Text

Suppls 860806 Transmittal of Changes to Final Draft Tech Specs,Fsar & Sser 3,necessary for Certification of Tech Specs,Editorial or Operational Flexibility.Expeditions Resolution of Items Requested
	ML18038A198
Person / Time
Site:	Nine Mile Point
Issue date:	08/19/1986
From:	Lempges T; NIAGARA MOHAWK POWER CORP.
To:	Adensam E; Office of Nuclear Reactor Regulation
References
	NMP2L-0827, NMP2L-827, NUDOCS 8608220029
	Download: ML18038A198 (338)
	v • d • e

REGULATORY INFORNATION DISTRIBUTION SYSTEN (RIDS)

ACCESSION NBR: 8608220029 DOC. DATE: 86/08/i9'OTARIZED: YES DOCKET FACIL: 50-4i0 Nine Nile Point Nuclear Station> Unit 2> Niagara Noha 05000410 AUTH. NAl'jE AUTHOR AFFIL I ATION LENP('ES> T E. Niagara Nohawl'ower Cor p.

REC IP. NANE REC IP IENT AFFILIATION ADENSAN> E. ('. Directorate 3 BNR Prospect

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SUBJECT:

Suppls 860806 transmittal of changes to final draft Tech Specs> FSAR 8c SSER 3> necessary for certification of Tech Specs> editorial or operational flexibility.Expediti ons resolution of items requested.

DISTRIBUTION CODE: 900iD COPIES RECEIVED: LTR ENCL SI ZE:

TITLE: Licensing Submittal: PSAR/FSAR Amdts 8'c Related Correspondence NOTES:

RECIPIENT COPlES REC lP I ENT COPIES ID CODE/NANE LTTR ENCL ID CODE/NAi~fE LTTR ENCL B(lR EB BNR EICSB 9(lR FOB j BMR PD3 LA B(lR PD3 PD BNR PSB i ig HAU('HEY> f'1 9WR RSB 01 2 2+

4 INTERNAL: ACRS 4i *Di'1/LFNB ELD/HDS3 IE FILE IE/DEPER/EPB 36 IE/DGAVT/GAB 2i NRR BWR ADTS NRR PNR-9 ADTS NR L NRR/DHFT/NTB E 04 RQNi

/NIB 0 EXTERNAL: BNL(ANDTS ONLY) DMS/DSS (ANDTS)

LPDR 03 1 NRC PDR 02 NSIC 05 PNL ('RUEL> R TOTAL NU}'jBER OF COPIES REQUIRED: LTTR 36 ENCL

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NIAGARA MOHAWK POWER CORPORATION/300 ERIE 80ULEVARD WEST, SYRACUSE, N.Y. 13202/TELEPHONE (315) 474-1511 August 19, 1986 NMP2L 0827 Ms. Elinor G. Adensam, Director BWR Project Directorate No. 3 U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Washington, DC 20555

Dear Ms. Adensam:

Re: Nine Mile Point Unit 2 ,

Docket No. 50-410 As a result of our continuing. review of the Final Draft Technical Specifications and Supplement 3 to the Safety Evaluation Report, Niagara Mohawk has identified changes to the following:

1. Technical Specifications The specific changes to the Technical Specifications and their justification (where appropriate) are provided in the enclosure.

2. Final Safety Analysis Report Where the Technical Specification changes affect the Final Safety Analysis Report, the changes to the appropriate pages of the Final Safety Analysis Report are provided in the enclosure. In addition, changes to the Final Safety Analysis Report are also included to correct inconsistencies between the Technical Specifications and the Final Safety Analysis Report.

3. Safety Evaluation Report Where a Technical Specification change affects the Safety Evaluation Report, this has been identified.

These changes are categorized as necessary for certification of the Technical Specifications, editorial, or for operational flexibility. The attached changes are in addition to our August 6, 1986 letter. If the previously mentioned letter contained changes on the same pages which are included in this letter, then the previous changes are also included. A list of the changes to the Technical Specifications and to the Final Safety Analysis Report is included to aid your staff in the review of these changes.

e,ao 8608220029 860819 PDR ADQCK

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0 Ms. Elinor G. Adensam, Director Page 2 Since certification of the Technical Specifications now appears to be the critical step in the Licensing of Nine Mile Point Unit 2, we would appreciate your expeditious resolution of these items.

Very truly yours, T. . Lempges Vice President Nuclear Generation WHB/ar 1917G xc: W. A. Cook, NRC Resident Inspector Project File (2)

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of Niagara Mohawk Power Corporation ) Docket No. 50-410 (Nine Mile Point Unit 2) )

AFFIDAVIT T. E. Lempges, being duly sworn, states that he is Vice President of Niagara Mohawk Power Corporation; that he is authorized on the part of said Corporation to sign and file with the Nuclear Regulatory Commission the documents attached hereto; and that all such documents are true and correct to the best of his knowledge, information and belief.

Subscribed and sworn to before me, a Notary Public in and.for t$ e State of New York and County of ~l d of 1986.

Notary Public in and for County, New York My commission expires:

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'3y Commiasion Ex ea March 30, 1

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LIST OF TECHNICAL SPECIFICATIONS AND FINAL SAFETY ANALYSIS REPORT PAGES CHANGED This Submittal Descri tion Document* ~Pa e ~Cate or Continuous Rod FSAR 15.4-3 Certification Withdrawal Analysis FSAR 15.4-4 FSAR 15.4-4a FSAR 15.4-21 9 Rod North Minimizer T.S. 3/4 1-16 Operational 10 T.S. 3/4 10-2 13 Reduced Feedwater T.S. 3/4 2-7 Certification Temperature 16 Signal-To-Noise Ratio T.S. 3/4 3-63 Certi fi cation 17 T.S, 3/4 3-88 18 T.S. 3/4 9-4 19 T.S. 3/4 10-8 22 Snubber Bases T.S. B 3/4 7-5 Certification 25 Auxiliary Boiler Pump T.S. 3/4 11-2 Certification 26 Seal Cooling Di scharge FSAR Table 11.5-2 (Service Water) Pg. 3of 4 27 FSAR 11.5-13 30 Secondary Containment T.S, 3/4 3-14 Certification Isolation Signals 35 Radi ol og i ca Env i r .

1 T.S. 3/4 12-7 Certification Monitoring Program 38 Specific Activity T.S. 3/4 4-21 Certification 41 Particulate Sample T.S. 3/4 11-9 Certification 44 Accident Monitoring T.S. 3/4 3-83 Certification 45 Instrumentation T.ST 3/4 3-84a 48 Semi-annual Radioactive T.S. 6-21 Certification Effluent Release Report

T.S. = Technical Specifications FSAR .- Final Safety Analysis Report 8608220029,

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LIST OF TECHNICAL SPECIFICATIONS AND FINAL SAFETY ANALYSIS REPORT PAGES CHANGED e

This Submittal Descri tion Document* ~Pa e ~Cate or 51 Recirculation System T.S. 3/4 4-1 Operational 52 Operation T.S. 3/4 4-2 And 53 T.S. 3/4 4-2a Certification 54 T.S. 3/4 4-4 54A T.S. 3/4 4-4a 54B T.S. B 3/4 4-1 54C T.S. B 3/4 4-la 54D T.S. B 3/4 4-2 55 T.S. B 3/4 4-2a 58 Emergency Core Cooling T.S. 3/4 5-4 Certification Systems Test Line Pressure 64 to Minimum Operable FSAR Q & R Table Certification 74 Channels 421.26-1 Pages 1 to 14 77 to Containment Isol ation FSAR Table 6.2-56 Certification 80 Signals Pages 3,4,17,17a 83 Service Water T.S. 3/4 3-111 Certification 84 T.S. 3/4 3-112 85 T.S. 3/4 3-113 86 T.S. 3/4 3-114 87 T.S. 3/4 3-115 88 T.S. 3/4 3-116 91 Other Items T.ST 2-4 Certification 92 T.S. B2-8 93 T.S. B2-9 94 T.S. 3/4 2-5 95 T.S 3/4 3-3 96 T.S. 3/4 3-13 97 T.S. 3/4 3-13a 98 T.S. 3/4 3-17 99 T.S. 3/4 3-18 100 T.S. 3/4 3-18a 101 T.S. 3/4 3-19 102 T.S. 3/4 3-26 103 T.S. 3/4 3-26a 104 T.S. 3/4 3-68 105 T,S, 3/4 6-13

T.S. = Technical Specifications FSAR = Final Safety Analysis Report

LIST OF TECHNICAL SPECIFICATIONS AND FINAL SAFETY ANALYSIS REPORT PAGES CHANGED This Submittal Descri tion Document* ~Pa e ~Cate or 106 Other Items FSAR Table 6.2-56 Certification (Conted) (page 6 of 24) 107 T.S. 3/4 6-25 108 T.S. 3/4 7-2 109 T.S. 3/4 7-2a 110 T.S. 3/4 7-3 111 T.S. 3/4 7-5 112 T.S. 3/4 7-5a 113 T.S. 3/4 7-6 114 T.S. 3/4 8-20 115 T.S. 3/4 8-22 116 T.S. 3/4 8-32 117 T.S. 3/4 8-33 118 T.S. 3/4 9-14 119 T.S. 3/4 9-15 120 T.S. 3/4 11-1 121 T.S. 3/4 11-2 122 T.S. 3/4 11-3 123 T.S. 3/4 11-4 124 T.S. 3/4 11-8 125 T.S. 3/4 12-1 126 T.S. 3/4 12-2 127 T.S. 3/4 12-8 128 T.S. 3/4 12-9 129 T.S. 3/4 12-14 130 T.S. 3/4 12-16 131 T.S. B 3/4 1-3 132 T.S. B 3/4 1-4 133 T.S. B 3/4 2-1 134 T.S. B 3/4 3-3 135 T.S. B 3/4 6-2 136 T.S. B 3/4 6-3 137 T.S. B 3/4 7-4 138 T.S. B 3/4 8-1 139 T.S. B 3/4 12-1 140 T.S. 6-20 Technical Specifications FSAR Final Safety Analysis Report

LIST OF TECHNICAL SPECIFICATIONS AND FINAL SAFETY ANALYSIS REPORT PAGES CHANGED This Submittal Descri tion Document* ~Pa e ~Cate or 142 Other Items T.S. V Editorial 143 T.S. X11 144 T.S. xviii 145 T.S. X1X 146 T.ST XX 147 T.S. xxi 148 T.S. xxiii 149 T.S. 3/4 3-1 150 T.S. 3/4 3-10 151 T.S. 3/4 3-34 152 T.S. 3/4 3-46 153 T.S. 3/4 6-24 154 T.S. 3/4 6-31 155 T.S. B 3/4 3-3 156 T.S. B 3/4 3-4 157 T.S. B 3/4 4-5

T.S. = Technical Specifications FSAR = Final Safety Analysis Report

Changes To Final Safety Analysis Report Section 15.4 To Include Continuous Rod Nithdrawal Analysis In the Start-Up Range

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Subject:

Justification for changes to Final Safety Analysis Report (FSAR) to include continuous rod withdrawal analysis in the start-up range.

The requested changes to the FSAR are enclosed.

Technical Specification 2.2.1 item 1 under Bases for Limiting Safety .System Settings references the FSAR for an analysis of continuous rod withdrawal in the start-up range. This pertains to an analytical case beyond the licensing basis. The Nuclear Regulatory Commission has imposed similar wording on BWR utilities since the issue was first debated on the Hatch 2 docket.

In order to make sure the Technical Specifications are well supported by the FSAR, Niagara Mohawk requests that FSAR Section 15.4 be revised to include an analysis of continuous rod withdrawal in the start-up range. This revision described a special study previously docketed for Hatch 2, Fermi 2, and LaSalle.

CHANGES REQUIRED FOR CERTIFICATION

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Nine Mile Point Unit 2 FSAR 15.4.1.1..5 Radiological Consequences An evaluation of the radiological consequences was not made for this event since no radioactive material is released from the fuel.

15.4.1.2 Continuous Rod Withdrawal During Reactor Startup 15.4.1.2.1 Identification of Causes and Frequency Classification The probability of initial causes of errors for this event alone is considered low enough to warrant its being categorized as an infrequent incident. The probability of further development of this event is extremely low because it i s contingent upon the simultaneous failure o f two redundant systems, the RSCS and the RWM system concurrent with a high worth rod, out-of-sequence rod selection contrary to procedures, plus operator non-acknowledgement of continuous alarm annunciations prior to safety system actuation.

15.4.1.2.2 Sequence of Events and Systems Operation Se ence of Events Control rod withdrawal errors are not considered credible in the startup and low power ranges. The RSCS and RWM prevent the operator from selecting and withdrawing .an out-of-secruence control rod. A special analysis described in Section 15.4.1.3 shows that, even for the unlikely event where the R'IlN and RSCS fail to block the continuous withdrawal of an out-of-sequence rod, the licensing basis criterion for fuel failure is still satisfied.

Continuous control rod withdrawal errors during reactor startup are precluded by the RSCS. The RSCS prevents the withdrawal ~

of an out-of-sequence control rod in the 100-percent to 75-percent control rod density range and limits rod movement to the b'anked position mode of rod withdrawal from the 75-percent rod density to the preset power level. Since only in-sequence control rods can be withdrawn in the 100-percent to 75-percent control rod density and control rods are withdrawn in the banked position mode from the 75-percent control rod density point to the preset power level, there is no basis for the continuous control rod withdrawal error in the startup and low power range. The low power range is defined as zero power to the RSCS low power set point, i.e., 20 percent of rated core power. For RWE above low power set point see Section 15.4.2. The banked position mode of the RSCS is described in Reference 1.

15.4-3

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Nine Mile Point Unit 2 FSAR Identification of 0 erator Actions No operator actions are required to preclude this event since the plant design as discussed above prevents its occurrence.

Effects of Sin le Failure and 0 erator Errors If any one of the operations involved the initial failure or error and is followed by another SEF or SOE, the necessary safety actions (e.g., rod blocks) are automatically taken prior to any limit violation. Refer to Appendix 15A for details.

15.4.1.2.3 Core and System Performance The performance of the RSCS and RWM prevent erroneous selection and withdrawal of an out-of-sequence control rod.

Thus, the core and system performance is not affected by such an operator error.

No mathematical models are involved in this event. The need for input parameters or initial conditions is not required as there are no results to report. Consideration of uncertainties is not appropriate.

15.4.1.2.4 Barrier Performance An evaluation of the barrier performance was not made for this event since there is no postulated set of circumstances for which this error could occur.

15.4.1.2.5 Radiological Consequences An evaluation of the radiological consequences is not required for this event since no radioactive material is released from the fuel.

15.4.1.3 Special Analysis The RSCS and RNM constraints on rod sequence will prevent the continuous withdrawal of an out-of-sequence rod. This analysis was performed to demonstrate that, even for the unlikely event= where an out-of-sequence control rod is withdrawn at the maximum allowable normal drive speed, the licensing basis criterion for fuel not be exceeded.

failure'ill 15.4-4

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Nine Mile Point Unit 2 FSAR This analysis is extracted for Nine Mile Point 2 application from a generic study (Reference 7) previously docketed for Hatch 2, Fermi 2, and LaSalle. The analysis concludes that as a result of continuous withdrawal of an out-of-sequence rod in the start-up range, the reactor is shut down and peak power is limited to 21'/ of rated thermal power with the peak fuel enthalpy well below the licensing basis fuel failure threshold of 170 cal/gm. Therefore, evaluation of barrier performance and radiological consequences is not necessary.

15.4.2 Rod Withdrawal Error at Power 15.4.2.1 Identification of Causes and Frequency Classification 15.4.2.1.1 Identification of Causes While operating in the power range in a normal mode of operation the reactor operator makes a procedural error and continuously withdraws the maximum worth control rod until the rod block monitor (RBM) system inhibits further withdrawal.

15.4-4a

Nine Mile Point Unit 2 FSAR 15.4.10 References

1. Paone, C.J. Banked Position Withdrawal Sequence.

NEDO-21231, September 1976.

2. General Electric Standard Application for Reactor Fuel, including United States Supplement, NEDE-24011. P-A and NEDE-24011-P-A-US (Latest approved revision).

3. USNRC . Standard Review Plan, NUREG-75/087, Washington, DC, November 24, 1975.

4. DRAGON 4 Computer Code, Dose and Radioactivity from Nuclear Facility Gaseous Outflows, NU-'115, Version 4, Level 1.

5. Horton, N.R.; Williams, W.A.; and Holtzclaw, K.W.

Analytical Methods for Evaluating the Radiological Aspects of General Electric Boiling Water Reactors.

APED-5756, March 1969.

6. Reactivity Control Data Book, GE Document No. 383HA587AI, Revision 1.

7. Stien, R.C. and Klapproth, J.F.; Continuous Control Rod Hithdrawal Transient in the Startup Range, NEON-23842, Class 1, April 1978.

15. 4-21

Changes to Technical Specifications in Area of Rod Horth Minimizer

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Subject:

Justification for changes to Technical Specifications in area of rod worth minimizer The requested changes to Technical Specifications are enclosed. The changes are a revision to the version transmitted by letter NMP2L-0807 dated August 6, 1986. This revision incorporates additional changes since the previous version to be consistent with the Hope Creek Technical Specification 3/4 10.2.

Technical justification for the changes to the Final Draft Technical Specification 3/4 10 ' is addressed in transmittal letter NMP2L-0807 .dated August 6, 1986.

