**

2 K

0.0013 (2250 psia operation only)

=

3 0.001 (2000 psia operation only)**

K

=

3 25 sec r

1 3 sec T

=

2 r3 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD

=

7, 2 sec for Rosemont or equivalent RTD

=

0 sec for Sostman or equivalent RTD

=

and f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startu) tests, where qt and q are the the top and bottom 1alves of the core resp,ectively, percent power in and qt + q3 is total core power in percent of rated power such that:

(a) for qt - q, within -17. +5 percent, f(AI) = 0.

(b) for each percent that the magnitude of qt - q, exceeds +5 3ercent, the AT trip setpoint shall be automatically reduced

]y ;

equivalent of 2.0 percent of rated power.

These values apply to Unit 2 following U2R22 and to Unit 1 following U1R24.

Prior to U1R24 the low pressurizer pressure reactor trip setpoint for Unit 1 is = 1790 psig.

These values apply to Unit 2 following U2R22 and to Unit 1 following U1R24.

Prior to U1R24. the values are:

T' s 573.9 F. P' - 2235 psig.

K - s 1.30. K = 0.0200 and K = 0.000791.

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l Unit 1 - Amendment No, M6.173 l

Unit 2 - Amendment No. M0.177

~ -

l (c) for each percent that the magnitude of qt - q, exceeds 1

-17 percent,'the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power i

i (5)

Overpower AT ( 1

)

1+r S 3

1

~

s AT, [K,-K (_r S

)( 1

)T-K [T(

1

)-T' ))

3 e

6 r S+1 1+r,S 1+ r,S s

i where (values are applicable to operation at both 2000 psia and 2250 psia) i-AT, indicated AT at rated power. F T

average temperature F

T' s 572.9 F*

1 F

K, s

1.09 of rated power

• l K

4, 3

0.0262 for-increasing T l

0.0 for decreasing i

=

K 0.00123 for T = T' 6

0.0 for T < T' l

=

10 sec r

3 r3 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD l

2 sec for Rosemont or equivalent RT0 7

4 0 sec for Sostman or equivalent RTD (6)

Undervoltage - =75 percent of normal voltage l

[

(7)

Indicated reactor coolant flow per loop -

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=90 percent of normal indicated loop flow r

(8)

Reactor coolant pump motor breaker open I

(a)

Low frequency set point =55.0 HZ l

l (b)

Low voltage' set point = 75 percent of normal voltage.

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These values apply to Unit 2 following U2R22 and to Unit 1 following U1R24.

Prior to U1R24, the values for Unit I are:

T' s 573.9 F and K, 51.09 of rated power.

t Unit 1 - Amendment No. 443. 173 Unit'2 - Amendment No. 446. 177 15.2.3-3

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t Other reactor trips:

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(1)

- High pressurizer water level - 595% of span I

(2)

Low-low steam generator water level -

m20% of narrow range instrument span

=5% of narrow range instrument span (Unit 1)*

-(3)

Steam-Feedwater Flow Mismatch Trip - s1.0 x 10' lb/hr i

1 (4)

Turbine Trip (Not a protection circuit) 1 I

(5)

' Safety Injection Signal i

j (6)

Manual Trip 1

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(

)

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This setting limit applies to Unit 1 until the narrow range lower tap is changed to the lower position consistent with Unit 2.

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Unit 1 - Amendment No.173 Unit 2 - Amendment No.177 15.2.3-3a 4

a power distribution, the reactor trip limit, with allowance for errors <2), is i

a' ays below the core safety limit as shown on Figures 15.2.1-1 and 15.2.1-2.

If axial peaks are greater than design as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip limit is 5

automatically reduced""7).

The overpower, overtemperature and pressurizer pressure system setpoints include the effect of reduced system pressure operation (including the effects of fuel 1

4 densification). The setpoints will not exceed the core safety limits as shown i

in Figures 15.2.1-1 and 15.2.1-2.

The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur.

The reactor is prevented from reaching the overpower limit condition by action of the nuclear overpower and overpower AT trips.

The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted.

The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The low

-pressurizer pressure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident"'.

The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or both reactor coolant pumps.

The setpoint specified is consistent with the value used in the accident analysis"'. The low loop flow signal is caused by a condition of less than 90 percent flow as measured by the loop flow

)

instrumentation. The loss of power <;gnal is caused by the reactor coolant pump breaker opening Unit 1 - Amendment No. 443. 173 Unit 2 - Amendment No. 146, 177 15.2.3-6

C.

