Text

Amend 83 to License DPR-3,changing Tech Specs in Partial Response to 810526 Proposed Change 139,Revs 4 & 5 as Revised on 840123 & 850226
	ML20132G327
Person / Time
Site:	Yankee Rowe
Issue date:	07/01/1985
From:	Zwolinski J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20132G320	List: ML20132G327;
References
	NUDOCS 8507190275
	Download: ML20132G327 (52)
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+ UNITED STATES j 0,k NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 L j h,.....,/

YANKEE ATOMIC POWER COMPANY YANKEE NUCLEAR POWER STATION i

DOCKET NO. 50-29 AMENDMENT TO FACILITY OPERATING LICEN5E Amendment No. 83 License No. DPR-3

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

t A. The application for amendment by Yankee Atomic Power Company (the licensee) dated May 26, 1981, as revised January 23, 1984 and February,26, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; .

B. The facility will operate in conformity with the application, .,

the provisions of the Act, and the rules and regulations of the Comission; .

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

ccnducted in compliance with the Comission'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 3 E. The issuance of this amendment is in accordance with 10 CFR Part 51 -

of the Comission's regulations and all applicable requiremente have been satisfied.

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'l 8507190275 850701 PDR ADOCK 05000029 P PDR i'

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this lice 6se _

amendment and Paragraph 2.C(2) of Facility Operating License No. DPR-3 is hereby amended to read as follows-(2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 83, are hereby incorporated in the ifcense. The licensee shall operate the facility in accordance with the Technical Specifications.

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3. This license amendment is effective as of the date of its issuance. .-,

FOR T}E( NUCLEAR REGULATCpY C0fMISSION John k DM .

. Zwolinski, Chief Opera ing Reactors Branch-#5 - .

Divisi n of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: July 1, 1985.

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a.TTACPMENT TO LICENSE AMENDMENT NO. g3 FACILITY OPERATING LICENSE NO. DPR-3 DOCKET NO. 50-29

-1 Revise Appendix A Technical Specifications by removino the pages identified below and insertino the enclosed pages. The revised pages are identi-f-ied by -

the captioned amendment number and marginal lines indicating the area of -

change.

PEPOVE INSERT 2-1 P-1 2-3 2-3 " .--

2-5 2-5 B2-1 82-1 P?-2 R2-2*

B2-3 B2-3 --

3/4 1-2 3/4 1-2 -

3/4 1-3 3/41-3*

3/4 1-4 3/4 1-4 3/4 2-1 3/4 2-1 3/4 3-3, 3/4 3-4 3/4 3-3, 3/4 3 4 3/4 3-12A 3/4 3-12A

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3/4 3-13 3/4 3-13 ,

3/4 3-28, 3/4 3-29 3/4 3-28, 3/4 3-29 3/4 3-34, 3/4 3-35 3/4 3-34, 3/4 3-35 i t

Overleaf page provided to maintain dccument completeness. No changes -

contained on this page. ,

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3/4 5-7, 3/4 5-8 3/4 5-7, 3/4 5-8 _

3/4 7-1 . 3 /4. 7.- 1 _ ..__ __ _ _. ._

3/4 7-2 3/4 7-2*

e 3/4 7-3 3/4 7-3 3/4 7-4 3/4 7-4*

i 3/4 7-29a, b 3/4 7-29a, b,.c d 3/4 8-1 3/4 8-1 -

1 j 3/4 8-2 3/4 8-2 ,

3/4 E-3, 3/4 8-4 3/4 8-3, 3/4 8-4 .-

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3/4 8-5 3/4 8-5 3/4 11-8, 3/4 11-9 3/4 11-8, 3/4 11-9 83/4 7-1 B3/4 7-1 i

R3/4 7-4 B3/4 7-4 i 6-1 6-1*

! 6-2, 6-3 6-2, 6-3 i'

, . 6-6 6-6 6-11, 6-12, 6-13 6-11, 6-12, 6-13 T1 6-15 6-15 -

, 6-18 6-18 I ~

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Overleaf page provided to maintain document completeness. No changes on this page. .

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f. 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, Main Coolant System pressure, and

' ' the highest operating loop cold leg coolant temperature shall not exceed the limits shown in Figure 2.1-1 for 4 loop operation. .. .

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l APPLICABILITY: MODES 1 and 2.

' ACTION:

Whenever the point defined by the combination of the highest operating loop cold leg temperature and THERMAL POWER has exceeded (is above and to the right of) the appropriate Main Coolant System pressure line, be in HOT ' _,

i STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MAIN COOLANT SYSTEM PRESSURE .

d 2.1.2 The Main Coolant System pressure shall not exceed 2735 psig. .

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

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ACTION:

MODES 1 and 2 Whenever the Main Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Main Coolant System pressure within its limit

, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3, 4 and 5 Whenever the Main Coolant System pressure has exceeded 2735 psig, reduce the Main Coolant System pressure to within its ilmit within 5 minutes.

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2-1 Amendment No. I 2 YANKEE-ROWE

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N TABLE 2.2-1 _

E rn REACTOR PROTECTIVE SYSTEM INSTRUMEPfrATION TRIP SETPOINTS t

FUNCTIONAL UNIT TRIP SETPOINT

1. Manual Reactor Trip Not Applicable i

2. Power Range, Neutron Flux Low Setpoint - f 35% of RATED THERMAL POWER High Setpoint - f 108% of RATED THERMAL POWER with 4 main coolant pumps operating l l Intermediate Power Range, High Setpoint - f 108% of RATED THERMAL POWER with 4 main coolant f

3. !

Neutron Flux pumps operating I '.

4. Intermediate Range, High f 5.2 decades / minutes Startup Rate

5. Source Range, Neutron Flinx Not Applicable

6. Low Main Coolant Flow > 80% of Design Flow (steam generator AP)

7. Low Main Coolant Flow > 240 Amperes, f 960 Amperes; with a response time of 1500 msec.

(main coolant pump current) I l

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ll 2.1 SAFETY LIMITS 1

BASES

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..: 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the -

j fuel and possible cladding perforation which would result in the release i of fission products to the main coolant. Overheating of the fuel cladding t' is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface I temperature is slightly above the coolant saturation temperature.

y Operation above the upper boundary of the nucleate boiling regime "."'

! could result in excessive cladding temperatures because of the onset of .

departure from nucleate boiling (DNB) and the resultant sharp reduction ,

in heat transfer coef ficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and main coolant temperature .

1 . and pressure have been related to DNB through the W-3 correlation. The -

W-3 DNB correlation has been developed to predict the DNB flux and the i

location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux ~-

that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30.

This value corresponds to a 95 percent probability at a 95 percent confidence l level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, l Main Coolant System pressure and cold leg temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. Because of flow instability, DNB may occur prematurely should the core exit quality become too great. N'J.

. The limiting core exit quality for preventing flow instability is taken .

'1 conservatively at 0.08.

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The limiting hot channel factors used in determining the thermal j limit curves are higher than those calculated at full power for the range .

j f rom all control rods fully withdrawn to maximum allowable control rod -

j ins ertion.

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B 2-1 Amendment No. II YANKEE-ROWE

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x, SAFETY LIMITS

.! BASES q '

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'$ The curv of 2.76; F(qs of are based on the following nuclear hot channel factors:

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F aEial power sM$pe.1.80; and a reference cosine with'I peak,of[1.44'for

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'i These limiting heat flux conditions are higher than those calculated j for the range of all control rods fully withdrawn to the maximum allowable 3 control rod insertion.

2.1.2 MAIN COOLANT SYSTEM PRESSURE .

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.I j The restriction of this Safety Limit protects the integrity of the j Main Coolant System from overpressurization and thereby prevents the -

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release of radionuclides contained in the main coolant from reaching the I containment atmosphere.