CHANGES REQUIRED FOR OPERATIONAL FLEXIBILITY

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REACTIVITY CONTROL SYSTEHS 3/4. 1.4 CONTROL ROD PROGRAM CONTROLS ROD WORTH MINIHIZER LIMITING CONDITIONS FOR OPERATION

3. 1.4.1 The rod worth minimizer (RWH) shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*When THERMAL POWER is less d EOE f EO T ERMA OVER, R setpoint.

AC1ION:

a. With the RWH inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console. Otherwise, control rod movement is permitted only by actuating the manual scram or by placing the reactor mode switch in the Shutdown position.

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.1.4. 1 The RWM shall be demonstrated OPERABLE:

aO In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of control rods for the purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months before RWH automatic initiation when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.

b. In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to withdrawal of control rods for the purpose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.

c. In OPERATIONAL CONDITION.1 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after RWM automatic initiation when reducing THERMAL POWER, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.

d. By demonstrating that the control rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer.

+" Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM before withdrawal of control rods for the purpose of bringing the reactor to criticality.

NINE HILE POINT - UNIT 2 3/4 1-16

10 SPECIAL TEST EXCEPTIONS 3/4. 10.2 ROD SE UENCE CONTROL SYSTEM rod Nor4h yntnin1i)et'(RwM) par Q /b>$ C+$ ICrn'. L.p. I aed py +4<

LIMITING CONDITIONS FOR OPERATION

3. 10.2 The sequence constraints imposed on control rod groups by the rod sequence control system (RSCS) per Specification 3. 1.4.2 may be suspended by means of bypass swit'ches for the following tests provided that %he-rod-w~

<Aa&@ter-4s-OPERABLE-p~pec+f+eatien-3 .1-./

ae Shutdown margin demonstrations, Specification 4.1.1.

b. Control rod scram, Specification 4.1.3.2.

C. Control rod friction measurements.

d. Startup Test Program with the THERMAL POWER less than 20K of RATED THERMAL POWER.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the requirements of the above specification not satisfied, verify that the RSCS-iS OPERABLE per SpeCifiCatiOnS3-.:h-.4~ S.l.+,I aced B.l.hj 2, r~Spu~~r'If<ly.

g,Qp/l and/or ~ DISCS SURVEILLANCE RE UIREMENTS

4. 10.2 When the sequence constraints imposed on-GonMol-rod-groups- by the RSCS are bypassed, verify: aM/or RWH Wi-'thm-8-hoursbe fe~ypaswing-an~quence-constraint an&at 1-east~nce-

-pe~hours-wQ-1-e-any-sequence-eonMra+n-t i~bypassed.

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1'4 4 d PBIABtE.t ~p That movement of control rods from 75K ROD DENSITY to the RSCS low-power setpoint is limited to the approved control rod withdrawal sequence during scram and friction tests.

b, c. Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit tech-nical staff.

contra/ rod nrorement prescribed for /his '/ectO'rtt /s 'loarificd bg a se n

/rranscd operator or o/her tcchnrca//rf )ua/'fred marcher of thc unit

/coho/ca/ strafe present at the race/or aonso/e 7hat nrorclerrt or/ contro/ rods durl'ng shutdoron margin Chrnonstratr'one ia /m/ Cd

Ate praSCrr'bad XCIuknke. par ZpCufrkatran 8 ikr. 8, NINE MILE POINT - UNIT 2 3/4 10-2

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Changes To Technical Specification 3.2.3 In The Area of Reduced Feedwater Temperature

12

Subject:

Justification for changes to Technical Specifications 3.2.3 in the area of reduced feedwater temperature.

The requested changes are enclosed. Niagara Mohawk Power Corporation will not operate the Nine MIle Point Unit 2 with reduced feedwater temperature for the purpose of extending the normal fuel cycle. Steady state operation with reduced feedwater temperature during the normal fuel cycle will be prohibited until plant specific analyses, justifying such operation, are provided by Niagara Mohawk and approved by your staff.

CHANGES REQUIRED FOR CERTIFICATION

13 POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO ODYN OPTION B LIMITING CONDITIONS FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater ul i' than the MCPR limit shown in Figure 3.2.3-1 times the Kf shown in Figure 3.2.3-2 end-adjus-'ted-as reqtA'ed-far-r educed feedwater temperature-wi-th~

II ave B A B where:

tA = 0.86 seconds; control rod average scram insertion time limit to notch 39 per Specification 3. 1.3.3, Nq tB = 0.688 + 1.65 n

[0. 0523, X N.

i=1 n

x ave

= i-'-1 N.t.

n i=1 n = number of surveillance tests performed to date in cycle N,. = number of active control rods measured in the i surveillance test, xi = average scram time to notch 39 of all rods measured

.th in the i surveillance test N>

= total number of active rods measured in Specification

4. l. 3. 2. a.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

~Add-0&3M~h~per~~CP~he~edwate~mperaturem~00~nd-

>8204F-o~d~~ompera&e~CP~he~edwate~mperatur~20~~nd-

>250-F justment~ar e-orAyr equh ed-far-steady-state-operati.These-de+ta-MCPR-ad on-when-feedwater~empe~u~~mduced NINE MILE POINT - UNIT 2 3/4 2-7

iS Changes To Technical Specifications In The Area of Signal-To-Noise Ratio For Source Range Monitors

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Subject:

Justification for changes to Technical Specifications in the area of signal-to-noise ratio for source range monitor (SRM).

The recent Technical Specifications allow the source count rate to be less than 3 counts per second (cps) for initial fuel loading and plant startup if the following conditions are met: the signal to noise ratio is greater than 2.0, and the signal is greater than 0.7 cps.

Niagara Mohawk requests that the signal to noise ratio be changed from 2.0 to 20 (twenty). The higher signal to noise ensures that the response from the SRM is proper with the lower count rate. The requested changes to Technical Specifications are enclosed.

The 20:1 signal to noise ratio was established by a statistical analysis in order to justify the lowering to the SRM down-scale setpoint of 0.7 cps. From this analysis, it was concluded that a signal to noise ratio of 20 is necessary for a count rate signal to trip at 0.7 cps in order to maintain a confidence level of greater than 95% for source count rate monitoring and a probability of better than 97 .7% for initiating rod b'lock.

CHANGES REQUESTED FOR CERTIFICATION

TABLE 3.3.6-2 (Continued) m CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS M

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

6. Reactor Coolant S stem Recirculation Flow

. a. Upscale <108K rated flow <111K rated flow

- g b. Inoperative NA NA

c. Comparator <10K flow deviation <11% flow deviation

7. Reactor Mode Switch

a. Shutdown Mode NA NA

b. Refuel Mode NA NA

=The rod block function is varied as a function of recirculation loop flow (W), and must be maintained in accordance with note (a) of Table 2.2. 1-1. The trip setting of this average power range monitor function must also be maintained in accordance with Specification 3.2.2 and note (a) of Table 2.2.1-1.

met: the signal to noise ratio is greater than ~

"* For initial loading and startup the count rate may be less than 3 cps 80 if the following conditions are and the signal is greater than 0.7 cps.

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17 INSTRUMENTATION MONITORING INSTRUMENTATION SOURCE RANGE MONITORS SURVEILLANCE RE UIREMENTS 4.3.7.6 (Continued)

c. Verifying, before withdrawal of control rods, that the SRM count rate is at least 3 cps'ith the detector fully inserted.

f For initial loading and startup the count rate may be less than 3 cps if the following conditions are met: (1) the signal-to-noise ratio is greater than and (2) the signal is greater than 0.7 cps.

ZO NINE MILE POINT UNIT 2 3/4 3-88

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18 REFUELING OPERATIONS INSTRUMENTATION SURVEILLANCE RE UIREMENTS 4.9.2 (Continued)

b. Performing a CHANNEL FUNCTIONAL TEST:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before the start of CORE ALTERATIONS, and

2. At least once per 7 days.

C. Verifying that the channel count rate is at least 3 cps"

1. Before control rod withdrawal,

2. Before and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS, and

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

d. Verifying, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months before and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during the time any control rod is withdrawn that the shorting links have been removed from the RPS circuitry, unless adequate shutdown margin has, been demonstrated per Specification 3.1.1 and the "one rod out" interlock is OPERABLE per Specification 3.9.1.

" For initial loading and startup.the count rate may be less than 3 cps if the following conditions are met: (1) the signal-to-noise ratio is greater than and (2) the signal is greater than 0.7 cps.

2o NINE MILE POINT - UNIT 2 3/4 9"4

K hi 19 SPECIAL TEST EXCEPTIONS SPECIAL INSTRUMENTATION - INITIAL CORE LOADING SURVEILLANCE RE UIREMENTS (Continued)

4. 10.7. 1 . (Continued)

b. The RPS "shorting links" are removed.

c. The reactor mode switch is locked in the REFUEL position.

4.10.7.2 Perform a CHANNEL FUNCTIONAL TEST for the above required SRM channels within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start and at least once per 7 days during initial core loading.

4.10.7.3 For at least one SRM channel, verify that the count rate is at least 0.7 cps":

a. Immediately following the loading of the first 16 fuel bundles.

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter during initial core loading.

roSo 2o Provided signal-to-noiseis > . Otherwise, 3 cps.

NINE MILE POINT " UNIT 2 3/4 10-8

.s' ~

20 Change. To Technical Specification Bases 3/4.7.5 on Snubbers

21

SUBJECT:

Justification for the changes to Technical Specification Bases 3/4.7.5 on Snubbers The requested change is enclosed. The change is necessary to make the Bases consistent with the surveillance requirements in 4.7 .5 item e.3 of the Technical Specifications.

CHANGE REQUESTED FOR CERTIFICATION

gl gC yt 4f I

<5 l lf f l

22 PLANT SYSTEMS BASES SNUBBERS 3/4. 7. 5 (Continued) before that interval has elapsed may be used as a new reference point to deter-mine the next inspection. However, the results of such early inspections per-formed before the. or iginally required time interval has elapsed, nominal time less 25K, may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

Mhen the cause of the rejection of a snubber is clearly established and reme-died for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from'being counted as inoperable. Generically susceptible snubbers are those snubbers that are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection or are similarly located or exposed to the same environmental conditions, such as temperature, radiation, and vibration.

Nen a snubber is found inoperable, an engineering, evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversly affected by the inoperability of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the supported component or system.

The service life of a snubber is evaluated via manufacturer input and informa-tion through consider ation of the snubber service conditions and associated installation and maintenance 'records, i. e., newly installed snubber, seal replaced, spring replaced, in high-radiation area, in high-temperature area, etc. The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records .and the snubber service life review are not intended to affect plant operation.

NINE MILE POINT - UNIT 2 B3/4 7-5

23 Changes To Technical Specification and Final Safety Analysis Report in the Area of Liquid Effluent Sampling for Auxiliary Boiler Pump Seal Cooling Service Hater Discharge.

0 24

SUBJECT:

Justification for the changes to Technical Specification and Final Safety Analysis Report in the area of liquid effluent sampling for auxiliary boiler pump seal cooling service water discharge.

The requested changes to Technical Specification and Final Safety Analysis Report are enclosed.

Service water is supplied to the heat exchangers for cooling the closed loop seal water to the auxiliary boiler recirculating pump. The service water discharge from the heat exchangers is routed to the auxiliary boiler building floor drains which is directed to the radwaste system evaporator.

During system testing, it was found that the quantity of the service water returned from the heat exchangers places unnecessary burden on the radwaste evaporator. The service water return was re'-routed to the equipment drain located in the diesel fire pump cubicle. This drain is discharged to the fire pump intake shaft. Therefore, Niagara Mohawk proposes to change the Technical Specification and Final Safety Analysis Report to include periodic local grab samples of the service water discharge for radiological analysis.

Niagara Mohawk also recommends change to the Safety Evaluation Report, page 11-10, from "All release points are continuously monitored for radioactivity before discharge" to "All release points are either continuously monitored or routinely sampled for radioactivity during discharge."

CHANGE REQUESTED FOR CERTIFICATION

4//,I-/'ABLE 4-.kX-:5-RADIOACTIVE LI UIO WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(a)

TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml )

1. Batch Waste P P Principal Gamma 5xl0-7 Release Each Batch Each Batch Emitters(c)

Tanks(b)

a. 2LMS-TK4A I-131 lxlo-6

b. 2LMS-TK4B

c. 2LWS"TKSA

d. 2LMS-TK5B P One Batch/M Dissolved and One Batch/M Entrained Emitters)'x10-s Gases (Gamma P ~ M H-3 lx10-s Each Batch Composite(d)

Gross Alpha lx10-7 P Sr-89, Sr-90 SxlO-Each Batch Composite(d)

Fe-55 lx10-e

2. Continuous Grab Sample Grab Sample Principal Gamma 5xl0" 7 Releases M(e) M(e) Emitters(c)

I"131 lx10 6

a. Service Water Dissolved and lxlO-s Effluent A Entrained Gases (Gamma Emitters)

b. Service Mater H"3 lxl0-s Effluent B Gross Alpha lx10-~

c. Cooling Tower Grab Sample Grab Sample Sr-89, Sr-90 5x10-8 Blowdown --9(e) 9(e)

Fe-55 lx10-e d s ~GV. ~ LI Av g G>*b 5+wpLe Qg <<g 5*~~yq Po i u C.) q<<l Q+m~ S N'1 0 -1 SalLep 9u~p E~x~pts Cc)

Sov M Caahiug

'3))sc. 4wtge

~nll62 Lt4~3 Gw+a Sg~Lc C ~~b 5~~)p i X)O-~

NINE MILE POINT UNIT 2 3/4 11-2

',~ ~ ~

4

~ g 0

lt t~i 4 I

Nine Hile Point Unit 2 FSAR TABLE 11.5-2 (Cont)

Grab Sample at Loca I Grab Grab Sample at Sam le Poin Location No Sam le Sta ion ~eam le Radia ion Honi or Anion regeneration tank effluent X Recovered acid tank effluent X Regeneration system effluent X Ultrasonic resin cleaner resin receiver tank effluent X Ultrasonic resin cleaner resin effluent X Low conductivity waste tank effluent X Deminera lizer waste neutralizing tank effluent X Dilute acid ef'fluent X X Recovered caustic tank effluent X Dilute caustic effluent X X Recovered water sump effluent X Condensate Hakeu and Orawoff S s em Condensate transfer line X Hakeu Hater S s em Oeminera Iizer water transfer line X Condensate S stem Condensa te pump discharge X Condenser hotwel ls (6)

LP heater drains (3) X Common effluent fourth point heaters X LP heater string common effluent X Reactor Feedwa er S s em Feedwater (after last heater)

Circulatin Wa er S s em Effluent (blowdown line) X Auxi liar S earn S s em Auxiliary boiler (steam outlet) X Feedwater (pump discharge) X Auxi I iary boi ler (blowdown) X Auxiliary boiler recirc pump seal X Heat exchanger outlet (Service Water) (2) 3of4

27 Nine Mile Point Unit 2 FSAR radioactivity alarm in the main control room. Tritium in the plant areas is determined on the basis of representative grab samples collected from the effluent points or ventilation exhaust ducts. Grab samples are obtained from locations indicated in Table 11 '-2. Samples are analyzed in the health physics laboratory, or by contracted 1aboratori es.

11.5.3 Effluent Monitoring and Sampling All potentially radioactive gaseous and liquid effluent discharge paths are either continuously monitored or routinely sampled for radiation level during discharge (Section 11.5.2). Solid waste shipping containers are monitored with gamma sensitive portable survey instruments. The following gaseous effluent paths are sampled and monitored:

1. Plant main stack exhaust.

2. Combined radwaste/reactor building ventilation exhaust.

The fol lowing monitored:

liquid ef fluent paths are sampled and

1. Liquid radwaste system effluent.

2. Circulating water system cooling tower blowdown line.

3. Service water system discharge.

All monitor ranges are listed in Table 11.5-1.

An isotopic analysis is performed periodically 'on samples obtained from each liquid effluent release path to verify the adequacy of effluent processing to meet the discharge limits to unrestricted areas.

This effluent monitoring and sampling program is comprehensive and provides the information for the effluent measuring and reporting programs required by 10CFR50 Section 36a, Appendix A, General Design Criterion 64, and Appendix I and Regulatory Guide 1.21 in semiannual reports to the NRC.

The frequency of the periodic sampling and analysis described in the technical specifications is a minimum and is increased if effluent levels approach technical speci-fication limits. Isotopic content of gaseous effluents is continuously monitored by offline monitors. All

28 Changes To Technical Specification Table 3.3.2.1-1 In The Area Of Secondary Containment Isolation Signals

ri 29

SUBJECT:

Justification for change to Technical Specification Table 3.3.2.1-1 in the area of secondary containment isolation signals.

The requested change is enclosed. The change is necessary to make Table 3.3.2.1-1 consistent with the Technical Specification Table 4.3.2-1 in the area of secondary containment isolation signals (page 3/4 3-27) and the Technical Specification 3.6.5.3 applicability requirements for the Standby Gas Treatment System.

This change is also consistent with the standard "GE-STS (BHR/5)"

Technical Specification Table 4.3.2.1-1, item 2.d, which was sent to Niagara Mohawk by your staff dated September 17, 1984 (attached).