MAXIMUM COOLANT ACTIVITY Sggification:

The specific activity of the reactor coolant shall be limit to:

1.

Less than or equal, to 0.8 microcurie per gram Dose Equivalent I-131.

l a.

If the specific activity of the reactor coolant is greater than 0.8 l microcuries per gram Dose Equivalent I-131 but within the allowable limit (below and to the left of the line) shown on Figure 15.3.1-5.

operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Reactor coolant sampling shall be in accordance with Table 15.4.1-2.

b.

If the specific activity of the reactor coolant is greater than 0.8 l microcuries per gram Dose Equivalent I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeds the allowable limit (above and to the right of the line) shown on Figure 15.3.1-5 the reactor will be shut down and the average reactor coolant temperature will be less than 500 'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

Less than or equal to 100/E microcuries per gram.

a.

If the specific activity of the reactor coolant is greater than 100/E microcuries per gram, the reactor will be shut down and the average reactor coolant temperature will be less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Reactor coolant sampling shall be in accordance with Table 15.4.1-2.

Basis:

The limitations on the specific activity of the reac"or coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriat'ely

{

small fraction of Part 100 limits following a steam generator tube rupture i

accident in conjunction with an assumed steady state primary-to-secondary steam j

generator leakage rate of 500 gpd in either steam generator. The values for the limits on specific activity represent limits based upon a parametric evaluation l

-by the NRC of typical site locations. These values are conservative for Point i

Beach Nuclear Plant.

i Unit 1 - Amendment No. N.. W. W.173 l.

Unit 2 - Amendment No M. M6. W.177 15.3.1-9

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Continued power operation for limited time periods with the reactor coolant's specific activity greater than 0.8 microcurie / gram Dose Equivalent I-131._ but l

within the allowable limit shown on Figure 15.3.1-5. accommodates possible

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l iodine spiking phenomenon which may occur following changes in thermal power.

' Operation with specific activity levels exceeding 0.8 microcurie / gram Dose l

)

Equivalent I-131 but within the limits shown on Figure 15.3.1-5 increase the i

2-hour thyroid dose at the site boundary by a factor of up to 20 following a j

l postul'ated steam generator tube rupture.

l Reducing T to less than 500 F normally prevents the release of activity m

l_

should a steam generator tube rupture since the saturation pressure of the l

reactor coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive l

specific activity levels in the primary coolant will be detected in sufficient l

time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

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Unit 1 - Amendment No. N. M3. M0.173 Unit 2 - Amendment No. M. M5. MB.177 15.3.1-10 l

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G.

@PBt.A_[10RA_LLIMITATIONS i

The following DNB related parameters shall be traintained within the limits shown during Rated Power operation:

+

1.

T,y shall be maintained *=557 F and sf/3 '.FF.

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2.

Reactor Coolant System (RCS) pressur:

precure shall be maintained:

22205 psig during operat:on at 2250 psia, ar 21955 psig during operation at 2000 psia.

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3.

Reactor Coolant System raw measurc '. Yotal Flow Rate shall be maintained =181.800 gpm.

Basis:

'The reactor-coolant system total flow rate of 181.800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178.000 gpm).

The raw measured flow'is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimetric at the-beginning of each cycle.

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Unit 1 - Amendment No. 46. 173 Jnit 2 - Amendment No. M9.177 15.3.1-19

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a FIGURE 15.3.15

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j 200 g

~~ UNACCEPTABLE

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OPERATION a

4

~~

p E

s

< 150 o

g u.

D uJ g

y I

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o>a: 100 l

g g

z a

j

_____:ACCEPTABL5 -

- ~ ~ " ~ ~

~~

o 8 50 OPERATl_O,N _,

N g

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20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL. POWER 1

DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.8 pCilgm Dose Equivalent 1-131 Unit 1 - Amendment No. 4. 173 l

. Unit 2 - Amendment No. M.177 s

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3.

A minimum of 13.000 gallons of water per operating unit in the condensate storage tanks and an unlimited water supply.from the lake via either leg of the plant Service Water System.

J 4.

System piping and valves required to function during accident conditions directly associated with the above components operable.

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5.

Both atmosph'eric steam dump lines shall be operable.

If either of the atmospheric steam dump lines is determined to be inoperable.

)-

restore the inoperable line to an operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I If operability cannot be restored, be in hot shutdown within six hours and cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

j B.

The dose equivalent I-131 activity on the secondary side of the steam generator shall not exceed 1.0 pC1/g.

C.