The reactor pressure vessel, pressurizer and pumps are designed to Section VIII of the ASPE Soiler and Pressure Vessel Code for Nuclear Power Plant, including all addenda through 1956, which pemits a maximum .,

transient pressure of 110%, 2735 psig, of design pressure. Pressure relief devices must be provided that will prevent pressure from exceeding ,

- 110 percent of the design pressure. The Main Coolant System piping and valves are designed to ANSI (fomerly ASA) Standards, Power Piping Code, Section B31.1, 1955 Edition, and B16.5,1957 Edition, respectively,

which allows the design to be based on nomal operating pressure and temperature and also allows exceeding the design conditions for periods of time. The stress level can be increased 15 percent above the Code

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allowable design value for not more than 10 percent of the design life and up to 20 percent above the allowable for up to 1 percent of the design life. Since normal plant operating pressure is 2000 psig, there is no conflict with either design condition. The setting of the Main Coolant System safety valves could allow pressure to increase to 2560 -a -

1 psig during a transient. The amount of time this condition is expected

.i to exist is well within the allowances of B31.1. The Safety Limit of -

1 2735 psig is therefore consistent with the design criteria and associated 1 code requirements.

j The entire Main Coolant System was hydrotested rt 3435 psig, 1384 k - of design pressure, to demonstrate integrity prior to initial operation.

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. 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES

'i 2.2.1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION SETPOINTS

'I The Reactor Trip Setpoint limits specified in Table 2.2-1 are the .

j values at which the reactor trips are set for each parameter. The Trip .

j Setpoints have been selected to ensure that the reactor core and Main-Coolant System are prevented from exceeding their safety limits.

j Manual Reactor Trip t

4 The Manual Reactor Trip is a redundant channel to the automatic

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1 protective instrumentation channels and provides manual reactor trip ~

! capability.

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'. l Power Ranae and Intermediate Power Range, Nuetron Flux The Power Range and Intermediate Power Range Neutron Flux channel

? high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by pressuriser water level protective -

circuitry. The Power Range low setpoint provides additional protection .

The

- in the power range for a power excursion beginning from low power.

trip associated with the low setpoint may be manually bypassed above 15 MW. and is ennustly reinstated at a power level below 15 MWe. The low i

' setpoint trip is not assumed in the accident analysis.

The prescribed s+tpoint, with allowances for . errors, is consistent with the trip point used in the accident analysis.

Intermediate Range, Neutron Flux, High Startup Rate The Intermediate Range High Startup Rate trip provides protection

' to limit the rate of power increase during low power conditions in the event of an uncontrolled rod withdrawal.

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% 4 YANKEE-ROWE B 2-3 Amendment No. 8 3

4 REACTIVITY CONTROL SYSTEMS 8

SURVEILLANCE REQUIREMENTS (Continued) i

e. Factors to consider:

1. Main Coolant System boron concentration, ,

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2. Control rod position,

3. Main Coolant System average temperature, .. . ,

4. Fuel burnup based on gross thermal energy generation, j

5. Xenon concentration, and 1

6. Samarium concentration.

l'j 4.1.1.1.1.2 The overall core reactivity balance shall be compared to *

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j predicted values to demonstrate agreement within + 0.8% ak/k at least once per l' .

31 Ef fective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.1.e, above. The predicted j

reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power -

Days titer each fuel loading.

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3/4 1-2 AMENDMENT No k , g YANxzE aows 6

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REACTIVITY CONTROL SYSTEMS SHUT 00WN MARGIN LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTLOWN MARGIN (with all control rods inserted) shall be 1 5.04 ok/k.

ARPLICABILITY: MODES 4 and 5.

ACTION:

With the SHUTOOWN MARGIN (with all control rods inserted) < 5.0% ak/k, immediately continue boration at 1 26 gpm of > 2200 ppm boron concentra-tion or equivalent and establish and maintain CONTAINMENT INTEGRITY until m,,,

the required SHUTOOWN MARGIN is restored. ,

SURVEILLANCE REQUIREMENTS '

4.1.1.2.1 The SHUTOOWN MARGIN (with all control rods inserted) shall be determined to be 1 5.0% ak/k

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a. Within one hour 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 thereaf ter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrip-pable, the SHUTDOWN HAR0!N chall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable 4

control rod (s).

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. Main Coolant System baron concentration,

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2. Control rod position, ,

3. Main Coolant System average temperature.

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and ,

6. Samarfum concentration.

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YANKEE-ROWE 3/4 1-3 i

9 lREACTIVITYCONTROLSYSTEMS SHUTDOWN MARGIN SURVEILLANCE REQUIREMENTS (Continued) 1

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YANKEE R0WE 3/4 1-4 Amendment No. 8 3 1

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. 3 /4.2 POWER DISTRIBUTION LIMITS PEAK LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 The peak linear heat generation rate (LHCR) shall not ~ exceed the -

limits of Figure 3.2-1 during steady state operation.

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APPLICABILITY: MODE 1. ",-

ACTION:

With the peak LHGR exceeding the limits of Figure 3.2-1:

a. Within 15 minutes reduce THERMAL POWER to not more than that fraction of the THERMAL POWER allowable for the main coolant pump combination in operation, as expressed below: , . ,,,

Traction of THERMAL POWER = Limitina LHCR ,

Peak Full Power LHGR .

b. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months reduce the Power Range and Intermediate Power Range .

Neutron Flux high trip setpoint to < 108% of the fraction of THERMAL POWER allowable for the main coolant pump combination.

SURVEILLANCE REQUIREMENTS __

4.2.1.1 The peak LHCR shall be determined to be within the limits of Figure 3.1-1 using incore instrumentation to obtain a power distribution maps

a. Prior to initial operation above 75% of RATED THERMAL poker af ter each fuel loading, and

b. At least once per 1,000 EFPH,

c. The provisions of Specification 4.0.4 are not applicable.

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YANKEE-ROWE 3/4 2-1 AmendmentNo.//,f,g,gg

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TABLE 3.3-1 (Continued) .

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I kEAtibk PkUTEL11WE SiSTEM IbSTRUMEh1 AT10h

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e IUTAL ho. CHAkhELS CNAkhELS APPLICABLE OF ChAhhELS TO ik1r OPERAhlt h0 DES ACTION i FUNCT105&L bhlT

11. Drhine Trip 1 1 1 1(3)(6) g i

i 12. Generator Trty 1 1 1 1(3)U) 8 1

2 1 2 1, 2 and

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13. Reacter Trip Breaker 2 1 2 1, 2 and

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$ 14. Amtematic Trip lagic 3

15. hela hteam Isolaties Trip Imgic 2 1 2 1, 2(4) 8 !

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TABLE 3.3-1 (Continued)

TABLE NOTATION

With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal. -

- The provisions of Specification 3.0.4 are not applicab-1e.'- - -

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f High voltage to detector is automatically de-energized above 5 x 10-9 Asperes on the Intermediate Ranga.

ff Or when other activities might increase reactivity.

(1) Power Range Neutron Flux Low Setpoint Trip may be manually bypassed at

> 15 MWe. Bypass shall be manually removed prior to decreasing below i T5 MWe. ~.~

(2) Intermediate Range Neutron Flux High Startup Rate Trip is automatically bypassed > 15 MWe. Bypass is automatically removed prior to decreasing .

below 15 RWe. *

(3) Trip may be manually bypassed g,15 MWe, typass is automatically removed prior to increasing above 15 MWe.

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(4) Trip may be manually bypassed when the reactor is not critical.

(5) Startup rate alarm setpoint g,1.1 decade / minute.

(6) Turbine shall be protected by at least the following protective trips:

rotor excessive axial movement, low bearing oil pressure; low condenser vacuus; and overspeed.

(7) Generator shall be protected by at least the following protective trips:

overcurrent; differential; and loss of field.

ACTION STATEKENTS T-ACTION 1 - With the number of chainels OPERABLE one less than required by the .