CHANGE REQUESTED'OR CERTIFICATION

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION m

VALVE GROUPS MINIMUM APPLICABLE I

foal OPERATED BY OPERABLE CHANNELS OPERATIONAL a TRIP FUNCTION SIGNAL a PER TRIP SYSTEM(b) CONDITION ACTION C)

2. RCIC Isolation Si nals (Continued)

e. RCIC Equipment Area Temperature - High 10 1 2 3 22

f. RCIC Steam Line Tunnel Temperature - High 10 1, 2, 3 22

g. Manual Isolation Push Button [RCIC](h) 10 1/Division I Only 1, 2, 3 26

h. Drywell Pressure - High(j) ll(i) 1, 2, 3 22

i. RHR/RCIC Steam Flow - High 10 1, 2, 3 22

3. Secondar Containment Isolation Si nal s

a. Reactor Building Above the Refuel (c)(d) 1. 2. 27 Floor Exhaust Radiation - High ff3.P'nd

b. Reactor Building Below the Refuel Floor Exhaust Radiation - High

( )(d) X, Z, 3, and ff j 27 pa~

TABLE 4.3.2.1-1 ISOLATION ACTUATION INSTRUMENTATION SURVI ILLANCE RE UIREHENlS CINNNEL Ol'F RATIONAL CIIANNEL FUNCTIONAL CIIANNEL CONDITIONS FOR WIIICH TRIP FUNCTION CIIECK TEST CALIOIIATION SURVEILLANCE RE( Ul RED

1. PRIHARY CONTAINMENT ISOLATION Reactor Vessel Water Level-

1) Low,'evel 3 S R 1,2,3

2) Low Low, Level 2 S R 1,2,3 Drywell Pressure - High (S) (R) 1,2,3 C. Hain Steam Line

1) Radiation - High S 1,2,3

2) Pressure - Low (S) (R) 1

3) Flow - High S R 1,2,3 Hain Steam Line Tunnel Temperature - High (S) (R) 1,2,3

e. Hain Steam Line Tunnel h Temperature - High (S) (R) 1, 2 3 f Condenser Vacuum - Low (S) (R) 3'4 (g Drywell and Suppression Chamber Radiation - High S H R I, 2, 3)

h. Hanual Initiation NA (H(')) (R) Nn 1, 2, 3

~

I.

2. SECONDARY CONTAINMENT ISOLATION Plant Exhaust Plenum Radiation - High S 1, 2, 3, and ""',

Drywell Pressure - High (S) '(R) 2, 3 Reactor Vessel Water Level - Low Low, Level 2 1, 2, 3, and.N Refueling Floor Exhaust "'"

Radiation - High S R 1, 2, 3, and "'"

e. Hanual Initiation NA (H( ))(R) NA 1, 2, 3, and

TABLE 4.3.2.1-1 (Co>>tinued)

ISOLATION ACTUATION INSTRUHENTATION SURVEILLANCE RE UIREHENTS CIIANNEL OPERATIONAL CHANNEL I UNCTIONAL CIIANNEL CONDITIONS FOR WIIICII TRIP FUNCTION CIIEK TEST CALIBRATION SURVEILLANCE RE UIRED

5. RIIR SYSTEH STEAH CONDENSING HODE ISOLATION

a. RIIR Flow - High (S) H (R) 1,2,3

b. Hanual Initiation NA (H(')) (R) NA 1,2,3 C.

6 RIIR SYSTEH SHUTDOWN COOLING HODE ISOLATION Reactor Vessel Water Level-Low, Level 3 2, 3 Reactor Vessel (RNR Cut-in Permissive). Pressure Iligh (S) H (R) 1,2,3 RNR Equipment Area A Temperature - High (S) H (R) 1,2,3 RIIR Area Cooler Temperature-Iiigh (S) H (R) 1,2,3

e. Hanual Initiation NA (H(')) (R) Nn 1, 2, 3'hen reactor steam pressure > (1043) psig and/or any turbine stop valve is ope>>.

" When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations

'with a potential for draining the reactor vessel.

¹ During CORE ALTERATION and operations with a potential far draining the reactor vessel.

((a) Hanual initiation switches shall be tested at least o>>ce per 18 months during shutdown. All other circuitry associated with manual initiation shall receive a CIIANNEL FUNCTIONAL TEST at least once per 31 days as part, of circuitry required to be tested for automatic system isolation.)

(b) Each train or logic channel shall be tested at least every other 31 days.

e 33 Change f To Techni cal Spec i i cation Table 3.12.1-1 "Radiological Environmental Monitoring Program"

34

SUBJECT:

Justification for the change to Technical Specification Table 3.12.1-1, "Radiological Environmental Monitoring Program" Niagara Mohawk requests a change to Technical Specification Table 3.12.1-1, item 4.c, from "predicted annual average ground level D/Q" to "calculated site average D/Q (based on all licensed site reactors)." The requested change is enclosed. The change is necessary to make item 4.c consistent with item 4.a of Table 3.12.1-1.

CHANGE REQUESTED FOR CERTIFICATION

~>

TABLE 3.12.1-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF .SAMPLES AND SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE LOCATIONS a COLLECTION FRE UENCY OF ANALYSIS

4. Ingestion (Continued)

c. Food One sample of each principal At time of harvest(m) Gamma isotopic(e)

Products class of food products from analysis of edible I

any area that is irrigated by portions (isotopic water in which liquid plant to include I-131) wastes have been discharged(1) of three different kinds Once per year during Gamma isotopic(e) 'amples of broad leaf vegetation (such -

the harvest season analysis of edible as vegetables) grown nearest each portions (isotopic of two different offsite loca- to include I-131) tions of highest yreNcted-annual average 'P/g (~

average-ground-levek-07+ ca(c elate <iK on all licensed site I~+~<'6 One sample of each of the similar Once per year during Gamma isotopic(e) broad leaf vegetation grown at the harvest season analysis of edible least 9.3 miles distant in a portions (isotopic least prevalent wind direction to include I-131)

36 Change To Technical Specification 3.4.5, Specific Activity

~l 37

SUBJECT:

Justification for the change to Technical Specification 3.4.5 on specific actlvi.ty.

The requested change to Technical Specification 3.4.5, Action c.3, is enclosed. The change from "at the SJAE" to "downstream of the recombiner" is necessary to make Action c.3 consistent with Technical Specification 3.4.5, Action c.2, and with Technical Specification 3.11,2.7.

CHANGE REQUESTED FOR CERTIFICATION

' a.;

I' 1

~

38 REACTOR COOLANT SYSTEM 3/4. 4. 5

~ ~ SPECIFIC ACTIVITY LIMITING CONDITIONS FOR OPERATION 3.4.5 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, and

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

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

a. In OPERATIONAL CONDITIONS 1, 2, or 3 with the specific activity of the primary cool ant

1. Greater than 0.2 microcuries per gram DOSE E(UIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE E(UIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or greater than 4.0 microcuries per gram DOSE E(UIVALENT I-131,

2. Greater than 100/E microcuries per gram, be in at least HOT SHUTDOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITIONS 1, 2, 3, or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE E(UIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4.a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

C. In OPERATIONAL CONDITION 1 or 2, with:

l. THERMAL POWER changed by more than 15X of RATED THERMAL POWER in 1 hour", or

2. The offgas level, downstream of the recombiner, increased by more than 10,000 microcuries per second in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during steady-state operation at release rates less than 75,000 microcuries per second, or

3. The offgas level, -atthe-58AE, increased by more khan 15'n 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during steady-state operation at release rates greater than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4.b of Table 4.4.5-1 until the specific activity of the primary coolant is restored within its limit.

" Not applicable during the startup test program I

NINE MILE POINT - UNIT 2 3/4 4-21

39 Change to Technical Specification Table 4.11.2-1 In the Area of Particulate Sample

'l+

40

SUBJECT:

Justification for the change to Technical Specification Table 4.11.2-1 in the area of particulate sample.

The requested change is enclosed. Sample analysis of gross alpha cannot be performed by compositing particulate filters. It is necessary to sample each filter separately for alpha analysis. Therefore, the requirement for gross alpha analysis will be included with each weekly filter change per Nine Mile Point Unit practices. This 1 proposed approach is conservative to what is presently in Final Draft version of Nine Mile Point Unit 2 Technical Specifications.

CHANGE REQUESTED FOR CERTIFICATION

TABLE 4.11.2-RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM ~

LOWER LIMIT OF SAMPLING ANALYSIS TYPE OF DETECTION (LLO)

GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml )

1. Containment(b) Each PURGE P Principal Gamma Emitters(c) lxl0-4 Each PURGE H-3 (oxide), Principal Gamma lx10-, 1xlO-~

Emitters(c)

2. Hain Stack M(d) H(d) Principal Gamma Emitters'(c) lxl0-~

Radwaste/Reactor Building Vent Grab Sample H(e) H-3 (oxide) lxl0-6 H(e)

Continuous(f) W(g) I-131 lx10-~~

Charcoal Sample Continuous(f) W(g) Principal Gamma Emitters(c) 1x10-Parti cul ate 6 ~ ~ 5 Lq h~ '~lO Sample

~FA+Auou x10="

~rt-ictAate Sam~

Continuous(f) Sr-89, Sr-90 1x10->>

Composite Particulate

.Sample

42 Change To Technical Specification Table 3.3.7.5-1, "Accident Honitoring Instrumentation" t

Z

't

43

SUBJECT:

Justification for the change to Table 3.3.7.5-1, "Accident Monitoring Instrumentation."

The requested change is enclosed.

The current Technical Specification Table 3.3.7.5-1, ACTION 81 provides for required operator action for accident monitoring instrumentation channels less than the "minimum channels operable" but does not specify the operator action when the number of channels is less than the "required number of channels". The creation of action statement 85 addresses operational situations when the number of operable channels is less than either the "required" or "minimum".

CHANGE REQUESTED FOR CERTIFICATION

0 TABLE 3. 3. 7. 5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION MINIMUM APPLICABLE REQUIRED NUMBER CHANNELS OPERATIONAL INSTRUMENT OF CHANNELS OPERABLE CONDITIONS ACTION

>2.0rywell High Range Radiation Monitors 12 IS,RHR Heat Exchanger Service Water 1/Heat Exchanger 1/Heat Exchanger 1, 2, 3 81 Radiation Monitor 13 lf. Refuel

~

Platform Area Radiation 82 Monitor 14 is, Neutron Flux'PRM 1, 2 80 IRM 1 2 80 SRM 1 80

~

15 ib,Primary Containment

~

Isolation

~

Valve Position Indication

~ ~ ~ ~

1 2

Acoustic'monitoring and tail pipe temperature

""When handling fuel, or components in the fuel pool or reactor cavity.

iNeutron flux indication is sufficient to meet the OPERABILITY requirement of this specification.

I j 1 I

45 ACTION 85 Hith the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channels to OPERABLE status within 7 days or initiate the preplanned alternate method of monitoring the appropriate parameter.

Hith the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or:

a. Initiate the preplanned alternate method of monitoring the appropriate parameter(s), and

b. In lieu of another report required by Specification 6.9.2,

'repare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3/4 3-84a

Changes To Technical Specification 6.9.1.8 In The Area Of Semi-annual Radioactive Effluent Release Report

<<1 47

SUBJECT:

Justification for the change to Technical Specification 6.9.1.8, "Semi-annual Radioactive Effluent Release Report."

The requested change is enclosed.

The current Technical Specification 6.9.1.8 (Semi-annual Radioactive Effluent Release Report) requires the use of real-time meteorology in the calculation of radiation dose assessment.

Niagara Mohawk requests the deletion of the requirement in Technical Specification 6.9.1.8 of using real-time meteorology. The Offsite Dose Calculational Manual describes the procedures for radiation dose assessment. We request the option to use average meteorology or real-time meteorology, The guidance for Radiological Effluent Technical Specifications given in NUREG-0133 recommends the use of average meteorology rather than real-time meteorology. The use of real-time meteorology is necessary in the event of specific, short time, high release situations.

CHANGE REQUESTED FOR CERTIFICATION

p>>

Al

'1 4i I

0%,

l yQ T

48 ADMINISTRATIVE CONTROLS REPORTING RE UIREMENTS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE'EPORT 6.9. 1.8 (Continued)

Plants," Revision 1, June 1974, with data summarized on a quarterly basis fol-lowing the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories:

class of solid wastes (as defined by 10 CFR 61), type of container (e.g., LSA, Type A, Type B, Large guantity), and SOLIDIFICATION agent or absorbent (e.g.,-cement, urea formaldehyde).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipftation (if measured),

or in the form of joint frequency distribution of wind speed, wind direction, and atmospheric stability." This same report shall include an assessment of the radiation doses from the radioactive liquid and gaseous effluents released from the unit during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC from their activities inside the SITE BOUNDARY (Figure 5.1.3-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, y-pVIAq a

~egaseous-pathway-.doses; dc~g shall be included, in these reports. -The-meteore4og+cab-cond-'i-'t4ens-cen~ent-

~t~he-Mme-o f release-of radieaet+ve-mater h&si-n-gaseous-ef f+uents ,as-de-The assessment of radiation doses shall be per-formed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show confor-mance with 40 CFR 190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contri-bution from liquid and gaseous effluents are"given in the ODCM.

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period.

In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorolog-ical data on site in a file that shall be provided to the NRC upon request.

NINE MILE POINT - UNIT 2 6"21

49 Changes to Technical Specifications in the Area of Recirculation System Operation

l f

50

SUBJECT:

Justification for changes to recirculation system Technical Specification The requested changes to Technical Specification are enclosed.

Nine Mile Point Unit 2 Safety Evaluation Report (February 1985), on page 4-8 states:

"The staff has reviewed the results of the stability analysis for NMP-2 and found that the maximum decay ratio is less than 0.8 for the end of the first cycle. Since the calculated maximum decay ratio is less than that of some of the operating plants (for example, Peach Bottom Unit 3 has the maximum decay ratio of 0.93), the staff concludes that NMP-2 core design stability is acceptable for cycle 1 on the condition that the Technical Specifications include appropriate limiting conditions for operation and surveillance requirements to address the concerns stated in SIL-380 and to avoid operation in regions of thermal-hydraulic instability. Also, in order to provide additional margin for stability, normal operation in natural circulation is prohibited."

In a letter from C. O. Thomas (NRC) to H. C. Pfefferlen (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8, Thermal Hydraulic Stability Amendment to GESTAR II," dated 4/24/85, the Nuclear Regulatory Commission Safety Evaluation report states:

"For GE reloads using Table 1 fuels in plants which have not yet implemented improved stability monitoring Technical Specifications the current practice of using the methods of NEDE-22277-P-1 to calculate a cycle specific decay ratio must be continued. This reload will be considered acceptable if the decay ratio is shown to be less than 0.80 for all possible operating conditions."

This statement is applied to all General Electric-designed fuels contained in Table 1 of the referenced Safety Evaluation Report which includes initial core fuel for Nine Mile Point Unit 2. Therefore, the conclusion of the Safety Evaluation Report is applicable to the Nine Mile Point Unit 2 initial core load, and no stability Technical Specifications (for two loop operation) are required, since the decay ratio is less than 0.8 for the initial core.

CHANGE REQUESTED FOR OPERATIONAL FLEXIBILITYAND CERTIFICATION

A 3/4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITIONS FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

-with.

~SIIBI '8 ' '-H-t~~

APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2".

R ACTION:

a0 With one reactor coolant system recirculation loop not in operation:

1. Within four hours:

a) Place the recircul ati on flow control system in the Loop Manual (Position Control) mode, and b) Reduce THERMAL POWER to < 70X of RATED THERMAL POWER, and, c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and, d) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0.81 times the two recirculation loop operation limit per Specification 3.2.1, and, e) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2. 1, 3.2.2 and 3.3.6.

f) Reduce the volumetric flow rate of the operating recirculation loop to < 41,000"" gpm.

" See Special Test Exception 3. 10.4.

"* This value represents the design volumetric recirculation loop flow which produces lOOX core flow at 100K THERMAL POWER. The actual value will be established during the Startup Test Program.

NINE MILE POINT - UNIT 2 3/4 4-1

52 REACTOR COOLANT SYSTEM RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITIONS FOR OPERATION Continued g) Perform Surveillance Requirement 4.4.1. 1.2 if THERMAL POWER is .

< 30K" of RATED THERMAL POWER or the recirculation loop flow 7n the operating loop is < 50K" of rated loop flow.

-2 .The-provisions-of Spoof ication-vh 0-.4-are-not-applicable.

-3 .Otherwise-be-in-a4 least-HOT-OHUTOOWN-witMn-the-next 1O-hours.

~itMno-r eaetor-cool-ant system reei-reul-ati-en loops i-n-oper&i-on,immedi-

-at~l~ieit+at~aet ion-t~reduce FHERMALPOWER-such-&at i-tH-not-whim

-the-res tr i-cted zone-of-Fi gur e-3~1 .1M-w-i-thi-n-two-hours,and-i-ni-ti-ate-

~easures-to-pl-ace-the-un'-0 i-n-avast STARTUP-wi-thin~i-x-hours-and in.

QG~HUTOGWN-wi-tM.n-'the-next-s-i-x-hours;

-c.Nkh-one-or-two-roaster-coolant-system-reoir oulmtion-loops-in-oper ation.

-and-tot~or-e-f-1 ow-1-ess than-45X-but-greater-than-39' rated-core

-f1-ow-and-'FHERMAL PGWER~i-thin-the-r es tweed-zone-of-Fi-gure-3-.4-.3~1-.