During power operation the requirements of 15.3.4.A.2.a and.b may be g

modified to allow the following components to be inoperable'for a i

specified time.

If the system is not restored to meet the requirements of i

15.3.4. A.2.a and b within the time period specified, the specified action must be taken.

If the requirements of 15.3.4.A.2.a and b are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the appropriate reactor (s) shall be cooled down to less than 350 F.

1.

Two Unit Operation - One of the four operable auxiliary feedwater pumps may be out-of-service for the below specified times. A turbine driven auxiliary feedwater pump may be out of service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

If the turbine driven auxiliary feedwater pump cannot be restored to service within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> time period the associated i

reactor shall be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. A motor I

driven auxiliary feedwater pump may be out of service for up to 7 days.

If the inoperable motor driven auxiliary feedwater pump l

cannot be restored to service within the 7 day time period both of the reactors shall be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Unit 1 - Amendment No. MO. 44.173

[

Unit 2 - Amendment No. M4. 44,177 15.3.4-2

For the purposes of determining a maximum allowable secondary coolant activity.

'the steam break accident is based on a postulated release of the contents of one j

3 steam generator to the atmosphere using a site boundary dose limit..The limiting dose for this accident results from iodine in the' secondary coolant.

l I-131 is the dominant. isotope because of its low derived air concentration and j

i because the other iodine isotopes have shorter half-lives and therefore cannot buildup to significant concentrations in the secondary coolant, given the limitations on primary system leak rate and activity.

It is assumed that the accident occurs at zero loid, which is when the maximum amount of water is contained in one steam gererator.

One ter.th of the contained iodine is assumed to reach the site boundary, making allowance for plate-out and retention in water droplets.

It is conservative to ineasure gross beta gamma activity except j

when the gross activity exceeds or equals 1.0 gC1/g. At this time the' iodine-l i

131 activity must be measured.

The maximum inhalation _ dose at the site boundary is then as follows:

Dose (rem) = C x V x B(t)x X x DCF

.10 0

where:

C

= secondary coolaat activity (1.0 pCi/g = 0.001 Ci/kg) l V

. water mass in cna steam generator (2877 ft' = 62.25J kg)

B(t) = breathing rate (3.47 x 10' m'/sec) x/Q = 5.0 x 10 sec/m'(

l DCF = 1.07 x 10' rem /Ci 1-131 inhaled l

The resultant dose is approximately 1.2 rem.

l

~

References:

FSAR Section 10 FSAR Section 14 Unit 1 - Amendment No. 173 Unit Amendment No. 177' 15.3.4-3 i

TABLE 15.3.5-1 (PAGE 1 0F 2)

ENGINEERED SAFETY FEATURES INITIATION INSTRUMENT SETTING LIMITS NO.

FUNCTIONAL UNIT CHANNEL SETTING LIMIT 1

High Contair, ment Pressure (Hi)

Safety Injection

• s6 psig 2

High Containment Pressure (Hi-Hi)

a. Containment Spray s 30 psig

b. Steam Line Isolation of Both Lines s 20 psig 3.

Pressurizer Low Pressure Safety Injection

• 2 1715 psig 4

Low Steam Line Pressure Safety Injection

• 2 500 psig Lead Time Constant 2 12 seconds lag Time Constant s 2 seconds 5

High Steam Flow in a Steam Line Steam Line Isolation of s d/p corresponding to Coincident with Safety Injection Affected line 0.66 x 10' lb/hr at and Low T.

1005 psig 2 540 F 6

High-high Steam Flow in a Steam Line Isolation s d/p corresponding to Steam Line Coincident with of Affected Line 4 x 10' lb/hr at Safety Injection 806 psig 7

Low-low Steam Generator Water Auxiliary Feedwater 2 20% of narrow range instrument Level Initiation 2 5% of narrow range instrument (Unit 1)**

8 Undervoltage on 4 KV Busses Auxiliary Feedwater a 75% of normal Initiation voltage Initiates also containment isolation. feedwater line isolation and starting of all containment fans.

This setting limit applies to Unit 1 until the narrow range lower tap is changed to the lower position consistent with Unit 2.

d/p means differential pressure Unit 1 - Amendment No. 444, 173 Unit 2 - Amendment No. 447, 177 Page 1 of 2

TABLE 15.4.1-2 MINIMUM FREQUENCIES FOR EOUIPMENT AND SAMPLING TESTS Tait Freauency 1.