Minimus Channels OPERA!LE requireeent, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDlY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip .

k breakers.

i Amendment No. p 3 YANKF.E R0WE 3/4 3-4 ,

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y TABLE 3.3-2 (Continued) +

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y EhC1hELkIhc SA&ELUAkDS SYSTEM th51RUhEh1 ATION I

hlklNbh TUTAL he. LMAbhELS CMAhhELS UF thamt E Ahh SLh50k$ a.%D hEhS0kS APPLICABLE I flRK.TichAL Dh1T AND SLh50kS 101kIF UrERABLE h0 DES ACTION

2. CDNTAIIIeEhT ISOLATION (Contimmed)

c. Actuettee Chaemel B 1 1 1 1, 2, 3, 4, 5(I) 10 ,

1) high Containment l Pressure Sensor 1 1 1 1, 2, 3, 4, 5(I) 10

2) Setety Injection (All Safety Injectica Initiating Functions and Requirements) i o w 3. hAlu STEAN ISOLAT10h y a. Imer Steae Line Pressure 3/Steae Line 2/ Steam Line 3/Steae Line 1, 2, 3 6**

b. Jhetonatic 1 rip Imsic 2 1 2 1,2,3() 8 .

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c. hammal Initiation 2 1 2 1, 2 8

> d. Nigh Containment Pressure I Trip tentalmeent Isolatica 2 1 2 1, 2 8 t

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TABLE 3.3-2 (continued)

TABLE NOTATION

- The provisions of Specification 3.0.4 are not applicable. .

(1) Trip function may be bypassed 1.1 this MODE with main. coolant- pressure .

<300 psig. .. .

(2) Trip function may be bypassed in this MODE with main coolant pressure

<1800 psig and main coolant temperature <4900F.

(3) Automatic initiation of Actuation Channel #1 may be bypassed in this !! ODE j during functional test of the Main Coolant Systen Loop 1 pressure channel.

Automatic initiation of Actuation Channel #2 may be bypassed in this MODE .

during functional test of the Main Coolant System Loop 2 pressure channel. .

ACTION STATEMENTS ACTION 10 - With the number of OPERABLE channels or sensors one.less than the .

total number of channels or sensors, be in at least NOT STAND 8Y 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 ; ~

however, one safety injection channel high containment pressure sensor may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ACTION 6 - With the number of OPERABLE channels one less than the total number of channels, STARTUP and POWER OPERATION may proceed provided both of the following conditions are satisfied

1. The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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2. The minimum channels OPERABLE requirement is met; however, one ;

additional channel say be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for "

surveillance testing per Specification 4.3.1.1.

J ACTION 8 - With the number of channels OPERABLE less than required by the ,

minimus channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . .

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M YANKF.C ROWE 3/4 3-13 Amendment NO pf, 8 3 -

TABLE 3.3-6 FIRE DETECTION INSTRUMENTS INSTRUMENTATION LOCATION MINIMUM INSTRUMENTS OPERABLE

1. Control Room Above Dropped ceiling 9

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Control Boards

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Main Control Board SI Panels 1/ Panel General Area 9

2. Cable Spreading Cable Tray House 2 Manhole No. 3 1

3. Switchgear Room 20 -

Battery Roon No. I 1 ,

Battery Roon No. 2 1

4. Diesel Generators

No. I 1 No. 2 1

h. 3 1

5. Safety Injection Pumps and No. 3 Battery 5

6. Charging Pump Cubicles No. I 1 No. 2 1 No. 3 1

7. 1 & 2 Charcoal Filters 1/ Filter l

8. Turbine Building l Transf ormer 011 Cooler Area 2 i

Turbine Lube 011 Reservoir 2

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9. Vapor Container 1

10. FICS Bu11dits 1

11. Non-Return Valve (NRV) Enclosure 2 ,.

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I YANKEt ROWE 3/4 3-28 AmendmentNO.g,[,S3 l f.2 - 49m gw -g-4 = wh 9

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INSTRUMENTATION ACCIDENT MONITORING INSTRLHENTATION LIMITING CONDITION FOR OPERATION .

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3.3.3.5 The accident monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE. , _. -

APPLICABILITY: MODES 1, 2, and 3.

ACTION

a. With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3-7, either restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . --.*'

4 j b. With the number of OPERABLE accident monitoring instrumentation channels .

{ 1ess than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3-7, J either restore the inoperable channel (s) to OPERABLE status within 48 -

Lours or bc in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

c. The provisions of Specification 3.0.4 are not applicable,

~

gII.1_ANCE REQUIREMENTS 5

4.3.3.5 Each accident monitoring instrumentation channel shall be ,

desor,**trated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIIW.ATION operations at the frequencies shown in Table 4.3-5.

'4 I :

r

,i YANKEE ROWE 3/4 3-29 Amendment NO. [ 8 9 .

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t i

F 1

. _ _ - _ _ . . = . . _ _ _ . . - - . - _ _ - _ - - - -

o 4

i

TABLE 3.3-8 (Continued)

ACTION STATEMENTS l+

Action 15 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases from the -

tank may continue provided that prior to initf acing the release: -

1. At least two independent samples of the tank's contenta-are- -

j analyzed in accordance with Specification 4:11.1.1.1; -

+ 2. At least two technically qualified members of the Facility i Staff independently verify the release rate calculations and j discharge line valving; otherwise, suspend release of radioactive effluents via this pa thway.

3 Action 16 - With the number of channels OPERABLE less than required by the -

'! Minimus Channels OPERABLE requirement, effluent releases via this -

I pathway may continue provided grab samples are analyzed for gross ,

radioactivity (beta or gamma) at a limit of detection of at least 1E-07 alcrocuries/ gram: -

i

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 __

microcuries/ gram I-131.

t *

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of i

the secondary coolant is less than or equal to 0.01 microcuries/ gram I-131.

5 Action 17 - With the number of channels OPERABLE less than required by the Minimus Channels OPERABLE requirement, effluent releases via this i pathway may continue provided the flow rate is estimated at least

.- once per four hours during actual releases. (Pump curves may be used to estimate flow.)

a I

1 i-4 1

I YANKEE R0WE 3/4 3-34 AMENDMENTNo.pd,$3 ,

J i

_ _ __ ~ _ . . - . _ _ . _ . _ . , . . . .-s _ _ _ _ _ _ _ . ._

TABLE 4.3-6

< RADIDACTIVE LIQUID EFFLUENT MONITORINC INSTRUMENTATION SURVEILLANCE REQUIREMENTS

{

? CHANNEL MODES IN WHICH ,

i CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED !

1. Cross Bets orCamma Radioactivity Monitors Providing Alarm and Automatic Isolation

c. Liquid Radweste Effluents Line D P R(2) Q(1) At all times

b. Steam Generator Blowdown Tank Effluent Line D M R(2) Q(1) At all times ;

2. Continuous Composite Samplers and d

Sample Flow Measurement Device f c. Steam Generator Blowdown Tank Effluent Line D N.A. R **

Q

b. Secondary Coolant and Condensate Leakage D N.A. R ** j Q

i Turbine Building Sump D N.A. R **

c. Q

3. F1 v Rate Measurement Devices N.A. **

o. Liquid Radwaste Effluent Line D(3) .R Q ,

b. Circulating Water System . ,

$' Discha rge* D(3) N.A. N.A. N.A. i **

E l i c. Steam Cenerator Blowdown j Tank Effluent N.A. N.A. R(4) N.A. **

k. cPump curves utilized for flow rate determination.

g QCDuring teleases vla this, pathway.

' i

. i i. 4

. .- i f.I .

i :.

L ,

i

1 MAIN COOLANT SYSTEM i

s, LIMITING CONDITIONS FOgERATION (Continued)

c. With the reactor vessel and connecting pressurizer systen isolated from the heat removal system by closing the loop

{ isolation valve (s), leak testing may be performed provided that -

the coolant temperature in the reactor vessel does not increase i at a rate exceeding 500 per hour, the maximum- temperature - -

increase during the test period does not exceed 1000F, and -

] pressurizer pressure does not exceed 2485 psig.

j SURVEILLANCE REOUIREMENTS

} 4.4.1.1.3.1 The required residual heat removal loop (s) shall be determined OPERABLE per Specification 4.0.5.