Deter H-ne-the-APRM-and-LPRM" *" norse 1-evil-s-perSpecA-f ica=

-tie a) At-1-east-onGe-per ~ht hours-o and-

-b) @+thin-30-minutes-af-ter-the-completion-of a THERMAL POWER

-increase-of at least SX-of-RATEDTHERMAL POWER;

~ W+th-the-APRM-or LRRM" "" neutron-ft-ux-nmse 1-evWs-greater tthan-three-

-times t+e~staH-ished-basel-ine-noi-se '}eve%,wi-thin-l&mmutws-

-ini-t+ate-correct+ve-act+on-t~restor e-the-noise-'lev&s-wi-thin-the-

-r equi-red Hmi swithin-two-hours-by i-ncreasi-ng-core-f kw-or-by-reduc-

+ng-THERMAL PGWER-.

Withone-or -tw~a toroo-oolat ns-yt smerec-ir oulationlop o4s-epos~in-nadn totwl-core-fl-ow-<-39X~nd-THERMAL PGWER-within the-rester i-cted-zone-of-

&i-gure-3-.4.1.1~within-l~mutes i-ni-tubate-cor rect4ve-acti-on to-reduce-

-THERMAL POWER to-witlrie-the-unrestricted-zone-of Rgure-8-.4 .1.1-1-or

" Initial values. Final values to be determined during Startup Testing based upon the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

" Value to be established during startup test program which is equivalent to minimum core flow for 2 recirculation pumps at high speed with minimum flow control valve position.

""" Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

NINE MILE POINT - UNIT 2 3/4 4-2

I 53 h) Perform surveillance requirement 4.4.1.1.4, if required, prior to entering the restricted zone of Figure 3.4.1.1-1.

i) Perform surveillance requirement 4.4.1.1.5 if total core flow is less than 45/ bu't greater than (39)%** of rated core flow, and THERMAL POWER is within the restricted zone of Figure 3.4.1.1-1.

2. With the APRM or LPRM*** neutron flux noise levels greater than three times their established baseline noise levels, within 15 minutes:

a) Initiate correct) ve action to restore the noise levels with) n the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow or by reducing THERMAL POWER, or b) Initiate corrective action to be within the unrestricted zone of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow or by reducing THERMAL POWER.

3. With total core flow less than or equal to (39)%** and THERMAL POWER within the restricted zone of Figure 3;4.1.1-1, within 15 minutes:

a) Initiate corrective action to reduce THERMAL POWER to be within the unrestricted zone of Figure 3.4.1.1-1 within 4 hours or b) Initiate corrective action to increase core flow to be greater than (39%)** within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. The provisions of Specification 3.0.4 are not applicable.

5. Otherwise, be in at lease HOT SHUTDOWN within the next 12 hours.

b. With no reactor coolant system recirculation loops in operation, within 15 minutes initiate action to reduce THERMAL POWER such that it is not within the restricted zone of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3/4 4-2a

s REACTOR COOLANT SYSTEM RECIRCULATION SYSTEM RECIRCULATION LOOPS SURVEILLANCE RE UIREMENTS Continued I

4.4.1.1.3 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by.:

a. Verifying that the control valve fails "as is" on loss of hydraulic pressure at the hydraulic control unit, and

b. Verifying that the average rate of control valve movement is:

1. Less than or equal to 13% of stroke per second opening, and

2. Less than or equal to 11X of stroke per second closing.

4-.4M4WQbl+s~ase+i-ne-APRM-and-LRM~neutr on-flux-nohe-vie-wi-thi-n-

-the-regions f~r-which-moni-toring 4-r equH ed-per-Speci-f ication-3-.4 ,1.1,. ACTCON-x,withe-two-hours-of-enteAg-the-r egi-on-for-wh-i-ch-moni-ter ieg-vrequi-red-

-urRess-basWiMog-has-pr~ou~een-pe~rmed in-the-region-s face-'the-1-as t veAeW-ng-outage

+ s<<FRT PRIE >JO +-fy.

g x gga, <, 4~ ~~knblishea'un y~+~~~'~~~ p~f ~~~

IUD/SNl 60 IVIllllll'N CONC flOHf pal /NO IZCltOllllltldN f Mhlp fpwcd urilVvtninlpnuw gloA canhol va(v4 pocH'on.

" Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRH string in the center of the core should be monitored.

NINE MILE POINT - UNIT 2 3/4 4-4

l.

P II L Pt P

t

54A 4.4.1.1.4. With one reactor coolant system recirculation loop not in operation, establish a baseline APRM and LPRM* neutron flux noise value at a point between 40% and 45/ of RATED THERMAL POWER with total core flow within (39)%** + 2% of rated core flow, prior to entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.

4.4.1.1.5 With one reactor coolant system recirculation loop not in operation, and total core flow less than 45% but greater than (39)%" of rated core flow, and THERMAL POWER within the restricted zone of Figure 3.4.1.1-1, determine the APRM and LPRM* noise levels and compare with the baseline noise levels established by Specification 4.4.F 1.4:

1) At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

2) Within 30 minutes after the completion of a thermal power increase of at least 5'/ of RATED THERMAL POWER.

3/4 4-4a

I "p V

t~

I l

lp>>

54B 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 RECIRCULATION SYSTEM The impact of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and con-trol rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6"2, respectively, MAPLHGR limits are decreased by the factor. given in Specifica-tion.3.2P,aad NCPR operating limits are adjusted per Section 3/4.2.3.

Additionally, surveillanc'e on the volumetric flow rate of the operating re-circulation loop is imposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below 30K" THERMAL POWER or 50K" rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

The objective of GE BWR plant and fuel design is to provide stable operation with margin over the normal operating domain. However, at the high-.power/

low-flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists, depending on combinations of operating conditions (e.g., rod pattern, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRM neutron flux noise levels should be monitored while operating in this region, Stability tests at operating BWRs were reviewed to determine a generic region of the power/flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0. 6 was chosen as the basis for

'determining the generic region for surveillance to account for the plant-to-plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45X of rated core flow and a THERMAL POWER Figure 3.4.1. 1-1. < SV8{js'n We NW~{i2d one Og Plant-specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case, the degree of conser-vatism can be reduced since plant-to-plant variability would be eliminated. In this case, adequate margin will be assured by monitoring the region which has a deca ratio reater tha or

~Neutron x ~eaT From pn hs B~/+ 0-I a.

ux norse >ms s are u

~ established a so to 0.8.

to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron "Initial values. The final values are determined during startup testing based upon the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head, preventing stratification.

NINE MILE POINT " UNIT 2 B3/4 4-1

fg 54C The value of 0.8 provides margin to account for calculational uncertainties in predicting the 1.0 threshold for instability. It has been determined for NMP-2 that a decay ratio of less than 0.8 exists for all conditions involving two recirculation loop operation. Consequently monitoring for instability during two recirculation loop operation is not required.

B 3/4 4-la

~ I

'4 pl l

54D REACTOR COOLANT SYSTEM BASES RECIRCULATION SYSTEM gnJSZR'T Phag Q3/9- 0 "Zw 3/4.4.1 {Continued) flux noise levels between 1X and 12K of rated power (peak-to-peak) have been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation. Neutron flux noise levels which signifi-cantly bound these values are considered in the thermal/mechanical design of GE BWR fuel and are found to be of negligible consequence. In ad"ition, sta-bility tests at operating BWRs have demonstrated that when stability-related neutron flux limit cycle oscillations occur, they result in peak-to-peak neutron flux limit cycles of 5'o 10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design"basis accident, increase the blowdown area and reduce the capability of ref looding the core; thus, the requirement for shutting down the facility when a jet pump is in-operable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation loop flow mismatch limits are in compliance with -the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop after a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other before startup of an idle loop. The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the NINE MILE POINT - UNIT 2 B3/4 4"2

4)

Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern).

Therefore, the baseline neutron flux noise level obtained at a specified core flow can be applied over a range of core flows. Initial baseline data should be taken near the minimum power/flow condition where surveillance is required as determined in Specification 4.4.1.1.5.

Baseline data taken at this condition can be applied to the entire region of surveillance since the'aseline values will be lower (and therefore conservative) at the lower power/flow condition. Since baseline noise levels do not vary significantly during a cycle, the baseline data is only required once per cycle.

B 3/4 4-2a

0 56 Changes to Technical Specification 4.5.lb in the Area of Emergency Core Cooling Systems Test Line Pressures

~ ~ S 57

Subject:

Justification for changes to Technical Specification 4.5.lb The requested changes to test line pressures are enclosed. The new pressures are based on site developed Emergency Core Cooling Systems test data.

CHANGE RE(}VESTED FOR CERTIFICATION

~ ~

58 EMERGENCY CORE COOLING SYSTEMS ECCS - OPERATING SURVEILLANCE RE UIREMENTS 4.5. 1 ECCS Oivision I, II and III shall be demonstrated OPERABLE by:

a. At least once per 31 days for the LPCS, LPCI, and HPCS systems

l. Verifying by venting at the'high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.

2. Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct" position.

b. Verifying that, when tested pursuant to Specification 4.0.5; each

1. LPCS pump develops a flow of at least 6350 gpm against a test line pressure greater than or equal to~psig.

Z.'tO flow of at least 7450 gpm against test

2. LPCI pump develops Owvkcp a

~ i' ~

1ine pressure greater than or equal to-M~ ~ psig ~A a ru4,B

3. HPCS pump develops flow of at least 6350 pm against a test line pressure greater than or equal to+85"psig.

SZ3 C. For the LPCS, LPCI and HPCSt systems, at least once per 18 months, performing a system functional test which includes simulated auto-matic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.

d. For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the condensate storage tank to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal.

" Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.

f Verify HPCS pump will auto-restart on low reactor vessel water level, level 2, .if the pump has been manually stopped.

NINE MILE POINT " UNIT 2 3/4 5-4

59 Changes to Final Safety Analysis Report Table 421.26-1 Minimum Operable Channels

h 0

60

Subject:

Changes to Final Safety Analysis Report Table 421.26-1 The requested changes to Final Safety Analysis Report Table 421.26-1 are enclosed. These changes are made to be consistent with the Technical Specifications and to address concerns raised by Carl Schulten, of your staff. The changes are defined by a vertical line in the righthand margin.

These changes will be incorporated in a subsequent amendment to the Final Safety Analysis Report.

CHANGES REQUIRED FOR CERTIFICATION

e

'll

Nine Mile Point Unit 2 FSAR TABLE 421. 26-1 REACTOR PROTECTION SYSTEM Mi nimum*

Total Operable Channels/ Channels/

Trip . Trip Functional Unit *

l. Intermediate Range Monitors

a. Neutron Flux - High

b. Inoperative

2. Average Power Range Nonitor a ~ Neutron Flux - Upscale, Setdown

b. .Flow Biased Simulated Thermal Power Upscale C. Fixed Neutron Flux - Upscale

d. Inoperative

3. Reactor Vessel Steam Dome Pressure - High 2

4. Reactor Vessel Water Level - Low, Level 3 2

5. Main Steam Line Isolation Valve-Closure 4

6. Main Steam Line Radiation - High

7. (Drywell) Pressure - High

8. Scram Discharge Volume Water Level - High

a. Transmitter /Trip Units 2

b. Float Switches 2

9. Turbine Stop Valve - Closure

10. Turbine Control Valve Fast Closure, Valve 2 Trip System Oil Pressure Low ll. Reactor Node Switch Shutdown Position

12. Manual Scram 1 of 14

From Tech. Spec. Table 3.3.1-1

~>

h,t

62 Nine Mile Point Unit 2 FSAR TABLE 421.26-1 (Cont)

Isolation Actuation Instrumentation Minimum*

Total Operable Channels/ Channels/

Trip Trip

~Sstem ~Sstem

1. Primary Containment Isolation

a. Reactor Vessel Hater Level

1. Low, Low, Low, Leve'1 1

2. Low, Low, Level 2

3. Low, Level 3

b. Drywell Pressure - High c ~ Main Steam Line

1. Radiation High 2 2

2. Pressure - Low 2 2

3. Flow High 2/line 2/line

d. Main Steam Line Tunnel

1. Temperature - High

2. Delta Temperature - High

3. Temperature - High MSL Lead Enclosure

e. Condenser Vacuum - Low

f. RHR Equipment Area Temp. High (HX's/A & B Pump Rooms)

g. Reactor Vessel Pressure High (RHR Cut-In Permissive)

h. SGTS Exhaust - High Radiation RHCU System

1. Delta Flow - High 'I

2. Delta Flow - High, Timer(1)

3. Standby Liquid Control, SLCS Initiation RHCU Equipment Area

1. Pump Room A Temp. High

2. Pump Room B Temp. - High

3. HX Room Temp. - High

k. Reactor Bldg. Pipe Chase

l. Azimuth 180 (upper), Temperature-High 1

2. Azimuth 180'lower), Temperature-High 2

3. Azimuth 40', Temperature-High 1

l. Reactor Bldg. Temp.'igh 5

m. Manual Isolation Pushbutton LNSSSSj 2 2

1 2

2 2 of 14

From Tech. Spec. Table 3 '.2-1

63 Nine Mile Point Unit 2 FSAR TABLE 421.26-1 (Cont)

Minimum*

To/a] Operable Channels/ Channels/

Trip Tr1p Tri Function 2 ~ Reactor Core Isolation Cooling System Isolation RCIC Steam Line Flow -High, Timer (1) 1 1 RCIC Steam Supply Pressure - Low 2 2 C. RCIC Steam Line Flow High 1 1

d. RCIC Turbine Exhaust Diaphragm Pressure - High

e. RCIC Equipment; Room Temperature High

f. RCIC Steam Line Tunnel Temperature High 1 1

g. Manual Isolation Pushbutton [RCICj 1/Div. 1/Div.

I only I only

h. Drywell Pressure High (2) '2 2

i. RHR/RCIC Steam Flow - High 1 1

3. Secondary Containment Isolation Signals a ~ Reactor Building Above the Refuel Floor Exhaust Radiation High

b. Reactor Building Below the Refuel Floor Exhaust Radiation - High 3 of 14

From Tech. Spec. Ta bl e 3.3. 2-1

I Nine Mile Point Unit 64 2 FSAR TABLE 421.26-1 (Cont)

Emergency Core Cooling System Actuation Instrumentation Hi n 1 mum*

Total Operabl e Channels/ Channels/

Trip Trip Tri Function

Function Function A. Division 1 Trip System

1. RHR-A (LPCI Node) 6 LPCS System

a. Reactor Vessel Water Level - I.ow .

Low Low, Level 1 b ..Drywe 1 1 Pressure High

c. LPCS Pump Discharge Flow - Low (Bypass) 1/Pum p 1/Pump

d. IPCS Injection Valve Permissive 1 1

e. LPCI Injection Valve Permissive 1 1

f. LPCI Pump A Start Time Delay Relay Normal Power I

g. LPCI Pump A Start Time Delay Relay Emergency Power

h. LPCS Pump Start Time Delay Normal Power

i. LPCS Pump Start Time Delay Emergency Power

j. LPCI Pump A Discharge Flow - Low (Bypass) 1/Pump 1/Pump

k. Manual Initiation 1 /Trl'p hrlp System System

2. Automatic Depressurization System Trip System "A"

a. Reactor Vessel Water Ievel - Low Low Low, Level 1

b. ADS Timer

c. Reactor Vessel Water Level Low, Level 3 (Permissive) 1 1

d. I,PCS Pump Discharge Pressure-High (Permissive)

e. LPCI Pump A Discharge Pressure-High (Permissive) 2 2

f. Manual Inhibit 1 1

g. 'anual Initiation 2/system 2/system 4 of 14

From Tech. Spec. Ta bl e 3.3. 3-1

0 Nine Nile Point Unit 2 FSAR 65 TABLE 421.26-1 (Cont)

Minimum

Total Operable Channels/ Channels/

Trip Trip Tri Function Function 'unction B. Division 2 Trip System

1. RHR .B and C (LPCI Mode)

a. Reactor Vessel Water Level Low Low Low, Level 1 2 2

b. Drywell Pressure - High 2 2

c. LPCI Injection Valve Permissive 1/valve 1/valve

d. IPCI Pump B Start Time Delay Relay Normal Power 1 1

e. LPCI Pump C Start Time Delay Relay Normal Power 1

f. LPCI Pump B Start Time Delay Relay Emergency Power 1

g. LPCI Pump C Start Time Delay Relay Emergency Room 1

h. LPCI Pump Discharge Flow - Low (Bypass) 1/pump

i. Nanual Initiation 1 /1/pump tri p sys'em 1/ tri p system

2. Automatic Depressurization System Trip System "B"

a. Reactor Vessel Water I evel - Low Low Low, Level 1

b. ADS Timer

c. Reactor Vessel Water Level Low, Level 3 (Permissive)

d. LPCI Pump (B and C) Discharge Pressure - High (Permissive) 2/pump 2/pump

e. Manual Inhibit 1 1

f. Manual Initiation 2/system 2/system C. Division 3 Trip System

1. HPCS System Reactor Spec. Vessel. Water level Low, Low, Level 2 4

b. Drywell Pressure - High 4 C. Reactor Vessel Water Level High, level 8 5 of 14

From Tech. Ta bl e 3.3. 3-1

.0 66 Nine Mile Point Unit 2 FSAR TABLE 421.26-1 (Cont)