Reactor Coolant Samples Gross Beta-gamma activity 5/ week (')

(excluding tritium)

Tritium activity Monthly czna Radiochemical E Semiannually Determination Isoto)ic Analysis for Every two weeks'"

Dose Equivalent I-131 Concentration Isotopic Analysis for a.) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Iodine including I-131.

whenever the specific I-133. and I-135 activityexceeds0.8pCi/l gram Dose Equivalent i

1-131 or 100/E gCi / gram.'6' b.) One sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a thermal power' change exceeding 15% of rated power in a one-hour period.

Chloride concentration 5/ week (8' Diss. Oxygen Conc.

5/ week (6' Fluoride Conc.

Weekly 2.

Reactor Cooicnt Boron Boron Concentration Twice/ week 3.

Refueling Water Storage Boron Concentration Weekly (

Tank Water Sample

~4.

' Boric Acid Tanks Boron Concentration Twice/ week and after each BAST concentration change when they are being relied upon as a source of borated water.

l

5.. Spray Additive Tank NaOH Concentration Monthly 6.

Accumulator Boron Concentration Monthly Unit 1 - Amendment No MB. W4 173 Unit 2 - Amendment No. MB. WA 177 Page 1 of 5

- _ _.~.. _ _. _ _... _. _. _ _ _ _ _ _.

l TABLE 15.4.1-2 (Continued) 1 Igst Frecuency

7. Spent Fuel Pit a) Boron Concentration Monthly b) Water Level i

Verification Weekly

]

8. Secondary Coolant Gross Beta-gamma Weekly"'

Activity or gamma isotopic analysis Iodine concentration Weekly when gross Beta-gamma activity equals or exceeds 1.0 yC1/g"'

[

9. Control Rods a) Rod drop times of all Each refueling or full length rods"'

after maintenance that could affect proper functioning"'

b) Rodworth measurement Following each refueling shutdown prior to commencing power operation

10. Control Rod Partial movement of Every 2 weeks""

all rods

11. Pressurizer Safety Valves Set point Every 5 years""

12. Main Steam Safety Valves Set Point Every 5 years""

13. Containment Isolation Trip Functioning Each refueling shutdown

14.. Refueling System Interlocks Functioning Each refueling shutdown

15. Service Water System Functioning Each refueling _ shutdown

16. Primary System Leakage Evaluate Monthly"'

17. Diesel Fuel Supply Fuel inventory Daily

18. Turbine Stop and Governor Functioning Annually")

Valves

19. Low Pressure Turbine Visual and magnetic Every five years i

Rotor Inspection")

particle or liquid i

penetrant j

4

20. Boric Acid System Storage Tank and Daily"

l piping temperatures

= tem 3erature required by Ta ale 15.3.2-1 i

Unit 1 - Amendment No. M8 -14,173 Unit 2 - Amendment No. % 3. N6. 177

i b.

The maximum potential seismic ground acceleration. 0.129. acting in I

the horizontal and 0.089 acting in the vertical planes simultaneously with no loss of function.

I r

3.

The' nominal Reactor Coolant System volume (both_ liquid and steam) at rated i

operating conditions and zero percent steam generator tube plugging is; j

Unit 1 - 6500 ft' Unit 2 - 6643 ft' l

References (1) FSAR Section 3.2.3 (2) Deleted-l 1

(3) Deleted I

(4) FSAR Section 3.2.3 (5) Deleted l

l (6) FSAR Table 4.1-9 i

f l

Unit 1 - Amendment 32.173 Unit 2 - Amendment (Chg dL 177 15.5.3-3

k e.

Reactor coolant activity The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 15.3.1.C.

The following information shall be included:

1.

Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the activity limit was exceeded:

2.

Results of the last isotopic analysis for radiciodine analysis i

prior to exceeding the limit results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit.

Eachresultshouldl include the date and time of sampling and the radiciodine concentrations:

3.

Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the activity limit was exceeded.

4.

Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries.per gram as a function of time for the duration of the specific activity above the steady state level: and 5.

The time duration when the specific activity of the primary j

coolant exceeded 0.8 microcuries per gram DOSE EQUIVALENT I 131. -l C.

Monthly Operating Reports 1.

Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis under the titles " Operating Data Report."

" Average Daily Power Levels" and " Unit Shutdowns" and " Power Reduction".

In addition. the report shall contain a narrative summary of operating j

experience that describes the operation of the facility, including major safety-related maintenance for the monthly report period.

L i

2.

Completed reports shall be sent by the tenth of each month following the calendar month covered by the report.

)

i i

Unit 1 - Amendment M. M. MB.173.

Unit 2. - Amendment 24. M. ~ M5.177 15.6.9-3

.