4.4.1.1.3.2 The required main coolant pump (s), if not in operation, shall be - . ,

determined to be OPERABLE once per 7 days by verifying correct breaker

alignments and indicated power availability.

4.4.1.1.3.3 Determine that the steam generator (s) associated with the main coolant loop (s) required to be in operation are capable of decay heat removal . -

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

! a. The Main Coolant System is closed and pressurized to!100 psi above __

saturation pressure,

b. The Main Coolant System loop cold and hot leg stop valves are fully open, with the bypass valve closed,

c. The steam generator water level is above the top of the tube bundle,

d. An inventory of over 85,000 gallons of primary grade feedwater is available.

i e. A boiler feed pump is OPERABLE.

{ 4.4.1.1.3.4 At least one coolant loop shall be verified to be in operation -.

! and circulating main coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . "E 4.4.1.1.3.5 Verify that the Shutdown Cooling System isolation valves are locked closed within one hour prior to increasing Main Coolant Systes pressure above 300 pois. ,

l 4.4.1.1.3.6 At least once per 18 months, during shutdown, demonstrate main j

coolant loop isolation valve operability by cycling each valve through at -

least one complete cycle of full travel from the control room.

1 i :-

a lYANKEE RowE 3/4 4-2c AmendmentNo.pd,$$ f l

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4

I I

A

'I

- EMERGENCY CORE COOLING SYSTE!!S SURVEILLANCE REQUIREMENTS (continued)

5. Verifying that each ECCS safety injection subsystem is aligned -

to receive electrical power from an OPERABLE emergency bus.

Verifying that each pair of ECCS recirculation subsystem-- --'

~

6.

redundant valves is aligned to receive eleclirical' power from 1}j separate OPERABLE busses.

j 7. Verifying that each pair of ECCS long-term hot leg injection subsystem redundant valves is aligned to receive electrical power from separate OPERABLE busses.

i j 8. Verifying that the charging header flow metering instrument is " *

OPERABLE by observing charging flow rate at least once per 12 .

hours. ,

'l

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the .-

pump suctions during LOCA conditions. This visual inspection shall

-l be performed:

1. For all accessible areas of the containment prior to l establishing containment integrity, and .

2. Of the areas affected within containment at the completion of

' each containment entry when containment integrity is established.

d. At least once per 18 months by 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 corrosion, t

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

i

! 1. Cycling each power-operated valve in the flow path through at l.

i ,

least one complete cycle of full travel,

2. Verifying that valve CS-MOV-532 actuates to its correct .

f position on a safety injection signal. -

i -

3. Verifying that each of the following pumps start automatically upon receipt of a safety injection signal:

1

,l (a) High Pressure Safety Injection (HPSI) pump i

'l -

! (b) Low Pressure Safety Injection (LPSI) pump i

YANKEE ROWE 3/4 5-7 AmendmentNo.f,[,8?.

I t

t

I i

t .

i EMERGENCY CORE COOLING SYSTEMS i

,; SURVEILLANCE REOUIREMENTS (Continued) 1 4. Verifying that two low pressure safety injection pumps develop a

1 combined flow 2 2180 gys. Test every LPSI pump at least once per 36 i sonths. .

! 5. Verifying that each charging pump stops autoestically upon receipt

- of a safety injection signal. -- _ . . _ - - ..

3

'i 6. Verifying that the charging header flow metering instrument is

d. OPERABLE by performing a CHANNEL CALIBRATION.

$ 7. Verifying the proper positioning of the HPSI throttle valves

SI-V-671, 672, 673, and 674 by performing an inspection to ensure 1 tha t

(a) Each valve locking device is in place and securely welded to y

.; the valve handle and to the valve yoke.

\

t 6 (b) The scribe mark on each valve body aligns with the scribe mark

'] on the valve yoke.

8. Verifying the proper positioning of hot leg injection throttle valve j SI-V-645 at least once per 36 sooths by flow testing.

J '

i j f. At least every 36 months, and/or any time either test under 4.5.2.e.7 is

f ailed, by developing a backpressure of. 875 pois in the high pressure -

safety injection header with two HPSI pumps operating as follows:

j 1. Pressure to the suction of the HPSI pumps to be 170 + 10 psi.

2. LPSI flow is isolated.

j' 3. Injection flow is to one loop with the other loops isolated by closing the appropriate injection gate valves CS-MOV-536, i

! CS-MOV-537, CS-MOV-538, and CS-MOV-539.

4. The flow to the injection loops shall not be less than 200 spa. _.

5. The above test shall be repeated to include the operation of all HPSI pumps.

1 .

4 ,-

1 -

1 (l

YANKEE R0WE 3/45-8 AmendnentNo.[. , 3$ + ;

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! . 3/4.7 PLANT SYSTEMS

,j 3 /4.7.1 TURBINE CYCLE

-t- SAFETY VALVES

^

LIMITING CONDITION FOR OPERATION 1

[j 3.7.1.1 All main steam line code safety valves associated with each steam generator of an unisolated main coolant loop shall be OPERABLE.--- -- - .-

q .. . .

4 1 APPLICABILITY: MODES 1, 2 and 3.

i ACTION:

, jl jj a. With 4 main coolant loops and associated steam generators in M

i operation and with one or more main steam line code safety valves 1

inoperable:

e , mr h 1. Operation in MODES 1, 2 and 3 may proceed provided, that ,

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either:

a) The inoperable valve (s) is restored to OPERABLE status, .

1 or b) Three Power Range Neutron Flux channels are OPERABLE **

-~

with:

,f 4 1) The Power Range coincidence selector switch in -

the single position,

2) The trip setpoints reduced per:

(a) Table 3.7-1 for 4 loop operation. l

'- 3) One Intermediate Power Range Neutron Flux channel in the tripped condition.

s,

2. OtF rwise, 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 *d and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

__]

) ~

b. The provisions of Specification 3.0.4 are not applicable.

s SURVEILLANCE REQUIREMENTS :*

a

?; 4.7.1.1 Each main steam line code safety valve shall be demonstrated -

? OPERABLE, with lif t settings and orifice sizes as shown in Table 4.7-1, i in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, j 1974 Edition, and Addenda through Summer,1975. ;

~

,i

.4

'.j

- One Power Range Neutron Flux channel may be made inoperable for up '

to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance per Specification 4.3.1.1. _

)

4 YANKEE-ROWE 3l' 7~1 Amendment No. p/, II i l

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YANKZE-ROWE 1/k7-2 Amendment No. II -

I APR 3 1'479

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i BLANK .

i-(INTENTIONALLY) .

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YANKEE ROWE 3/4 7 3 Amendment No. p,g*

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. . ~ . . . . . . . , - - - _ . u. L w . - w. - . - . . . , - .;.. .

~

TABLE 4.7-1 5

e.s STEAM LINE SAFETY VALVES PER LOOP m

$> V5LVE MlHRER_ LIFT SETTING (1 3Q ORIFICE SIZE *

O

i, c .

a. SV-409 E F, G or 11 935 psig K

b. SV-409 A, B, C or D 985 psig K 2

c. SV-409 I, J, K or L 1035 psig Q !

w N

h

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E-3 i

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i :I l *K = 1.838 sciuare inc!ies '

K = 2.545 square inches l Q2 = 11.05 square inches ,

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i PLANT SYSTEMS

. SURVEILLANCE REQUIREMENTS (Contin _ued) _

2. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and .

compression. -

3. Snubber release rate, where required, is within the specified -

range in compression or tension. For snubbers specifically -

required not to displace under continuous load the ability of the snubber to withstand load without displacement shall be verified.

e. Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and --.

maintenance records on which the designated service life is based ~

shall be maintained as required by Specification 6.10.2.n.

l Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance -

records for each snubber listed in Table 3.7-4 shall be reviewed to verify that the indicated service lif a has not been exceeded or will not be exceeded prior to the next scheduled snubber service life .,

review. If the indicated service lif e will be exceeded prior to the next scheduled snubber service life review, the snubber service life

' l 4 shall be re-evaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of

' the next scheduled service life review. This re-evaluation, replacement, or reconditioning shall be indicated in the records.