Minimum*

Total Operable Channels/ Channels/

Trip Trip F Function Function

d. Condensate Storage Tank Level Low .2 2

e. Suppression Pool Water level - High 2 2

f. HPCS System Flow Rate Low (Bypass) 1 1 g Pump Discharge Pressure - High (Bypass) 1 1

h. Hanual Initiation 1/system 1/system D. Loss of Power (Div. 1, 2 and 3)

l. 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) 3/bus 2/bus

2. 4.16 kV Emergency Bus Undervoltage (Degraded Voltage) 3/bus 2/bus 6of 14

From Tech. Spec. Table 3.3.3-1

Nine Mile, Point Unit 2 FSAR TABLE 421.26-1 (Cont)

ATWS Recirculation Pump Trip System Instrumentation Minimum" Total Operable Channels/ Channels/

Trip Tl lP Tri Function

1. Reactor Vessel Water Level Low Low, I.evel 2 2

2. Reactor Vessel Pressure - High I'

7 of 14

From Tech. Spec. Table 3.3.4.1-1

Nine Mile Point Unit 2 FSAR .68 TABIE 421.26-1 (Cont)

End-of-Cycle Recirculation Pump Trip System Instrumentation Minimum*

Total Operable Channels/ Channels/

Trip Trip Tri Function

1. Turbine Stop Valve - Closure

2. Turbine Control Valve - Fast Closure 2 8 of 14

From Tech. Spec. Table 3.3.4.2-1

Nine Mile Point Unit 2 FSAR 69 TABLE 421.26-1 (Cont)

Reactor Core Isolation Cooling System Actuation Instrumentation

Minimum Total Operable Channels/ Channels/

Tr ip Tr1p Tri Function

a. Reactor Vessel Water Level Iow Iow, Level 2

b. Reactor Vessel Water Level High, Level 8

c. Condensate Storage Tank "A" Water Level - Low

d. Manual Initiation 1/system 1/system 9 of 14

From Tech. Spec. Table 3.3.5-1

Nine Mile Point Unit 2 FSAR 70 TABLE 421.26-1 (Cont)

Control Rod Block Instrumentation Minimum

Total Ooerabl e Channels/ Channels/

Trip Trip Tri Function Function Function

1. Rod Block Monitor

a. Upscale

b. Inoperative

c. Downscale

2. APRM a ~ Flow Biased Neutron Flux Upscale

b. Inoperative c Downscale

d. Neutron Flux, Upscale, Startup

3. Source Range Monitors

a. Detector not full in 3'2(3>

b. Upscale 3'

c. Inoperative 3'

e. Downscale 3 2

4. Intermediate Range Monitors

a. Detector not full in 8

b. Upscale 8

c. Inoperative 8

d. Downscale 8

5. Scram Discharge Volume

a. Water level - High, Float Switch

6. Reactor Coolant System Recirculation Flow

a. Upscale

b. Inoperative

c. Comparator

7. Reactor Mode Switch

a. Shutdown Mode 2'

b. Refuel Mode 10 of 14

From Tech. Spec. Table 3.3.6-1

Nine Mile Point Unit 2 FSAR TABLE 421.26-1 (Cont)

Plant Systems Actuation Instrumentation Mi ni mum" Total Operable Tri Function Channels Channels

1. Feedwater System/Main Turbine Trip System

a. Reactor Vessel Water Level - High, Level 8

2. Service Water System a ~ Discharge Bay Level 2 2

b. Intake Tunnel 1 and 2 1/Division 1/Division Water Temperature C. Service Water Bay

d. Service Water Pumps Discharge Strainer Differential Pressure Train "A" 1/strainer 1/strainer

e. Service Water Pumps Discharge Strainer Differential Pressure Train "B" 1/strainer 1/strainer Service Water Supply Header Discharge Water Temperature g, Service Water Inlet Pressure for EDG-2 (HPCS, Div. III)

1) Division I Supply Header 2 Division II Supply Header 11 of 14

From Tech. Spec. Table 3.3.9-1

72 Nine Mile Point Unit 2 FSAR TABLE 421.26-1 (Cont)

Radiation Monitoring Instrumentation Minimum*

Total Channels, Channels Operable

1. Hain Control Room Ventilation Radiation Honitors 2/system 2/system 12 of 14

" From Tech. Spec. Table 3.3.7.1-1

ill@ \

1

73 Nine Mile Point Unit 2 FSAR TABLE 421.26-1 (Cont)

Radioactive Liquid Effluent Monitoring Instrumentation Minimum Total Channels Instrument *, Channels Operable

l. Liquid Radwaste Effluent Line Radioactivity Honitor 3.a. Liquid Radwaste Effluent Line Flow Rate Honitor 13 of 14

From Tech. Spec. Table 3.3.7.10-1

f 0

74 Nine Mile Point Unit 2 FSAR TABLE 421.26-l (Cont)

Radioactive Gaseous Effluent Monitoring Instrumentation Minimum Total Channels Instrument *,I Channels O~erable

l. Offgas System

a. Noble Gas Activity Monitor-Providing Alarm and Automatic Termination of Release

2. Offgas System Explosive Gas Monitoring System

a. Hydrogen Monitor Train A

b. Hydrogen Monitor Train B Delta flow timer is part of flow circu1t. Technical Spec1fications incorporate the timer in the list for safety reasons (i.eee to check for drifts).

(2) Flow specifications retain this tr1p function because there are two measurements along this line that branch off to RCIC: (a) RHR/RCIC high flow - used to monitor the flow in the ceenon RCIC/RHR steam supply line, and (b) RCIC high flow - used to monitor the flow in the RCIC 11ne.

(3) MOC/TF of 3 is for startup, 2 is for refueling.

(4) See Technical Specification for valve groups and associated isolation I signals, key to isolation signals, and primary containment isolation valves.

l4 of 14

From Tech. Spec. Tab'le 3.3.7.11-1

75 CHANGES TO FINAL SAFETY ANALYSIS REPORT TABLE 6.2-56 IN THE AREA OF CONTAINMENT ISOLATION SIGNAL

76

Subject:

Changes to Final Safety Analysis Report Table 6.2-56 in the area of Containment Isolation Signal.

As requested by Ms. M. Haughey, Niagara Mohawk will revise Table 6.2-56 of the Final Safety Analysis Report.

Table 6.2-56, "Containment Isolation Provisions for Fluid Lines," lists the containment isolation valves and their respective containment isolation signals'everal valves on the table also receive system actuation signals, and these are also listed. Technical Specification Table 3.6.3-1, "Primary Containment Isolation Valve," lists only containment isolation signals and therefore, does not coincide with the Final Safety Analysis Report.

The attached shows the change to Table 6.2-56 of the Final Safety Analysis Report which differentiates between system actuation and containment isolation signal to valves:

2CSH*MOV ill 2RHS*MOV 33A, 8 2NCS*MOV 112 These changes will be incorporated in a subsequent amendment to the Final Safety Analysis Report.

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SUBJECT:

Justification for deletion of items 2.h, 2.i, and 2.j from Plant Systems Actuation Instrumentation Tables 3.3.9-1, 3.3.9-2, and 4.3.9.1-1 in the Technical Specifications.
Niagara Mohawk requests the deletion to items 2.h, 2.i, and 2.j from Plant Systems Actuation Instrumentation Tables 3.3.9-1, 3.3.9-2, and 4.3.9-1 in the Technical Specifications because these parameters do not indicate the operability of the service water system. The parameters being deleted from these tables are associated with the control room chiller subsystem which is addressed in Limiting Conditions For Operation (LCO) 3.7.3, single asterisk footnote. The deleted parameters did not assure operability of the chiller subsystem by themselves. They are only three of the numerous parameters and components monitored by the operators to assure operability of the chi llers and compliance with LCO 3.7.3.
CHANGE REQUESTED FOR CERTIFICATION
TABLE 3.3.9-1 PLANT SYSTEMS ACTUATION INSTRUMENTATION APPLICABLE TRIP FUNCTION INSTRUMENT NUMBER OPERABLE 2 ll OPERATIONAL

1. Feedwater S stem/Main Turbine Tri S stem 2222222HH'INIMUM LSH lt 2.a,>,C Reactor Vessel Mater Level - High, Level 8 I 140

2. Service Water S stem

a. Discharge Bay Level 2SWP"LS30A,B 1,2,3,4,5 142

b. Intake Tunnel 1 8 2 Water Temperature 2SWPATSL64A,65A 1/Division 1,2,3,4,5 144 2SWP*TSL64B,65B 1/Division 1,2,3,4,5 144 c.

d.
e.
f.
Service Mater Bay Service Mater Differential Service Water Differential Pumps Pumps Discharge Strainer Pressure - Train "A" Discharge Strainer Pressure - Train "B" Service Water Supply -Header Discharge Mater Temperature 2SWP
~
2SWPALS73A,B
$ 'OSA "PO~~
'POSH TY 2SWP9P5lA, ih,C,E, L~)~P'22 B
1/Strainer 1/Strainer 1,2,3,4,5 1,2,3,4,5 1,2,3,4,5 1,2,3,4,5 143 146 146 147 EDG-2 (HPCS, Division III) et+el-BuiMmgWer 22eee-W Temp erat-ure-
>~9)cc: &AT) w 'XN~ TR~ssv~ poR.
E (Hf C9 ) DWLS)ow ~)
) Iv) Slo+ ~ pvpPL-'I QeAog+ gsU)w ~ ps< qsp l) Z,+) +) 5 lHG Z) D) VLS\og W SU>'Pi% HBPQc~ ZSRPW P4965 l 1
) Z)3) l )5 l~
W 1,
-IHSTRUMEHT-
~UMBER ates-5 m-gont4nuW SWP"TT35
'~
VE pl M eve+Hence-vri44oat
~Hag-th
'-+Hyped-cond-i-t-io, n&s&f'v4c~e leve4-wQch-may-be-placed i-r~~operatAe-W~s fee-up-to~hours-without-pkac4og-@henri@-System 58l'y Nt'
4 Ip
85 TABLE 3. 3.9-1 (Continued)
PLANT SYSTEMS ACTUATION INSTRUMENTATION ACTION ACTION 140 - a. With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the number of OPERABLE channels two less than required

~
by the Minimum OPERABLE Channels requirement, restore at least one of the inoperable channels to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least STARTUP within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 141-Nrt'.sek
-dddk ddd-:dd-d d>> ddd'dd 'dd
~d d <<H
~c-i-HfH-ter-tra~inoperatrl-e-and-take-t+~CRGN-required dd I
ACTION 142- Monitor discharge bay level continuously if level reaches trip setpoint, provide an alternate flow discharge path by locking closed 2SWP"MOV30A or 2SWP"MOV30B.
ACTION 143- Monitor service water bay level continuously if level reaches Trip Setpoint provide an alternate intake to the service'ay by locking open 2SWP"MOV77A or 2SWP*MOV77B.
ACTION 144- Place intake heaters in service if lake temperature < 39'F or take the ACTIONS required by Specifications'.7. l. 1 and 3.7. 1.2, as appropriate.
ACTION 145- Lock closed 2SWP"MOV95A or 2SWP*MOV95B and declare EDG-2 (HPCS, Division III) inoperable and take the ACTION required by Specification 3.8. 1.
ACTION 146- Monitor the effected pump discharge pressure and the applicable service water loop header pressure to determine the differential pressure across the strainer; if the differential pressure exceeds the setpoint manually start the strainer or declare-the effected service water pump inoperable and take the ACTION required by Specifications 3. 7. 1. 1 and 3. 7. 1. 2, as appropriate.
ACTION 147- Monitor service water local discharge temperature indicators as applicable per Specification 4. 7. l. l. l. a. 2 or 4. 7. 1. 2. l. a. 2.
NINE MILE POINT - UNIT 2 3/4 3-113 JUN 25 lg86
I' 86 TABLE 3.3.9-2 PLANT SYSTEMS ACTUATION INSTRUMENTATION SETPOINTS ALLOMABLE TRIP FUNCTION TRIP SETPOINT VALUE

1. Feedwater S stem/Main Turbine zo% >

a. Reactor Vessel Water Level - High <202.3 in.* in.

Level 8

2. Service Mater S stem

a. Discharge Bay Level <275'lev. <275'-3/4" Elev.

b. Intake Tunnel 1 4 2 >39 F >38 F Mater Temperature

c. Service Mater Bay >234'lev. >233'-l/4" Elev.

d. Service Water Pumps Discharge <10 psid <14.5 psid Strainer Differential Pressure-Train "A"

e. Service Water Pumps Discharge <10 psid <14.5 psid Strainer Differential Pressure" Train "B"

f. Service Mater Supply Header NA Discharge Water Temperature P

+or-EBS+PCS ,9m&-mn-&H-)-
Gont~o4-BA4dmg-Mate~
-i.Contre@-Bm-M&g-Service.
Rater IA-e-t Temperatur o
-j.Control-Bui-I~g-Servic Mater-Outl-et Temperate r~

See Bases Figure B3/4 3-1.

Sz,rvi'ca Agee X let t'tzcSura
)br EDQQZ C.HPcS~ pr'uisio~~)
I) bh/rsi'on X S'upply Hider >'5'rig 5 l75p5(
2P 3>iv>sion X Supp/y h'~PA- > Z2 pSig l7.Gp&If NINE MILE POINT - UNI 2 -ll
N I
I
TABLE 4.3.9.1-1 PLANT SYSTEHS ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREHENTS CHANNEL OPERATIONAL CONDITIONS CHANNEL FUNCTIONAL CHANNEL FOR WHICH SURVEILLANCE Cl TRIP FUNCTION CHECK TEST CALIBRATION IS RE UIRED M

1. Feedwater S stem/Hain Turbine Tri S stem

a. Reactor Vessel Mater Level - High Level 8

2. Service Mater S stem

a. Discharge Bay Level 1, 2, 3, 4, 5

b. Intake Tunnel 1 8 2 Water Temperature W 1, 2, 3, 4, 5

c. Service Mater Bay 1, 2, 3, 4, 5

d. Service Mater Pumps Discharge Strainer S 1, 2, 3, 4, 5 Differential Pressure - Train "A"

e. Service Water Pumps Discharge Strainer S 1, 2, 3, 4, 5 Differential Pressure - Train "B"

f. Service Water Supply Header Discharge 1,2,3,4,5 Water Temperature

&GAPES-,~iv-ts-io~~-
tpggigg tti-h}ing-Ma&r-Re
" Calibration excludes sensors; a comparison test of the four RTDs will be done.
Service Wab,r T~tat Fr~ssu<a Gory 2 /pe&, Pivisl+n &)
1, z~g,g 5 g 9;vision Z'up p~y
-8 1>,-~'s.~ ~ S.e
I "sP E 4 I
0
~ ~ ~
m WtAN~'fS~SMCTUkTK~STRUHEHTiRM~RVETttAN~ tMEHENT5 I
CHANNE- tOeWPr-m~ONVrrra~
CANNEL- +UNGTRNAt ~ANN% FOfHIHICHMWRVHttANCE
~!EP-FUNCTMN mmmWmON- -+S-R teaEO~
I er v4ee-latm s ee-gostAnueQ HW '4d'Temp~re 8
-TeeperaCure-
I 89 Changes to Technical Specifications on Other Items
90

Subject:

Changes to Technical Specifications for items required for certification The requested changes to Technical Specifications are enclosed. These changes are requested for certification and reflect the Nine Mile Point Unit 2 design.
TABLE -1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS M
ITl FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE I
GATI

8. Scram Discharge Volume Water C) Level - High M

4g,+

a. Level Transmitter/Trip in. <49.5 in.

Unit
. b. Float Switch <48.5 in. <49.5 in.

9. Turbine Stop Valve - Closure <5X closed <7X closed

10. Turbine Control Valve Fast >530 psig >465 psig Closure, Trip Oil Pressure-Low ll. Reactor Mode Switch Shutdown Position

12. Manual Scram NA

'2 BASES FOR LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2. 1 (Continued)

5. Main Steam Line Isolation Valve - Closure The main steam line isolation valve (MSIV) closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIVs are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, h>g earn tunnel temperature, and low steam line pressure. The MSIV's closure scram anticipates the pressure and flux transients that could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal/hydraulic Safety Limits.

6. Main Steam Line Radiation - Hi h The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected, a trip is initiated to reduce the continued failure of fuel cladding. At the same time the main steam line isolation valves are closed to limit the release of fission products.

The trip setting is high enough above background radiation levels to prevent spurious trips, yet low enough to promptly detect gross failures in. the fuel cladding.

7. Dr ell Pressure - Hi h High pressure in the drywell could indicate a break in the primary pressure boundary systems. The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant.

The trip setting was selected as low as possible without causing spurious trips.
~

8. Scram Dischar e Volume Water Level - Hi h The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume up to a point where there is insufficient volume to accept the displaced water fill at pressures below 65 psig, control rod insertion would be hindered. The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressur QW below'65 psig when they are tripped. The Trip Setpoint for each scram discharge volume is equivalent to a contained volume of approximately~ allons of water. This corresponds to a level indicating switch reading of-46-.7-inches above an instrument zero level of elevation 263 feet 10 inches. 43,+

Turbine Sto Valve - Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves. With a C
NINE MILE POINT - UNIT 2 B2-8
93 BASES FOR LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2. 1 (Continued) trip setting of 5X of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient.,

10. Turbine Control Valve Fast Closure Tri Oil Pressure - Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection with or without coincident failure of the turbine bypass valves. The reactor protection system initiates a trip when fast closure of the control valves is initiated by the fast-acting solenoid valves and in less than 30 milliseconds after the start of control valve fast closure. This is achieved by the action of the fast-acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure switches whose contacts form the one-out-of-two-twice logic input to the reactor protec-tion system. This trip setting, a slower closure time, and a different valve characteristic from that of the turbine stop valve, combine to produce transients which are very similar to that for the stop valve. .Relevant transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis Report.

11. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position provides additional manual reactor trip capability.

12. Manual Scram The manual scram pushbutton switches provide a diverse means for initiating a reactor shutdown (scram) to the automatic protective instrumentation channels and provides manual reactor trip capability.

~F4 ~,~~~ ~~ Q Sc '~~
4~+
ddkW <4
~ ~3 e.VZ NINE MILE POINT - UNIT 2 B2-9 JN 85 586
94 I
POWER DISTRIBUTION LIMITS 3/4. 2.2 E ~ AVERAGE POWER RANGE MONITOR SETPOINTS LIMITING CONDITIONS FOR OPERATION 3.2.2 The Average Power Range Monitor (APRM) flow-biased simulated thermal power-upscale scram trip setpoint (S) and flow-biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:
TRIP SETPOINTV ALLOWABLE VALUEt S < (0.66 (W-hW) + 51X)T S < (0.66 (W-nW) + 54K)T SRB
< (0.66 (W-hW) + 42X)T SRB
< (0.66 (W-hW) + 45M)T where:
S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million lb/hr.
T = The ratio FRACTION OF RATED THERMAL POWER divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY.
T is applied only if less than or equal to 1.0.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or E T E P EE.
ACTION:
With the APRM flow-biased simulated thermal power-upscale scram trip setpoint and/or the flow-biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or SRB to be consistent with the Trip Setpoint value within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
" With CMFLPD greater than the FRTP rather than adjusting the APRM setpoints, the APRM gain may be adjusted so that APRM readings are greater than or equal to 100X times CMFLPD provided that the adjusted APRM reading does not exceed lOOX of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel. ~a~ RodBlo~ k erg g The Average Power Range Monitor Scram FunctionsvaHes as a function of recirculation loop drive flow (W). hW is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between two loop and single loop operation at the same core flow.
hW = 0 for two loop operation. hW = 5X for single loop operation.
NINE MILE POINT - UNIT 2 3/4 2-5
~,'W
~ V4
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEH INSTRUHENTATION APPLICABLE HINIHUH OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEH (a) ACTION

5. Hain Steam Line Isolation Valve-Closure 1(e)

6. Hain Steam Line Radiation-High 1, 2(d)

7. Drywell Pressure - High 2(r) 2(g)

8. Scram Discharge Volume Water Level - High

a. Transmitter Trip Units l. 2 5(h)

b. Float Switches 1, 2 5(h)

9. Turbine Stop Valve - Closure 4(i)

10. Turbine Control Valve Fast Closure, Valve Trip System Oil Pressure - Low 2(J) ll. Reactor Hode Switch Shutdown Position 1, 2 1 3, 4 7 5 .3

12. Hanual Scram 1, 2 3, 4 5

3.3.
TABLE (Continued)
ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM APPLICABLE OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL(a) PER TRIP SYSTEM(b) CONDITION ACTION

1. Primar Containment Isolation Si nals (Continued)

h. SGTS Exhaust - High Radiation 1 2 3 27

i. RMCU System

1) h,Flow - High 6,7 1,2,3 22

2) BFlow - High, Timer 6,7 1 2 3 22

3) Standby Liquid Control, SLCS, Initiation 6(f),7(f) 1 1, 2, 5t 22

~~g/~

j. CM~uipmentMrea Temperature Hi-gh

~X+ARB-Pump-Rooms) s-lz~
~peratur ~~

k. Reactor-BuAdmg-P-ice-Chase-1.

~
Reactor Building Temperature High 3/+ 3"Ha.
5,10 1 2 3 22 S

m. Manual Isolation Pushbutton I.NSSS] 1 1, 2, 3 25 A

2,4,5 26 5 3,6,7 1, 2, 3 26 8 1, 2, 3 25,27 9 1, 2, 3 27

2. RCIC Isolation Si nals

a. RCIC Steam Line Flow - High, Timer 10 1 2 3 22

b. RCIC Steam Supply Pressure - Low(g) 10, ll 1 2.3. 22 C. RCIC Steam Line Flow - High 10 1 2.3 22

d. RCIC Turbine Exhaust Diaphragm Pressure - High(g) 10 1 2 3 22

I I SO VLO N RTRVmE. 'TALON APAJC.A'aL.C MAL WG Knack. OPeaAVL 0~ e At"T'tO<
~O~

1. Peeirneeeev Cour*a roam J ~reou R>sueeu.c(cour)

~. Q 4) C U E ~L PmL= tuV ARED

1. PumP 4o~ A~emPaIt.A'TU~e. - HLSH 4,1 1,'2P 22 Z. PUrn&RloavnR 1KmPBR,MvvE, QLGH 1,2,S 22

s. HX 'Boo~ )e~emc~uze- Hie~ Q,l 1,2,3 22 Q. ~ /~i AC.%OR. SLOE PL PG C HARK m m'> I >o (c~<R), Tarsperate(ra. l4()4 S)&) 1, 1 o 1 I,23 Z. A&1 ~ UTER'I PO Ql OILIER) Te~perofgylp 81)4 6, b,'f,10 z 1,2)3

'3. QWLrn u7H }0 ~ T<LLL~a&m. 1+Lgg 5,4,, t,10 I
TABLE 3.3.2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

1. Primar Containment Isolation Si nals (Continued)

a. Reactor Vessel Water Level"

1) Low, Low, Low, Level 1 >17.8 in. >10.8 in.

2) Low, Low, Level 2 >108.8 in. >101.8 in.

3) Low, Level 3 >159.3 in. >157.8 in.

b. Drywell Pressure - High <1.68 psig <1.88 psig

c. Main Steam Line

1) Radiation - High <3x Full Power Background <3.6x Full Power Background

2) Pressure - Low >766 psig >746 psig .

3) Flow - High <103 psid <109.5 psid

d. Main Steam Line Tunnel 1) 2)

3)
Temperature Temperature
- High temperature - - High High MSL Lead Enclosure
<175
<50
<140'F F
F < ~
<181 F
&2.8'F
~50~ ~

e. Condenser Vacuum Low >8.5 in Hg vacuum >7.6 in. Hg vacuum

f. RHR Equipment Area Temperature - High <135'F <+PP ~<+.S'F (HXs/A8B Pump Rooms)

g. Reactor Vessel Pressure - High <128 psig <148 psig (RHR Cut-in Permissive) CM~

h. SGTS Exhaust - High Radiation ~&60~ pCi/cc ~&6 pCi/cc i5:7x'/0 ~ l >XIO-g.

1 TABLE 3.3.2-2 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE

1. Primar Containment Isolation Si nal s (Continued)

i. RWCU System z5 Flow - High <150.5 gpm <165.5 gpm h Flow - High, Timer

~ ~
~
<45 sec <47 sec o Standby Liquid Control, SLCS, Initiation
~ ~
Area Femperature=. H~
WWCU-Equ-ipment
~HXs&CB-Pump-Rooms-)- ~ ~~+ pf~ p g-<

k. -Reactor

~ ~%1 ~ </g Bui-l-kg-P-ipe-Cha
~mperatur e Hi-gh- 5-<S~
l3o,k IZ+

1. Reactor Building Temperature High <~ F <-38~ F

m. Manual Isolation Pushbutton [NSSS] NA

2. RCIC Isolation Si nals

a. RCIC Steam Line Flow - High, Timer >3 sec, <13 sec 13 sec

b. RCIC Steam Supply Pressure - Low >75 psia >70 psia

c. RCIC Steam Line Flow - High <184. 5 in. H~O"" <193.0 in. H~O""

d. RCIC Turbine Exhaust Oiaphragm Pressure - High <10 psig <20 psig

e. RCIC Equipment Area Temperature - High <135 F <+R'F lho,S

h
3.3 e wkly o Ac.To w>ou T M. u fne.N Amo
> 2 thlTs

1. 'PR<ynagy Co~ve <~eeet'~c~avio< Siaweia(eogv<Nueo)

R(A l" O E QoLPrnE,NT'REA

1. POFAp k(bO~ 9 I GFAp+Q,AVUQ.E'- Qt& H 2 TuvnP Rloorr

~ 6 ~KmPE'R~~ -AUGH ~ 15cPp-HX loom RrnPGRAT'oleic QtG8 135 F ROC~ ~C Pi Pe CV~~e

l. 4 BIYAUtH I R'o (<PPEv) TRNlplKT'c(hL~ glfg 1't+.5 t=

O Z, A+i~u YH 1 VO (QouoE~) TO&I 47efllVQ, fflf4 lw.g r-A&l ~o-w 'to, TiepQra+cre, - H>g 4 )qo.5 I=
TABLE 3.3.2-2 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ALLOMABLE TRIP FUNCTION TRIP SETPOINT VALUE

2. RCIC Isolation Si nals (Continued)

f. RCIC Steam Line Tunnel Temperature - High <135 F L to.S '~

g. manual Isolation Push Button [RCIC] NA

h. Orywell Pressure - High <1.68 psig <1.88 psig

i. RHR/RCIC Steam Flow - High <96 in. HzO"" <104.5 in. H~O"*

3. Secondar Containment Isolation Si nals

a. Reactor Building Above the Refuel Floor Q.$ 4 Q..+4 Exhaust Radiation - High <+;7-x 10-~ pCi/cc <2-.05 x 10-~ pCi/cc

b. Reactor Building Below the Refuel $ ,3$ ~,+to Floor Exhaust Radiation - High <1-.7- x 10-~ pCi/cc <~& x 10-~ pCi/cc

" See Bases Figure B3/4 3-1.
"* Preliminary setpoint - actual setpoint to be determined during startup test program and submitted in writing within 90 days of their determination..
~'
TABLE 4. 3. 2. 1-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREMENTS OPERATIONAL CONDITIONS FOR WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE TRIP FUNCTION CHECK FUNCTION TEST CALIBRATION IS RE UIRED

1. Primar Containment Isolation Si nals (Continued)

h. SGTS Exhaust - High Radiation NA R 1 2 3

i. RWCU System

1) hFlow " High S R 1, 2, 3

2) hFlow - High, Timer NA R 1, 2, 3

3) Standby Liquid Control, SLCS, NA NA 1, 2, 5tt Initiation (iiÃ7%8rBWum~oom&j ~

QW~uipmentMrea Temperatu
~ g/~
k.
temperature
~actor-BuH-ding-P-i-pe-eh Ngh-

1. Reactor Building Temperature High S- M R(b) 1, 2, 3 S 'A

~Pj]
Stkazgcg

m. Manual Isolation Pushhutton INSSS) NA . M(c) 1, 2, 3

2. RCIC Isolation Si nals paaeea

a. RCIC Steam Line Flow - High, Timer NA 1, 2, 3 Low

b. RCIC Steam Supply Pressure R(a) 1, 2, 3

c. RCIC Steam Line Flow - High R(a) 1, 2, 3

d. RCIC Turbine Exhaust Diaphragm Pressure High R(a) 1, 2, 3

II if ff il t
WASw f.3. Z. l-TS<LATl04 A<ToAT LOW ZN'ST RONENTP Tiara 50'R'VGlL~ANc.g Mgoin.grnewvs OP CONo.
C4a<Ne PcR. 4)a~
puuemava,~ C.gANN& S'uRv lS
~'R iP t=o Hc Tl O~
l.PRI TEARY CON TAl t4 MSNTK'Ro<ATlau SLG NITS (MH llnueo)

y. RVJC U EQolPMKHm gQ.K+

1. VU~P Roe~ A l E ~PBQATOAE, HlGQ S' 'R(g) 1)Z)3

~. Pomp Roc rn S TE,~@GOAT v%c - ALGA WC~) 1 )Z)3
>. HX Wi~ lr.~r~h~~e.- Aiaq S v-Cb) l)z)3 J4. RC<CTnR. hulL OlSt" 'PiPE'gASe.
1- Aai mo >h 'PcP CUPP6a), Taepe~fz~ P S
'1 l(g R(s) 1) ZP

z. Ka> vnuwH I R> Cwowaa) Temp~atgra - S Z(n) 1,2,3

~. ~>> ~~~4 ~o, Hi/
. Empt,ratur< W-)4 S W(b> l,2',3
TABLE 3.3.7.1-1 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP INSTRUMENTATION OPERABLE CONDITIONS . SETPOINT (a) ACTION

1. Hain Control Room 2/System(b) 1, 2, 3, 5, and * <~x!i < pCi/cc(c)

Ventilation Radiation 583.x, <o~
Monitors

2. Area Monitors

a. Criticality Monitor <1. OxlO~ mR/hr(d) 76 (New Fuel Storage Vault)

b. Control Room Direct At all times <2. 5x10-~ mR/hr(d) 76 Radiation Honitor

CONTAINMENT SYSTEMS f'JAPAN. 89, PRIMARY CONTAINMENT PRIMARY CONTAINMENT PURGE SYSTEM SURVEILLANCE RE UIREMENTS 4.6.1.7.2 At least once per 92 days each 12- and 14-inch drywell and suppres-sion chamber purge supply and exhaust isolation valve'ith resilient material seals shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 4.38 scf per hour pet 14-inch valve and 3.75 scf per hour per 12-inch valve when pressurized to Pa, 39.75 psig. W~~~ ~~~~ z<~
4I40 EXHPIOST K EDL-AT ldll VAI 'vE'8 LiSfED ON ~Aau= 3.4. l.2-l soAu Ra. %Re.ssom~c ~ I fo. o,'z >N, NINE MILE POINT - UNIT 2 3/4 6-13 JUN 85 58S
0 C)ae ct& tots<< lait 7 alla vsatC 4 7Q)a less<<l
~ e>>eaelw eels Ir 15 QI Isr/
~~ .IS ~ 'at I' esto vsa a I\ee >>ewwa es ~e aaa 1st ele Ct aewac I ~ Ie sae>>w 1 lao aa: ss as ~ v>> ~ IQ ~ ~ Qelveart
)a I I sle 51<<sel << 1 as SWICe
<> I r>>>> I I Q Ie AI>>>> S>>e 1+4>> "
~ aw ~ et I ICNSNCJ r57 scessl Cee>>
as 4al w~ 7)555555 )r CABAL
~
6<<H tlat rt,s ~ &I I
~ ~
fwla<<5.)75 4>> are <<55am Iv ~ ~ ~ 4 ~5 goer yr 4155 le asssal Ct' Saa ape>> 7IE aa C>CCI )~ ~ ts 7 Iae Ie IC5 4,) Is a IIW ~ ew a era Isaac Otee 5 71l l>>eas>>r twe>>s~e sa aaa
~ 4 Sl, I ~e
~
DW
~ tIC<<e see<<le a ree 'I ~ I'l r'1 ~ saao
~ IICJ Cs a>>a w W~
act Csaasl lass Cloeah ay W ~ lc
~ a Ce ~ SC>>5
~ Iee aeCC ~ I at r 57
~ a 4 aeaC
~C ~
asses a>>est sa 4 s<<a>>a s I la>> e Iea ~ 714<<>>e114 4\I ras4 ~ 5 rae aa<< Cta a ~ easel Ceasel C) seel C la W Ial ~4 c~
~ a>>ae)eaa als CC
~
l Iree>> Q W se I'4
~ 1 alw a>>la>>