~

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AN E R0WE , 3/4 7-29a Amendment No. ,' ,3 g _

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_ _ -_ __ 1 _I . . t. . _ _ . __ _ _ _ _ _ _ . _ _ _ _ _ . . __ . .

.....w. ~~ - .

4 TABLE 3.7-4 M r y SAFETY-kELAYED hECHANICAL SNUBBERS *

($

a 58 System Snubber HIE h hadiation Snubber Installed On Accessible or Zone Especially Difficult k

ho. Location Inaccessible Du rinR Shu td ow n*

to Remove (A or 1) (Yes or ho) (Yes or No) i CRU-SNB-19A Pres'surizer kellef A No ho Valve, PR-SV-182 CRU-S NB-198 Pressurizer Relief A No No l Valve, PR-SV-182 :

CRU-SNB-20A Pressurizer Relief A ho No Valve, PR-SV-181 ;

i A No No t*, CRU-SNB-20B Pressurizer Relief !

Valve, PR-SV-181 18 SG-ShB-lR S/C No. 1, Right Side Yes Yes o.

1 SG-ShB-1L S/G ho. 1, Let t bide 1 Yes Yes i

i SG-SNB-2R S/C ho. 2, hight Side 1 Yes Yes SG-Shb-2L S/G No. 2, Left Side 1 Yes Yes i

. SG-SNB-3R S/G No. 3 Right Side I Yes Keq E. SG-SNB-3L S/G ho. 3, Lef t Side 1 Yes Ye s t 8 i i SG-SNb-4k S/G ho. 4, hight Side I Yes Ieb f m: I !

. SG-SNB-4 L S/G No. 4 Let t Side 1 Yes Ye s i d'

'll A Yes

.s PkSH-SNh-1 Safety injection in No ;

N VC - Loop 3 g

I cm PkSit-S h h-2 Safety injection in A kes No - g i t se : VC - Loop 3 fy

' I f

"* l l6 -

l-I n'- l%l ,

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a--.-- ~ -.J L-. ~ .: a ..~^ -..:...: ~...x .-. n .. . . . . . . . .

E-.; . . . . . . . . , . . . . _ ; . . . _ . .

_.t 1,

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1 TABLE 3.7-4 (Continued) ,

rY E. System Snubber High Radiation <

~

8 Snubber Installed Dn Accessible or 4one Especially Dif ficult w No . Location Inaccessible During Shutdown ** to Remove

$ ( A or 1) (Yes or ho) (Yes or No) l e :

PRSH-S NB-3 Safety injection in A Yes No VC - Loop 1 i

PRSH-SNB-4 hafety Injection in A Yes No ,

]. VL - Loop 1 :

PRSH-SNE-5 Safety injection in A Yes No VC - Loop 2

(

PR SH-S N B-6 Safety Injection in A Yes No l

w VL - Drain Box 20 t '

ss PRSH-SNB-7 Safety injection in A Yes No de VC - Loop 4 J E

PRSH-S NB-8 Safety Injection in A Yes No VC - Loop 4

) PRSH-SN B-9 Safety injection in A Yes No I VL - Drain Box ,

i I

PR SH-SNB-10 Safety injection in A Yes No i

f E.

VL - Drain Box : i j

i

] g H PSI-S ND-1 Safety Injection in A Yes No u VL - Drain Box

I n ,

z-

O HPSI-SNB-2 Safety Injection in A Yes Np VC - Drain Box ,

is . .

l

' A Yes No H PSI-S NB-3 Safety Injection in

, VC - Drain Box :

PHLH-SNE-1A Shutdown Cooling in A Yes No j 3

i ' ;

1 VC . to Loop 4

! . 4 l $6 I

, f .

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. .- a ..._ . __,._ n. .-.. ..

A TABLE 3.7-4 (Continued) ,

i If i E- System Snubber High kadiation .

tone Especially Dif ficult !

U Snubbe r Installed On Accessible or ps No. Loca t ion Inaccessible During Shutdown ** to Remove j E (A or 1) (Yes or No) (ies or No) '

s PRCH-SNB-1B Shutdown Cooling in A Yes No l 1 VC - to Loop 4 Shutdown Cooling in A Yes No ?

PRCH-S NB-2A '

VC - to Loop 4 I

t PRCH-ShB-2B Shutdown Cooling in A Yes No l VC - t o Loo p 4 i w PRCH-ShB-3A Shutdown Cooling in A Yes No 30 VC - to Loop 4

., y Shutdown Cooling in A Yes No w PRC H-S hB-3B l $ VC - to Loop 4 i

PRCH-SNB-4A Shutdown Cooling in A Yes No VC - to Loop 4 PRCH-S hB-4B Shutdown Cooling in A Yes No i VC - to Loop 4

Snubbers may be added to safety-related systems without prior i License Amendment to Table 3.7-4 provided that a revision to :i '

1able 3.7-4 is included with the next License Amendment request. ,

5o- ;

a ** Modifications to this column due to changes in high radiation . !

a '

$l areas may be made without prior License Amendment provided tisat l a revision to 1able 3./-4 is included with the next License l ,

Amendment request.

i i !

' 8 l'

l I lI i

( . . , .- 14 ,

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--a __

t

- 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 AC SOURCES OPERATING I

~

LIMITING CONDITION FOR OPERATION _

3.8.1.1 As a minimum, the following AC electrical powef soure'es'shall be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system, and

b. Three separate and independent diesel generators:

1. Each with separate day fuel tank containing a minimum volume ,

of 210 gallons of fuel, equivalent to a 3/4 full tank, and .

. 2. With a fuel storage system containing a minimum volume of 8000-gallons of fuel, equivalent to a tank level of 4'6.5". l APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With either an offsite circuit or diesel generator of the above required AC electrical power sources inoperable, demonstrate the OPERABILITY of the remaining AC sources by performing Surveillance Requirements 4.8.1.1.1.s and 4.8.1.1.2.a.5 within one hour 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 two offsite circuits and three diesel generators to OPERABLE status within l

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 .

I i<g Ii b. With one oft. tite circuit and one diesel generator of the above 3

required AC electrical power sources inoperable, demonstrate the ;3 e

OPERABILITY of the remaining AC sources by performing Surveillance Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and _

J at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereaf ter; restore at least one of the j inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in ,

ll 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 -

l- within the following 30 s _

l ' :

l-d -

is -

i

.i 4

YANKEE-ROWE 3/4 8-1 Amendment No. 8 $

i ,

2 ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

. hours. Restore at least two offsite circuits and three diesel a generators to ODERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frcm the time of initial loss or be in at least HOT STANDBY within the next 6 ~

hours and in COLD SHUTDOWN within the followrng 30] hours-- ,

i c. With two of the above required offsite A.C. circuits inoper- l

' able, demonstrate the OPERABILITY of three diesel generators i by performing Surveillance Requirement 4.8.1.1.2.a.5 within one hour and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the i 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 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 .

d. With less than two of the above required diesel generators l" OPERABLE demonstrate the OPERABILITY of two offsite A.C.

circuits by performing Surveillance Requirement 4.8.1.1.1.a .,

within one hour 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 two 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 fol-lowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore three 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 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 rg a

4.8.1.1.1 Two physically independent circuits between the offsite _

transmission network and the onsite Class lE distribution system shall

. be:

' a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and -

i N b. Demonstrated OPERABLE at least once per 18 months during shut-ij down by manually transferring unit power supply from one 2 1

independent circuit to the second independent circuit.

j l .