l. ) la Se>>aa il~

\ 77 )55 ee Itv ~ ee ~ seat C L,7 't4 eae ~ Iar Ceca 4 sees s )srasw<<I<<a res IIII) d ca 441 twasw atc ccwt oa))1 aswa cess Clessl aet C)ao ~ n I)5<<ac
'a Q ~ I ssc 5 tao n
~,4 s;I c" ~ Cw>>t aav ~ Iaa>>u ah Csea<< eats) des>>a elaea Otee Ctassa ~ 175VCC
~ eeeC Isa ~ es>>C 4,7 )4
~
e\e ees sae Qar
~ ~ I r
~.
7sr. ~ e>>
)er I Ne Isa 7ta<<e ew IVV Cl'II essa ea) I
~ Ile <<C 4 1 as ~
avv aee ~
llac
~ Iaa
~ eoasl heess't asar<<
dsasa cas>>IC CCees
~ aors Cteea Pla ral ~,taa CC 17
)7 ~ ts 5 la>>a srsa ~ II Ie aaea
~ le 4 C) IQIC 5 7 ala>>Car ~ to 55 I 7) ~ CS Carat ~ e<<SC s.) )a e>>a ta Se relet 4) waco<< ~ lais CIIC ~ assai al 5
~ Ias Se<<
~
4 CVC ee I rr 1>> le ede r lt
~ 'I LI<< IWC ace asl Cts 4 4) CtsaeC '45 aeP S,C Ce 17 ats I stale aoo 1 .Ctec ~
Salas aa S<<>>as M 7. )5 Wa ls>> ee Ie 1>>s I~ ~
M \I I'se e'I << r>>err aae ~ Is >>I
)
~ ~, 1@I M '\1 ~ loaseswl ,sees e ~ sear 5 75 ~ IS llseee as 1>>o 5aa Cata 51 A ~ tv II l>>I II O'e C 1 Iee'I ~ QI Iar I \I 'I I, ~ eel Iv valaass>>al a>>aae>> Sts w
~ I' I ~ Qee aa Iae 5>>s asia 17
~ a'I Sl Is aea ~ Iss I apl hae IC Ine I>> s 'Ie ~
54 oslalswst )n ~ Qe 1 as ee Ine-s 4 d 74
'I J
(el C
TABLE 3.6.3-1 (Continued)
PRIHARY CONTAINMENT ISOLATION VALVES ISOLATION VALVE ISOLATION MAXIMUM CLOSING VALVE NO. VALVE FUNCTION GROUP SIGNAL(a) TIME (SECONDS) 2CPS*SOV132 Nitrogen to 2CPS"AOV107 Outside IV B,F,Y,Z,RH 2CPSASOV133 Nitrogen to 2CPS"AOV109 Outside IV B,F,Y,Z,RM 2LHS"SOV152(i) LHS from Drywell Inside IV B,F,Z,RH 2LMS"SOV153(i ) LHS from Drywell Outside IV B,F,Z,RH 2LMS"SOV156(i ) LHS from SP Inside IV B,F,Z,RM 2LHS"SOV157(i) LHS from SP Outside IV B,F,Z,RH 2RCSASOV65 A,B(1) Hyd. Unit to RCS FCVs Outside IV's B,F,Z,RH 2RCS"SOV66 A,B(l) Hyd. Unit to RCS FCVs Outside IV's B,F,Z,RM 2RCS"SOV67 A,B(l) Hyd. Unit to RCS FCVs Outside IV's B,F,Z,RH 2RCSASOV68 A,B(l) Hyd. Unit from RCS FCVs Outside IV's B,F,Z,RH 2RCS*SOV79 A,B(1) Hyd. Unit to RCS FCVs Inside IV's B,F,Z,RM 2RCS*SOV80 A,B(l) Hyd. Unit to RCS FCVs Inside IV's B,F,Z,RM 2RCS"SOV81 A,B(1) Hyd. Unit to RCS FCVs Inside IV's B,F,Z,RH 2RCSASOV82 A,B(1) Hyd. Unit from RCS FCVs Inside IV's B,F,Z,RH 2 ICS"HOV121 RCIC Steam Supply Outside IV 10 K,H,H,Z,RH,BB,CC,DD 14 2ICS"HOVlg8(n) RCIC Steam Supply Inside IV 10 K,H,H,QRH,BB,CC,DD 14 2ICS"HOV170 RCIC Warmup Valve Inside IV 10 K,H,H,V,RH,BB,CC,DD 10 2WCS*HOV102 WCS Supply from'RCS 8 RPV Inside IV B,J,U,S,Z,RM,DD 14 Outside IV ~ e5 2WCS"HOV112 WCS Supply from RCS 8 RPV B,J,U,S,Z,RH,DD F 14 2 I CS"MOV148 RCIC Vacuum Breaker Outside IV H 8 F, RH 18 pa~
2NMS"SOV1 A, B, Traversing Incore Probe B,F,Z,RH C, D, E Ball Outside IV's 2GSN*SOV166 Nitrogen Purge to TIP Indexing B,F,Z,RM Mechanism Outside IV
108 PLANT SYSTEMS PLANT SERVICE WATER SYSTEM PLANT SERVICE WATER SYSTEM " OPERATING LIMITING CONDITIONS FOR OPERATION

3. 7. 1. 1 (Continued)

ACTION:

f. With less than the required Division I and Division II heaters OPERABLE within one hour initiate action to 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 and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS 4.7. 1.1.1 The plant service water system shall be demonstrated OPERABLE:
a ~ By verifying the plant service water supply header discharge water temperature to be less than or equal to 76 F:

l. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the last recorded water temperature is greater 'than or equal to 70 F, and

3. At least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when the last recorded water temperature is greater than or equal to 74 F.

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water level at the service water pump intake is greater than or equal to elevation 233. 1 feet.

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

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

l. After a simulated test signal, each automatic valve servicing non-safety-related equipment actuates to its isolation position, en-an NAaV~nhest-s~n~

Each-as~oci-ated service-water system-cross-connect and-pump-des to-supply-fl-ow-to i-~ise+a5ion-pos-i-tion ,and.
aharg~alwe-actuates-au&mahica&y-to
-thata-s-ieg4e-serv4ce-water-pump-starts-automati-ca&y-i~ach-di~
-sion-and tha4-@he-assoc~ted-pump-cVNcharge-v~e-reopens-automa-
-t+@ably ,in-order the-system-safety-re+ated- ~
~mponent~
rdsGR,T P86E /0 7->~.
NINE MILE POINT - UNIT 2 3/4 7-2
109

2. After a simulated test signal, each service water system cross connect and pump discharge valve actuates automatically to its isolation position, and

3. For each service water pump, after a simulated test signal, the pump starts automatically and the associated pump discharge valve opens automatically, in order to 'supply flow to the system safety-related components.

3/4 7-2a
I 4k TP e
PLANT SYSTEMS PLANT SERVICE MATER SYSTEM PLANT SERVICE WATER SYSTEM - OPERATING SURVEILLANCE RE UIREMENTS

4. 7.1.1.1. d (Continued) g P. Each pump runs and maintains service water pump discharge pressure equal to or greater than 80 psig with a pump flow equal to or greater than 6500 gpm.

5 g. The resistance to ground is > 28 ohms for each feeder cable hat powers the intake deicing heater systems.
4.7. 1. 1.2 The Intake Deicing Heater System shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the intake tunnel water temperature is greater than or equal to 390F, or

b. At least once per 7 days by verifying that the current of the heater feeder cables at the motor control centers is 10 amps" or more (total for three phases) at > 518 volts per divisional heater in each intake struc-ture.

lck5'f Sic< foRr I8 knot&, LC)Q IC Sy57+tn FO@C7 lomb I 78675 o$ Ae Service ouehor pump ekrtiutj loadie shell be pertorNeJ.
" For 7 heater elements in operation.
NINE MILE POINT - UNIT 2 3/4 7-3
xC I I V
I N g 7:
PLANT SYSTEMS PLANT SERVICE WATER SYSTEM PLANT SERVICE WATER SYSTEM - SHUTDOWN LIMITING CONOITIONS FOR OPERATION 3.7.1.2 (Continued)
ACTION:

f. With less than the required Oivision I and Oivision II heaters OPER-ABLE, suspend CORE ALTERATIONS and all operations that have a poten-tial for draining the reactor vessel.

SURVEILLANCE RE UIREMENTS 4.7. 1.2.1 The plant service water system shall be demonstrated OPERABLE:

a. By verifying the plant service water supply header discharge water temperature to be less than or equal to 76'F:

l. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. At least once pet 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the last recorded water temperature is greater than or equal to 70 F, and

3. At least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months when the last recorded water temperature is greater than or equal to 74'F.

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the water level at the service water pump intake is greater than or equal to elevation 233. 1 feet.

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

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

1. After a simulated test signal earth automatic valve servicing non-safety-related equipment actuates to its isolation position,e~rr mokati~n-teston~

2. Each associated service water system cross connect and pump discharge valve actuates automatically to its isolation position, and that a single service water pump starts automatically in each division and that the associated pump discharge valve reopens automatically, in order to supply flow to the system safety-related components.

ggScaT fW6E, 3/0 l-Sa.
NINE MILE POINT - UNIT 2 3/4 7-5
~
~ a
2, After a simulated test signal, each service water system cross connect and pump discharge valve actuates automatically to its isolation position, and

3. For each service water pump, after a simulated test signal, the pump starts automatically and the associated pump discharge valve opens automatically, in order to supply flow to the system safety-related components.

~
3/4 7-5a
'"4 tJV' 113 PLANT SYSTEMS PLANT SERVICE WATER SYSTEM PLANT SERVICE WATER SYSTEM - SHUTDOWN SURVEILLANCE RE UIREMENTS 4.7.1.2.1. d (Continued)'ach pump runs and maintains service water pump discharge pressure equal to or greater than 80 psig with-each-pump flow equal to or greater than 6500 gpm.
5. The resistance 4o-ground-is 28 ohms or more for each feeder cable that powers the intake deicing heater systems.
I 4.7.1.2.2 The Intake Deicing Heater System shall be demonstrated OPERABLE:

a. At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 's by verifying the intake tunnel water temperature is greater th'an or equal to 39'F, or i

b. At least once per 7 days by verifying that the current of the heater feeder cables at the motor control centers is 10 amps" lofti~ or more (total for 3 phases) at > 518 volts 'per divisional heater in each intake structure.

8. N'ets't ouch par )$ 'ing% s, Lo5ld $ $ 57F pl pudc7iorJ/l-L TGGT~

op Ac service ~+~ p~~i sH~tig skull Le. pargrmek l
" For 7 heater elements in operation.
NINE MILE POINT - UNIT 2 3/4 7-6 JUN 25 1986
114 ELECTRICAL POWER SYSTEMS 3/4. 8. 3
~ ~ ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION - OPERATING LIMITING CONDITIONS FOR OPERATION 3.8.3.1 The following power distribution system divisions shall be ene'rgized with tie breakers open between Division I and Division II buses:

a. AC power distribution

1. Division I, consisting of:

a) 4160-volt AC bus b) 600-volt AC load center/MCCs/distribution panels c) 240/120-volt AC and 120-volt AC distribution panels, energized from inverter 2VBA"UPS2Af

2. Division II, consisting of:

a) 4160-volt AC bus b) 600-volt AC load center/MCCs/distribution panels c) 240/120-volt AC and 120-volt AC distribution panels, energized from inverter 2VBA"UPS2Bf

3. Division III, consisting of:

a) 4160-volt AC bus b) 600-volt AC MCCs/distribution panels c) 240/120-volt AC and 208/120-volt AC distribution panels M)~P~verter-energi-zed-from-0-i-vH-ion-I-H bat teri-es-

b. DC power distribution

1. Division I, consisting of 125-volt DC switchgear, MCC and associated distribution panels: 2BYS"PNL 201A; 2BYS*PNL 202A; 2BYS"PNL 204A

2. Division II, consisting of 125-volt DC switchgear, MCC and associated distribution panels: 2BYS*PNL 201B; 2BYS"PNL 202B; 2BYS"PNL 204B

3. Division III, consisting of 125-volt DC distribution panel~ DCESWXPeL 9/+

-2~NW201C,2BYWPNL 2eZC 2eVS PNL 2e4C APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
i'he UPS shall be energized from their normal AC supply or their backup DC s"pply-NINE MILE POINT - UNIT 2 3/4 8-20
ll 115 ELECTRICAL POWER SYSTEMS ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION - SHUTDOWN LIMITING CONDITIONS FOR OPERATION 3.8.3.2 As a minimum, the following power distribution system divisions shall be energized:

a. For AC power distribution, Division I or Division II, and when the HPCS system is required to be OPERABLE, Division III, with:

1. Division I consisting of:

a) 4160-volt AC bus b) 600-volt AC load center/MCCs/distribution panels c) 240/120-volt AC and 120-volt AC distribution panels, energized from inverter 2VBA".UPS2A or alternate supply

2. Division II consisting of:

a) 4160-volt AC bus b) 600-volt AC load center/MCCs/distribution panels c) 240/120-volt AC and 120-volt AC distribution panels, energized from inverter 2VBA*UPS2B or alternate supply

3. Division III consisting of:

a) 4160-volt AC bus b) 600-volt AC MCCs/distribution panels c) 240/120-volt AC and 208/120-volt AC distribution panels
~~PC~i-nverte~nergized-~m&m~menMH batteri+s

b. For DC power distribution, Division I or Division, II, and when the HPCS system is required to be OPERABLE, Division III, with:

1. Division I consisting of 125-volt OC switchgear, MCC, and distribu-tion panels

2. Division II consisting of 125-volt DC switchgear, MCC, and distribu-tion panels

3. Division III consisting of 125-volt OC distribution panels APPLICABILITY: OPERATIONAL CONDITIONS 4, 5, and ".

" When V
handling irradiated fuel in the secondary containment.
NINE MILE POINT " UNIT 2 3/4 8-22
116 ELECTRICAL POWER SYSTEMS ELECTRICAL E UIPMENT PROTECTIVE DEVICES REACTOR PROTECTION SYSTEM ELECTRIC POWER MONITORING (RPS LOGIC)
LIMITING CONDITIONS FOR OPERATION 3.8.4.4 Two RPS UPS electrical protection assemblies for each inservice UPS set or alternate source shall be OPERABLE.
APPLICABILITY: At all times.
ACTION:

a. With one RPS electrical protection assembly for an inservice RPS UPS inoperable, restore the inoperable electrical protection assembly to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or remove the associated RPS UPS from

'service.

b. W',th both RPS electrical protection assemblies for an inservice RPS UPS inoperable, restore at least one electrical protection assembly to OPERABLE status within 30 minutes or remove the associated RPS UPS from service.

SURVEILLANCE RE UIREMENTS 4.8.4.4 The above specified RPS electrical protection assemblies instrumen-tation shall be determined OPERABLE:

a. At least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST and;

b. At least once per 18 months by demonstrating the OPERABILITY of over-voltage, undervoltage and underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actua-tion of the protective relays, tripping logic and output circuit breakers verifying the following setpoints. 'nd

l. Overvoltage Bus A: < 132 volts AC, volts Ad,~

Bus B: < 132 AC,

2. Ud I 2 2 ~: 117.1 It Bus B: . > 115.75 volts AC, -0~-2-.5X-

1. Ud I 2 7 272*,~

NINE MILE POINT " UNIT 2 3/4 8-32
0 ELECTRICAL POWER SYSTEMS Flf)PL 0<iQ[>>
ELECTRICAL E UIPMENT PROTECTIVE DEVICES REACTOR PROTECTION SYSTEM ELECTRIC POWER MONITORING (SCRAM SOLENOIDS)
LIMITING CONDITIONS FOR OPERATION 3.8.4.5 Two RPS UPS electrical protection assemblies (EPAs) for each inservice RPS MG set or alternate source shall be OPERABLE.
I APPLICABILITY: At all times.
I ACTION:
a.'ith one RPS or alternate power I
electrical protection assembly for an inservice supply'noperable, restore the inoperable RPS MG EPA to set OPERABLE status within 72; hours or remove the associated RPS MG set or alternate power supply from service.

b. With both RPS electrical protection assemblies for an inservice RPS MG set or alternate power supply inoperable, restore at least one EPA to OPERABLE status within 30 minutes or remove the associated RPS HG set or alternate power supply'from service.

SURVEILLANCE RE UIREMENTS 4.8.4.5 The above specified RPS electrical protection assemblies shall be determined OPERABLE:

a. At least once per 6 month's by performance of a CHANNEL FUNCTIONAL TEST and;

b. At least once per 18 months by demonstrating the OPERABILITY of over-voltage, undervoltage and'" underfrequency protective instrumentation by performance of a CHANNEL CALIBRATION including simulated automatic actua-tion of the protective re'lays, tripping logic and output circuit breakers and verifying the following setpoints.

1. Overvoltage Bus A:

Bus B:
I 128.8 volts AC~0
&O~o+~
~%-
L~.o V0l 5'oo-I-WS QC.

2. Undervoltage Bus B

A:: >
B:,,
~
MB:6 (le.l H~~&
3.. Ud f 0 y 57lf*,~ votes A,c.~
NINE MILE POINT " UNIT 2 3/4 8-33
REFUELING OPERATIONS CONTROL R00 REMOVAL MULTIPLE CONTROL ROD REMOVAL LIMITING CONDITIONS FOR OPERATION

3. 9. 10.2 Any number of control rods and/or control rod drive mechanisms may be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core.

ao The reactor mode switch is OPERABLE and locked in the Shutdown position or in the Refuel position per Specification 3.9. 1, except that the Refuel position "one-rod-out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below.

b. The source range monitors (SRMs) are OPERABLE per Specification 3.9.2.

C. The SHUTDOWN MARGIN requirements of Specification 3.I.1 are satisfied.
All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

e. The four fuel assemblies surrounding each control rod or control rod drive mechanism to be removed from the core and/or reactor vessel are removed, from the core cell.

APPLICABILITY: OPERATIONAL CONDITION 5.
ACTION:
With the requirements of the above specification not satisfied, suspend removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and initiate action to satisfy the above requirements.
/
$, All fuel toad;~ epra+ions halva. lo~vl 5zzpzpJ~
NINE MILE POINT " UNIT 2 3/4 9-14
1'p 1
'I a t,r
p REFUELING OPERATIONS h f 119 CONTROL ROD REMOVAL MULTIPLE CONTROL ROD REMOVAL SURVEIL'LANCE RE UIREMENTS 4.9.10.2.1 Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months before the start of removal of control rods and/or control rod drive mechanisms from the core and/or reactor pressure vessel and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter until all control rods and control rod drive mechanisms are reinstalled and all control rods are inserted in the core, verify that:

a. The reactor mode switch is OPERABLE per Surveillance Requirement 4.3. 1.1 or 4.9.1.2, as applicable, and locked in the Shutdown position or in the Refuel position per Specification 3.9.1.

b. The SRM channels are OPERABLE per Specification 3.9.2.

c. The SHUTDOWN MARGIN requirements of Specification'.1.1 are satisfied.

-. d. All other control rods are either inserted or have the surrounding four fuel assemblies removed from the core cell.

e. The four fuel assemblies surrounding each control rod and/or control rod drive mechanism to be removed from the core and/or reactor vessel are removed from the core cell.