3/4 8-2 Amendment No. I 3 YANKEE-ROWE _

i

,,....-s--..- . - . . . _ . . . . - ._. . . _ . . . . _ . . _ . . . . .

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d i ELECTRICAL POWER SYSTEMS

A a SURVEILLANCE REOUIREM_ENTS (Continu_td) l Each diesel generator shall be demonstrated OPERABLE

4.8.1.1.2 ,

'i

",l a. At least once per 31 days on a STAGGERED TEST RASIS by:

1. Verifying the fuel level in the day fuel tank, . .

j 2. Verifying gravity flow from the storage system to the day tanks,

3. Verifying the diesel starts from ambient condition and th.

i generator voltage reaches >432 volts within 14 sec .ods,

4. Verifying the generator is synchronized, loaded to g,200 kW, and operates for >2 hours, and

5. Verifying the diesel generator is aligned to provide standby -

power to the associated emergency buses.

b. At least once per 31 days by verifying the fuel level in the fuel .

storage tank,

c. At least once per 92 days by verifying that a sample of diesel fuel "' '

from the fuel storage tank is within the acceptable limits specified r in Table 1 of ASTM D975-68 when checked for viscosity, water, and sediment,

d. 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 t,275 a1 peres without tripping,

3. Simulating a loss of off-site power in conjunction with a safety 'dj injection test signal, and: .

a) Verifying de-energization of the energency buses.

1 :

i .

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J .

H 1

YANKEE 3/4 8-3 Amendment No. II _

.c , . ~. . .. ,. - . . . - . - . .. -. . _ . . .

g ELECTRICAL POWER SYSTEMS

.) SURVEILLANCE REOUIRE_ME_N_TS (Continued) _ __ -..

. i.

.. b) Verifying the diesel starts from ambient condition on the l auto-start signal, the diesel generator voltage reaches

>432 volts within 14 seconds, energizes the emergency buses with permanently connected loads, energizes the tj

~'

auto-connected emergency loads through the load sequencer

'y-and operates f or >5 minutes while its generator iricaded i " ' -

with the emergency loads.

i

4. Verifying the diesel generator operates for >60 minutes while

~ ;.

loaded to >400 kW.

~la

5. Verifying that the high pressure safety injection pump breakers d on each emergency bus delay 10 + 3 seconds in closing on the bus.

d.

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! ELECTRICAL POWER SYSTEMS

} -

SHUTDOWN 1:]

".g.4 1 LIMITINC CONDITION POR OPERATION

,. 1

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. .T i$ 3.8.1.2 As a minimum, the following AC electrical power sources shall be -

] OPERABLE:

a. One circuit between the offsite transmission network nd he nsite

'l Class 1E distribution system, and d

j b. One diesel generator with:

}l

'.: 1. Day fuel tank containing a minimum volume of 210 gallons of fuel, equivalent to a 3/4 full tank, and

'I

?

j 2. A fuel storage system containing a minimum volume of 4000 -

gallons of fuel, equivalent to a tank level of 2'4.5". l ,

APPLICABILITY: MODES 5 and 6.

ACTION:

fj With less than the above minimum required AC electrical power sources ~

j OPERABLE, suspend all operations involving CORE ALTEP.ATIONS or positive f reactivity changes until the minimum required AC electrical power sources -

', are restored to OPERABLE status.

1 1s SURVEILLANCE REQUIREMENTS

!k i'*

I4.8.1.2 The above required AC electrical power sources shall be demonstrated -.

b. OPERABLE by the performance of each of the Surveillance Requirements of "E L. - 4.8.1.1.1 and 4.8.1.1.2 except for Requirement 4.8.1.1.2.a.4. ,

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TABLE 4.11-2 (Continued)

TABLE NOTATION NOTE 1 - The ventilation header channels air through the ventilation system to the plant vent stack. The following ventilation systems -

discharge directly into the ventilation header. -

o Auxiliary Building o Fuel Storage -Area - - - - - - - '

> o Radwaste Building o Condenser Air-Ejector -

o Containment Purge o Waste Gas Holdup System Discharge i The steam generator blowdown vent discharges directly into the plant l vent stack.

f l a. The lower limit of detection (LLD) is defined in Table Notation a.

.l of Table 4.12-1 of Specification 4.12.1.1.

s b. The principal gaena emitters for which the LLD specification applies I exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, i

Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, .

Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these .-

nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall

~

also be identified and reported. Nuclides which are below the LLD _.

for the analyses should not be reported as being present at the LLD level for that nuclide. ,-

c. Sampling and analysis shall also be performed following a THERMAL POWER change of greater than 15 percent of RATED THERMAL POWER

within one hour. A grab sample for noble gas analysis shall be taken within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the THERMAL POWER chenge. This requirement does not apply ift (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased more than a factor of 3.

d. Samples shall be changed at least once per 7 days and analyses shall "I l

be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months af ter changing or af ter removal from .

samplers. Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following a THERMAL F0WER change of greater than 15 percent of RATED THERMAL POWER within one hour. ,

Samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will be analyzed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of '

- changing, and the corresponding LLDs may be increased by a factor of

10. This requirement does not apply if t (1) analysis shows that -

the DOSE EQUIVALENT I-131 concentration of the primary coolant has

not increased more than a factor of 3; and (2) the noble aos '

activity monitor shows that the effluent activity has not increased ;

., more than a factor of 3. , ,

t

e. The ratio of the sample flow rate to the sospled stress flow rate shall be known for the time period covered by each dose or dose rate f calculation made in accordance with Specifications 3.11.2.1, ,

3.11. 2. 2, a nd 3.11. 2. 3. . .

3 YANKEE R0WE 3/4 11-9 AMENDMENT No. ,g

l 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 saEety va'1ves ensures that the secondary system pressure will be limited to within its design pressure of 1035 peig 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 il condenser heat sink (i.e., no steam bypass to the condenser). I The specified valve lift settings and relieving capacities are in m ',,

accordance with the requirements of Section VIII of the ASME Boiler and Pressure Code, 1956 Edition. Thetotg1relievingcapacityforallvalves j

on all of tr e steam lines is 3.1 x 10 1bs/hr 6 which is 129 percent of the total second.iry steam flow of 2.4 x 10 lbs/hr at 100% RATED THERMAL POWER.

I A minimum ot 2 OPERABLE safety valves per OPERABLE steam generator ensures .

that suf ficient relieving capacity is available for the allowable THERMAL l

POWER restriction in Table 3.7-1.

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 system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Plux channels. The reactor trip setpoint reductions are derived on the following bases:

For 4 loop operation SP = (X) - (Y)(V) x (108)

X 741 Where: -

' SP = Reduced reactor trip setpoint in percent of RATED THERMAL POWER ;

V = Maximum number of inoperable safety valves per steam _

Generator YANKEE-ROWE B 3/4 7-1 Amendment No. jWI, gp ,g g f' I

i

.y._ , - _ , _ . . , - - _ . _ _ _ _ _ _ . . _ _ . _, _ ,.,, - ....__, - . _ - _ . _ _ _ _ . _ , _ . _ _ - - - _ . _ - - . _

4 e PLANT SYSTEMS j BASES 1 .

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION t

The limitation on steam generator pressure and..temperaturt ensures ._

that the pressure induced stresses in the steam generators da not exceed .

i the maximum allowable fracture toughness stress limits. The limitations '

are based on a steam generator initial RTNUT plus 60'F and are sufficient

! to prevent brittle fracture.

3/4.7.3 PRIMARY PUMP SEAL WATER SYSTEM (Deleted) ,

3/4.7.4 SERVICE WATER SYSTDf .

1 (Deleted) 3/4.7.5 CONTROL ROOM VENTILATION SYSTEM EMERGENCY SHUTDOWN The operability of the control room ventilation system emergency shutdown enhances the opportunity for the control room to remain habitable .

for Operations personnel during and following accident conditions.

3 /4.7.6 SEALED SOURCE CONTAMINATION The limitation on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more of ten than those which are not. Sealed Ili sources which are continuously enclosed within a shielded mechanism (i.e.. ~

sealed sources within radiation monitoring or boron measuring devices) are -

considered to be stored and need not be tested unless they are removed from the shielded mechanism.

i 4

YANREE-ROWE B 3/4 7-4 Amendment No. JH[, d 3 _

6.0 ADMINISTRATIVE CONTROLS

~

[ Administrative controls are the written rules, orders, instructions. '

- procedures, policies, practices, and the designation of authorities and responsibilities by the management to obtain assurarjce..or safety __. _ .

' and quality.of operation and maintenance of a nuclear power reactor. .

These controls shall be adhered to.

' 6.1 RESPONSIBILITY 6.1.1 The Plant Superintendent shall be iesponsible for overall facility operation and shall delegate in writing the succession to this .-

responsibility during his absence.

6.1.2 In all matters relating to the operation of the plant and to ,

these Technical Specifications, the Plant Superintendent shall report ~

to and be directly responsible to the Manager of Operations in the Yankee Atomic Electric Company.

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

FACILITY STAFF ,

6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

s.

Each on duty shift shall be composed of at least the minimum -l a.

shift crew composition shown in Table 6.2-1.

i l

b. At least one licensed Operator shall be in the control room when fuel is in the reactor. ;

c. At least two licensed Operators shall be present in the -

control room during reactor start-up, scheduled reactor <

shutdown and during recovery from reactor trips.

~

l d. An individual qualified in radiation protection procedures -

shall be on site when fuel is in the reactor. Operating l

crew personnel trained in radiation protection procedures

- fill this requirement.

': YANKEE-ROWE 6-1 Amendment No. 46 ij L

,______'__(___._

1 #

[ t 6.0 ADMINISTRATIVE CONTROLS (Continued)

[

e. All CORE ALTERATIONS af ter the initial fuel loading shall be F

directly supervised by either a licensed senior reactor operator or

{ ., senior reactor operator limited to fuel handling who has no other concurrent responsibilities during this operation.

.~

! f. A fire brigade of at least 5 members shall be maintained on-site at ,

all times. The fire brigade shall not include the ainla^us' shift ~~

j ,. l

! crew necessary for safe shutdown of the plant, 2 licensed operators, p; or any personnel required for other essential functions during a j fire emergency.

3 Administrative procedures shall be developed and implemented to r limit the working hours of the unit staff who perform safety-related

> functions; e.g., senior reactor operators, reactor operators, health

!. physicists, auxiliary operators, and key maintenance personnel.-

! Adequate shift coverage shall be maintained without routine heavy . l

use of overtime. The objective shall be to have operating personnel I i work a normal 8-hour day, 40-hour week while the plant is -

!' operating. However, in the event that unforeseen problems require , ,

I substantial amounts of overtise to be used, or during extended -

periods of shutdown for refueling, major maintenance or major plant l'

modifications, on a temporary basis, the following guid,elines shall ""

be followed:

i 1. An individual should not be permitted to work more than 16 -!

hours straight, excluding shif t turnover time. l i

il 2. An individual should not be permitted to work more than 16

!i hours in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any ll 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period, all N excluding shift turnover time. j

,i 9' 3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time.

l 4 Except during extended shutdown periods, the use of overtime '.4 [

should be considered on an, individual basis and not for the

entire staff on a shift. -

A, j Any deviation from the above guidelines shall be authorised by the [

j; Plant Superintendent or his delegated representative, or higher levels of management, in accordance with established procedures and

l

with documentation of the basis for granting the deviation.

.j ontrols shall be included in the procedures such that individual overtise shall be reviewed monthly by the Plant Superintendent or i

)

his delegated representative to assure that excessive hours have not I been assigned. Routine deviation from the above guidelines is not 2 ,

!' authorised.

i T

i, YANKEE-ROWE 6-2 AmendmentNo.f,f,82 I

______,._.___,,__m...,,,,_,____,m. . , _ ___, _ _ , , - _ _ . _ _ . _ , , _ _ , , . _ _ _ . _ _ _ - _ . ,

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ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff listed below shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Shif t Technical Advisor who shall have a bachelor's degree or -

equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and -- - -

accidents, and the Radiation Protection Manager who shall meet the minimum qualifications of Regulatory Guide 1.8, Revision 1.

a. Plant Superintendent

b. Assistant Plant Superintendent

c. Plant Chemistry Manager - .a

d. Plant Operations Manager .

e. Reactor Engineering Manager ,

f. Plant Maintenance Manager 3 Maintenance Supervisor

h. Instrument and Control Supervisor

~

i

1. Shif t Supervisors

j. Radiation Protection Manager I

k. Technical Director I

i 1. Shift Technical Advisor

m. Technical Services Supervisor 6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility NRC ,

licenses staff shall be maintained under the direction of the Training Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of-ANSI N18.1-1971 and Appendix "A" of 10CFR Part 55. .

6.4.2 A training program for the Fire Brigade shall be maintained under the -

j

' direction of a member of the plant staf f appointed to perform the duties of -

Fire Protection Coordinator and shall meet or exceed the requirements of Bection 27 of the NFPA Code-1975, except for Fire Brigade training sessions l which shall be held at least quarterly.

O 6-6 Amendment No. 4, d,8 3 ~

YANKEE ROWE L

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

i, i

ADMINISTRATIVE CONTROLS

b. Proposed changes to equipment or systems which inv,olve an unreviewed -

safety question as defined in Section 50.59,10 CFR.

c. Proposed test or experiments which involve an usteviewed safety- -- " -

question as defined in Section 50.59,10 CFR. .

t- .

d. Proposed changes to Technical Specifications or this operating ,

License.

e. Violations of codes, regulations, orders. Technical Specifications ,

license requirements, or of internal procedure's or instructions having nuclear safety significance.

~ '~

f. Significant operating abnormalities or deviations fr'on normal and -

expected performance of plant equipment that affect nuclear safety.

defined as Plant Information Reports.

! 3 Events requiring 24-hour written notification to the Commission. .

h. Reports and s'esting minutes of the Flant Operation Review Committee.

3 1. Perform special reviews and investigations and render ' reports thereon as requested by the Vice Frasident of Operations. ,

AUDITS 6.5 2.9 Audits of facility activities shall be performed under the cognizance of the NSAR Coasittee. These audits shall encompass:

l

a. The conformance of facility operation to provisions contsined within the Technical Specifications and applicable license conditions at least once per 12 months,125%.

b. The performance, training and qualification of those sembers of the facility staff who have a direct relationship to operation, _ _ ,

maintenance or technical aspects of the plant , at least once per 12 months,j;252. _

c. The results of actions taken to correct deficiencies occurring in facility equipment , structures , systems or method of operation that affect nuclear safety at least once per 6 months,125%.

d. The performance of activities required by the Operational Quality '

1 Assurance Frogram to meet the criteria of Appendix ~5",10 CFR 50, at least once per 24 months , 3; 252. ,

e. The Facility Emergency Plan and implementing procedures at least once per 12 months , 3; 252. . - l_

i j TAxxrt-Rouz s-11 4.end.ent no. 76 ,176, q 3 ,

8 e

.. , m, - - . -- - - - - - - . , , , . - . -. - . , - - - - - - -_ ,_ _.._ ,,, y-m .,_.,- _. _ ,_

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- . , - . . .. . .-. . . - - ~ . - . . . - . . . . . . _ . . - . . . .- - :

. 4 ADMINISTRATIVE CONTROLS

~

f. The Facility Security Plan and implementing procedures at least once.

Fer 12 months, i 25%. ,_ . _ _ , _ _ . _ ..

l '

~ '

3 The Facility Fire Protection Program and impleaenting" procedures at least once per 24 months,125%. -

T- '

k. The radiological environmental monitoring program and ths results thereof at least once per 12 months,125%.

1. The Offsite Dose Calculation Manual and implementing procedures at least once per 24 months, 1 25%.

..w

j. The PROCESS CONTROL PROGRAM and laplementing procedures for processing and packaging of radioactive vaste at least once per 24 ~

nonths 1 25%.

k.* The performanc's of activities required by the Qua!!ty Assurance -

Program to meet the provisions of Regulatory Guide 1.21. Revision 1, June 1974 and, Regulatory Cuide 4.1, Revision 1. April 19,75 at least once per 12 months, i 25%. ,

1. Any other area of facility operation considered appropriate by the NSAR Committee or the Vice President. -

AUTHORITY 6.5.2.10 The NSAR Committee shall report to and advise the Vice President on

those areas of responsibility specified in Sections 6.5.2.8 and 6.5.2.9.

RECORDS i

! 6.5.2.11. Minutes of each NSAR Committee meeting shall be prepared and

,j ,

forwarded to the Vice President and each member of the Consmittee for review within 20 working days following each meeting. The meeting minutes shall ~j j include , where applicable , reports of reviews encompassed by Section 6.5.2.8; -

i and reports of audits encompassed by Section 6.5.2.9. The review of the _

! minutes shall be completed within 60 days of the date of their distribution.

f 6.5.3 INDEPENDENT AUDIT AND REVIEW

.i ~

i 6.5.3.1 An independent fire protection and loss prevention program inspection and audit shall be performed at least once per 12 months utilizing either _

, qualified offaite licensee personnel or an outside fire protection firm. -

. 2 g,g,o .$

r m tz-Ro 6-12 A.end..nt ao.

. _ _ _ _ _ _ . - _ _ _ _ . . _ . _ , -- . _ . . - _ , , _ __ _ . . , . _ , _ . , , ___m__- . , - ._.

.--- . - - - . . . . - - a. ..

s ADMINISTRATIVE CONTROLS 6.5.3.2 An inspection and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

l 6.6 REPORTABLE OCCURRENCE ACTION .

t 6.6.1 The following actions shall be taken for REPORTABLE.0CCURRENCESt--- - ..

[ a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

l b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the t Commission shall be reviewed by the PORC and submitted to the NSAR f Committee and the Manager of Operations.

6.7 SAFETY LIMIT VIOLATION u,.,

. 6.7.1 The following actions shall be taken in the event a Safety Limit is -

violated: ,

a. The facility shall be placed in at least HOT STANDBY vithin one hour. -

b. The Safety Limit violation shall be reported to the Commission, the Manager of Operations and to the NSAR Committee within .24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . .

c. A Safety Limit Violation Report shall be prepared. The report shall -

be reviewed by the Plant Operation Review Committee. 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.

. d. The Safety Limit Violation Report shall be submitted to the Commission, the NSAR Committee and the Manager of Operations within

,j 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained "[

. that meet or exceed the requirements and recommendations of Sections 5.2-5.2.9 ~

and 5.3 of ANSI N18.7-1976 and Appendix "A" of Regulatory Guide 1.33, Revision l 2, except as provided in 6.8.2 and 6.8.3 below. The written procedures shall also cover the activities relating to:

i>

a. FIRE PROTECTION PROCRAM foplementation.

b. FRDCESS CONTROL PROGRAM implementation. .

,, c. OFFSITE DOSE CALCULATION MANUAL implementation.

!j d. Quality Assurance Frogram for effluent and environmental sonitoring, I: using the guidance in Regulatory Guide 1.21, Revision 1 June 1974 _

' and Regu3 story Guide 4.1, Revision 1 April 1975. -

l.

li YANKEE R0WE 6-13 AmendmentNo.j[,pb,g3

~

1; .

j

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

I gP ADMINISTRATIVE CONTROLS Startup Reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality,

i whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption

- or commencement of commercial power operation), supplementary reports shall be -

submitted at least every 3 months until all three events have been completed.

Annual Report. Annual Reports covering the activitiFs of the unit as -;

6.9.2 described below for the previous calendar year shall be submitted prior to March 1 of each year.

Reports required on an annual basis shall include:

a. A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures.

4 greater than 100 ares /yr and their associated man-res exposure according to work and job functions (a); e.g., reactor operations "

and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, and refueling. The dose assignment to various duty functions may be -

7 estimates based on pocket dosimeter, TLD, or film badge seasurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the whole body dose received from external sources --

shall be assigned to specific major work functions.

b. Any other unit-unique reports required on an annual basis.

c. Documentation of all challenges to the pressurizer Power-Operated Relief Valves (PORVs) or safety valves.

6.9.3 Monthly Operating Report. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Of fice of Management Information and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to tt.e appropriate Regional l' Office, to arrive no later than the 15th of each month following the calendar -

4 month covered by the report.

5

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(a) This tabulation supplements the requirements of Section 20.407 of

10CFR Part 20.

YANKEE-ROWE 6-15 Amendment No. pd, yd, g 3 i

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i ADMINISTRATIVE CONTROLS Note: This item is intended to provide for reporting of potentially generic problems.

(10) Exceeding the limits in Specification 3.11.2.6 for the storage -

of radioactive materials in the listed tanks. The written follow-up report shall include a schedule and description- of-- -

activities planned and/or taken to reduce thW contents to -

within the specified limits.

(11) Failure of the pressurizer PORVs or safety valves.

b. Thirty Day Written Reports. The REPORTABLE OCCURRENCES discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of.the

event. The written report shall include, as a minimum, a completed " = ".

copy of a licensee event report form. Information provided on the '

licensee event report form shall be supplemented, as needed, by additional narrative material to provide a complete explanation of the circumstances surrounding the event.

(1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the Technical Specifications but -

which do not prevent the fulfillment of the functional requirements of affected systems. ,

l (2) Conditions leading to operation in a degraded mode permitted by

! a limiting condition for operation or plant shutdown required l by a limiting condition for operation.

Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurstions as described in Items b(1) and b(2) need not be reported except where test results themselves reveal a degraded mode

, as described above.

<1 (3) Observed inadequacies in the implementation of administrative e lI or procedural controls which threaten to cause redretion of _

l l- degree of redundancy provided in reactor protection systems or engineered safety feature systems.

L (4) Abnormal degradation of systems other than those specified in l

Item a(3) above designed to contain radioactive material -

resulting from the fission process. .

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r I 6-18 AmendmentNo.j,pd,g3

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ADMINISTRATIVE CONTROLS

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n. Records of the service lives of all mechanical snubbers listed on l' ~

Table 3.7-4, including the date at which the service life commences and associated installation and maintenance records. _ . _ _ , _ _ _ , _ _

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6.11 RADIATION PROTECTION FROCRAM Frocedures for personnel radiation protection shall be prepared consistent -

with requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposures.

i 6.12 RICH RADIATION AREA a

6.12.1 Paragraph 20.203 " Caution signs, labels, signals, and controls". 'In . . ,

lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is -

- 1000 nrea/hr or less shall be barricaded and conspicuously posted as a high ,

radiation area, and entrance thereto shall be controlled by requiring issuance of a Radiation Ucrk Feruit.* An individual or. group of individus1s per=itted ;

~

to enter such areas sha,11 be provided with one or more of the following:

a. A radiation monitoring device which continuously indicat.as the ""

radiation dose rate in the area.

ll l '

! b. A radiation monitoring device which continuously integrates the -

radiation dose rate in the area, and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been

established and personnel have been made knowledgeable of them.

A Health Physics qualified individual (i.e., qualified in radiation c.

protection procedures), with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities l

l within the area and who will perform radiation surveillance at the frequency specified in the RWP. The surveillance frequency will be established by the Flant Health Physicist.

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Health Physics personnel shall be exenpt from the RVF issuance requirement i during the performance of their assigned radiation protection duties. _

ff Providing they are following plant radiation protection procedures for .

entry into high radiation areas.

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iq TANxrE-ROWE 6-24 AmendmentNo.J,pd,83 I m

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ML20132G327

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