4.9. 10.2.2 Following replacement of all control rods and/or control rod drive mechanisms removed in accordance with this specification, perform a functional test of the "one-rod-out" Refuel position interlock, if this function had been bypassed.
p[) age.l loadie'parrows hate. been suspu~ecl.
NINE MILE POINT - UNIT 2 3/4 9-15
C.
b
~ IC' 1 ~
W
120 3/4. 11 RADIOACTIVE EFFLUENTS
-3/4.11.1 LI UID EFFLUENTS CONCENTRATION LIMITING CONDITIONS FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall be limited to the concentrations specified in 10 CFR 20, Appendix 8, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-~ microcurie/ml total activity.
APPLICABILITY: At all times.,
ACTION:
With the concentration of radioactive material released in, liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits.
SURVEILLANCE RE UIREMENTS

4. 11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.:HM.

4. 11.1.1.2 The results of the'adioactivity analyses shall be used in accord-ance with the methodology and parameters in the ODCM to assure that the con-centrations at the point of release are maintained within the limits of Specification 3.11.1.1. I NINE MILE POINT - UNIT 2 3/4 11-1

II 11 I L
121 TABLE ~
~,Ll,l+
RADIOACTIVE LI UID WASTE SAMPLING AND ANALYSIS PROGRAM LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(a)
TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml)

1. Batch Waste P P Principal Gamma 5xlO-7 Release Each Batch Each Batch Emitters(c)

Tanks(b)

a. 2LWS-TK4A I"131 lx10-s

b. 2LWS-TK4B Emitters)'OWER

. c. 2LWS-TK5A

d. 2LWS"TK5B P One Batch/M Dissolved and lx10"s One Batch/M Entrained Gases (Gamma P M H-3 lx10"s Each Batch Composite(d)

Gross Alpha lx10-7 P Sr-89, Sr-90 Sx10-s Each Batch Composite(d)
Fe-55 lx10-6

2. Continuous Grab Sample Grab Sample Principal Gamma 5x10-7 Releases M(e) M(e) Emitters(c)

I-131 Ix10-s

a. Service Water Dissolved and lxlO-s Effluent A Entrained Gases (Gamma Emitters)

b. Service Water H-3 lx10 s Effluent B Gross Alpha lx10"7

c. Cooling Tower Grab Sample Grab Sample Sr-89, Sr-90 5xlO s Blowdown Q(e) Q(e)

Fe"55 lx10-s NINE MILE POINT - UNIT 2 3/4 11-2
y "m,t Vi f,
V.
a
122 TABLE ~~
4)lid-i (Continued)
RADIOACTIVE LI UID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (a) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95K probability with only 5X probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
4.66 sb LLD ED V-2. 22x106 Y.exp(-A4t)
Where:
LLD = the before-the-fact lower limit of detection (microcurie per unit mass or volume),
sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),
= the counting efficiency (counts per disintegration),
= the sample size (units of mass or volume),
2.22x106 = the number of disintegrations per minute per microcurie,
= the fractional radiochemical yield, when applicable,
'= the radioactive decay constant for the particular radio-nuclide (sec-~), and
= the elapsed time between the midpoint of sample collection and the time of counting (seconds).
Typical values of E, V, Y, and ht should be used in the calculation.
It should be recognized that the LLD is defined as a before-the-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measurement.
(b) A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM to assure representa-tive sampling.
NINE MILE POINT - UNIT 2 3/4 11-3
e 4
I
~C
)e
123
~,i<,i-)
TABLE ~l~ (Continued)
RADIOACTIVE LI UID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (c) The principal gamma emmiters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137 and Ce-141. Ce-144 shall. also be measured, but with an LLD of 5 x 10-e. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.8 in the format outlined in RG 1.21, Appendix B, Revision 1, June 1974.
(d) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.
(e) If the alarm setpoint of the effluent monitor, as determined by the method presented in the ODCM, is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists. Frequency of analysis shall be increased to daily for principal gamma emitters and an incident composite for H-3, gross alpha, Sr-89, Sr-90, and Fe-55.
NINE MILE POINT " UNIT 2 3/4 11-4
124 RADIOACTIVE EFFLUENTS 3/4.11.2
~ ~ GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITIONS FOR OPERATION 3.11.2.1 The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than 'or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, for iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.
ACTION:
With the dose rate(s) exceeding the above limits, immediately restore the release rate to within the above limit(s).
'SURVEILLANCE RE UIREMENTS 4.11.2.1.1 The dose rate from noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

4. 11.2. 1.2 The dose rate from iodine-131, iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis pro'gram specified in. Table <M&&.

4, ll.k-i NINE MILE POINT - UNIT 2 3/4 11-8
C 125 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12.
~ ~ 1 MONITORING PROGRAM LIMITING CONDITIONS FOR OPERATION

3. 12.1 The Radiological Environmental Monitoring Program shall be conducted as specified in Table H~

S. lk il-)
APPLICABILITY: At all times.
ACTION:
ao With the Radiological nvironmental Monitoring Program not being conducted as specified in Table ~f1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Speci-fication 6.9. 1.7, a description of the reasons for not conducting the pro-gram as required and the plans for preventing a recurrence.
5,'g. l - Q.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the re-porting levels of Table H~~ when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Specifi-cation 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose" to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifica-tions 3.11.1.2, 3. 11.2.2, or 3.11.2.3. When more than one of the radio-

~2&
nuclides in Table shall be submitted if: ~ are detected in the sampling medium, this report G, Q,)-g.
concentration 1 concentration 2 reporting level 1 reporting level 2 s.>z,l-Z When radionuclides other than those in Table<..12~ are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose" to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of Specifica-tion 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Specification 6.9.1.7.
" The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
NINE MILE POINT - UNIT 2 3/4 12-1
pQ
~'
I pa J"
126 RADIOLOGICAL ENVIRONMENTAL MONITORING MONITORING PROGRAM I
LIMITING CONDITIONS FOR OPERATION

3. 12. 1 (Continued)

ACTION:

c. Mith milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table-3%&1; identify specific loca-tions for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program. The speciffc locations from which samples were unavailable may then be deleted from the monitor-ing program. Pursuant to Specification 6.9. 1.8, submit in the next Semi-annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location(s) for obtaining samples.

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

SURVEILLANCE RE UIREMENTS 4.12.1 pursuant to Table ~~
The radsologscal environmental monitoring samples shall be collected from the specific locations given in the table and figure(s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 1 and the etection capabilities required by Table ~R=:l-.
4(ia l -1 NINE MILE POINT - UNIT 2 . 3/4 12-2
M P
I qI L
127 TABLE 3.12.1-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATIONS (a) Specific parameters of distance and direction sector from the enterline of one reactor, and additional description where pertinent hall be pro-vided for each and every sample location in Table -B-.MW in a table and figure(s) in the ODCM. Refer to NUREG-0133, "Preparation of Radiol'ogical Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position. on Environmental Monitoring, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable because of such circumstances as hazardous conditions, seasonal unavailability," or mal-function of automatic sampling equipment. If specimens are unobtainable because sampling equipment malfunctions, effort shall be made to complete corrective action before the end of the next sampling period. All devia-tions from the sampling schedule shall be documented in the Annual Radio-logical Environmental Operating Report pursuant to Specification 6.9. 1.7.
It is recognized that, at times, may not be possible or practical to it continue to obtain samples of the media of choice at the most desired loca-tion or time. In these instances, suitable alternative media and locations may be chosen for the particular pathway in question and appropriate sub-stitutions may be made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to Specification 6.9. 1.8, submit in the next Semiannual Radioactive Effluent Release Report a revised figure(s) and table for the ODCM reflecting the new location(s) with sup-porting information identifying the'ause of the unavailability of samples for that pathway and justifying the selection of new location(s) for obtaining samples.
(b) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be 'used in place of, or in addi-tion to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.
Film badges shall not be used as dosimeters for measuring direct radiation.
(c) The purpose of these samples is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites, which provide valid background data, may be substituted.
(d)' Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the previous yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
(e) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
"Seasonal unavailability is meant to include theft and uncooperative residents.
NINE MILE POINT - UNIT 2 3/4 12-8 gUN 2s >9aG
128 TABLE 3:12.1-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATIONS The "upstream" sample shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall be taken in an area beyond but near the mixing zone.
(g) In this program, representative composite sample aliquots shall be col" lected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a re-presentative sample (refer to the ODCM for definition of representative composite sample).
(h) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination (see ODCM for discussion).
Drinking water samples shall be taken only when drinking water is a dose pathway (see ODCM for discussion).
Analysis for I-131 may be accomplished by Ge-Li analysis provided that the lower limit of detection (LLD) for I-131 in water samples found on Table 4-.3;,~can be met. Doses shall be calculated for the maximum organ and agefgroup; using the methodology in the ODCM.
~A Q..)-l (k) In the event two commercially or recreationally important species are not available, after three attempts of collection, then two samples of one species or other species not necessarily commercially or recreationally important may be utilized.
This specification applies only to major irrigation projects within 9 miles of the site in the general "downcurrent" direction (see ODCM for discussion).
(m) If harvest occurs more than once a year, sampling shall be performed dur-ing each discrete harvest. If harvest occurs continuously, sampling shall be taken monthly. Attention shall be paid to including samples of tuberous and root food products..
=
NINE MILE POINT " UNIT 2 3l4 12 9 Qgl P.~
0 129 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITIONS FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a dis-tance of 5 miles the location in each of the 16 meteorological sectors of the nearest milk animal and the nearest residence, and the nearest garden" of greater than 500 square feet producing broad leaf vegetation. For elevated releases as defined in RG 1.111, Revision 1, July 1977, the land use census shall also identify within a distance of 3 miles the locations in each of the 16 meterological'sectors of all milk animals and all gardens* greater than 500 square feet producing broad leaf vegetation.
APPLICABILITY: At all. times.
ACTION:
a0 Mith a land use census identifying a location(s) that yields a calculated dose, dose commitment, or D/g value greater than the values currently being calculated in Specification 4. 11. 2. 3, pursuant to Specification 6. 9. 1. 8, identify the new location(s) in the next Semiannual Radioactive Effluent Release Report.
3,ix,i-i With a land use census identifying a location(s) that yields a calculated dose, dose commitment, or 0/g value (via the same -exposure athway) signi-ficantly greater(50K)than at a location from which samples are currently being obtained in accordance with Specification&~M, add the new loca-tion(s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The sampling location(s), excluding the control station location, having the lowest calculated dose, dose commitment(s) or 0/g value, via the same exposure pathway, may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.
Pursuant to Specification 6.9. 1.8 submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table(s) for the ODCM reflecting the new location(s) with information supporting the change in sampling locations.
C. ,The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
" Broad leaf vegetation sampling of at least three different kinds of vegeta-tion, such as garden vegetables, may be performed at offsite locations in each of two different locations with the highest predicted 0/gs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table %~-1-, Part 4.c, shall be followed, including analysis of control 5 8 lllp I e 5 .
NINE MILE POINT " UNIT 2 3/4 12"14
130 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12.3
~ ~ INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITIONS FOR OPERATION 3.12.3 Analyses shall be performed on all radioactive materials, supplied as part of Commission, that correspond to samples required by Table ~-
an Interlaboratory Comparison Program that has been approved by the
. Participa-tion in this program shall include media for which environmental amples are routinely collected and for which intercomparison samples are available.
APPLICABILITY: At al 1 times.
ACTION:

a. Mith analyses not being performed as required above, report the correc-tive actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specifi-cation 6.9. 1.7.

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

SURVEILLANCE RE UIREMENTS 4.12.3
~ ~ The Interlaboratory Comparison Program shall be described in the ODCM.
A summary of the results obtained as part of the above required Interlaboratory Comparison Program, shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6. 9.1.7.
NINE MILE POINT - UNIT 2 3/4 12"16
131 REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued}
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and, therefore, this check must be performed before achieving criticality after
-completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and, therefore, that other parameters are within their limits, the control rod position indica-tion system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 6 inches in the event of a housing failure. The amount of rod reactivity that could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and yet limited in frequency to avoid causing excessive wear on the system components.
3/4. 1.4 CONTROL ROD PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum in-sequence individual control rod or control rod segments that are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal/gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 20K of RATED THERMAL, POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal/gm. Thus requiring the RSCS and RWM to be OPERABLE when THERMAL POWER is less than or equal to 20K of RATED THERMAL POWER provides adequate control.
The RSCS and RWM provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4 of the FSAR the techniques of the analysis are presented in a topical report (Ref-
~
and erence 1} and two supplements (References 2 and 3}.
designed to automatically prevent fuel damage in the event of t
'The RBM is ~
erroneous rod withdrawal from locations of high power density during high-power
~
NINE MILE POINT - UNIT 2 B3/4 1"3
l 13,2 REACTIVITY CONTROL SYSTEMS ASES CONTROL ROD PROGRAM CONTROLS 3/4. 1.4 (Continued) operation. Two channels are provided. Tripping one .of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. T&~sys4em-back~
~p-4he-wH-iten-sequence-used-by-the-operator-f~r-withdraw s-.
3/4. 1.5 STANDBY LI UID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern. To meet this objective, it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel. To allow for potential leakage and imperfect mixing, this con-centration is increased by 20K. The required concentration is achieved by having a minimum available quantity of 4418 gallons of sodium-pentaborate solution con-taining a minimum of 5500 lb of sodium-pentaborate. This quantity of solution is' net amount which is above the pump suction, thus allowing for the portion that cannot be injected. The minimum pumping rate of 41.2 gpm per pump provides
~
~
~ ~
negative reactivity insertion rate over the permissible penetaborate solution
~ ~ ~ ~
olume range, which adequately compensates for the positive reactivity effects
~ ~
from temperature and xenon during shutdown. The temperature requirement is
~ ~
necessary to ensure that the sodium pentaborate remains in solution.
~ ~ ~
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time when the system is inoperable or for longer periods of time when one of the redundant components is inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concen-tration will not vary unless more boron or water is added; thus a check on the temperature and volume once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular i.ntervals will assure that these valves will not fail because of deterioration of the charges.

References:

C. J; Paone, R. C. Stirn, and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," GE Topical Report NED0-10527, March 1972.
C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NED0-10527, July 1972.

3. J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, "Exposed Cores,"

Supplement 2 to NED0-10527, January 1973.
NINE MILE POINT - UNIT 2 83/4 1"4
0 133 3/4. 2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature the postulated design basis loss-of-coolant accident will not exceed 'ollowing the 2200'F limit specified in 10 CFR 50.46.
3/4.2. 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature,(PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure-dependent steady-state gap conductance and rod-to-rod local peaking factor. 1-he-Teehni~-Speci+~he WVERAG~LANAR-LINEAR-HEAT-GENERATIeN-RA-FE-(APLHGR) mthe-LHBR-o fthe-highest
-powered-rod-deeded-by it~ca}-peaking faetm". The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3 for two-recirculation-loop operation.
The calculational procedure used to establish the APLHGR shown on Figures

3. 2. 1-1, 3. 2. 1-2, and 3. 2. 1-3 is based on a loss-of-coolant accident analysis.

~ ~ ~ ~ ~
~
The analysis was performed using General Electric (GE) calculational models
~ ~
~ ~
which are consistent with the .requirements of Appendix K to 10 CFR 50. A
~ ~
~
complete discussion of each code employed in the analysis is presented in
~ ~ ~ ~
Reference l.~ Differences in this analysis compared with previous analyses can
~ ~ ~ ~ ~
be broken down as follows.
~lt. Cl Corrected vaporizati'on calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.

2. Incorporated more accurate bypass areas - The bypass areas in the top guide were recal,culated using a more accurate technique.

3. Corrected guide tube I thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

b. ~Nd I Cd

1. Core CCFL pressure differential, 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1-psi pressure drop in core.

2. Incoporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.

NINE MILE POINT - UNIT 2 B3/4 2-1
134 INSTRUMENTATION BASES 3/4. 3. 4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATMS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an antic'ipated transient. The response of the plant to this postulated event falls within the envelope of study events in General
.Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979; and Section 15.8 of the FSAR.
The end-of-cycle recirculation pump trip (EOC-RPT) system is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end of -cycle. The physical phenomenon involved is that the void reactivity feedback from a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each EOC-RPT system reps o h recirculation pumps., reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the othe", two turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast
.closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.
I Each EOC-RPT system may be manually bypassed by use of a switch which is administratively controlled by procedures. The manual bypasses and the auto-matic Operating Bypass at less than 30K of RATED THERMAL POWER are annunciated in the control room.
The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i. e.,
190 milliseconds. Included in this time are:. the time from initial valve movement to reaching the Trip Setpoint, the response time of the sensor, the response time of the system logic, and the time alloted for breaker arc supression; Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference NINE MILE POINT - UNIT 2 B3/4 3-3
135 CONTAINMENT SYSTEMS BASES PRIMARY CONTAINMENT 3/4.6.1.4 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the design pressure of 45 psig in the event of a loss-of-coolant acci-dent (LOCA). A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3/4.6. 1.5 DRYWELL AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 39.75 psig does not exceed the design pressure of 45.0 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 4.7 psi. The limit of 14.2 to ~-psia for initial positive containment pressure will limit the total pressure to 39.75 psig, which is less than the design pressure and is consistent with the safety analysis.
3/4.6.1.6 DRYWELL AVERAGE AIR TEMPERATURE i~,As The limitation on drywell average air temperature ensur es that the containment peak air temperature does not exceed the design temperature of 340'F during steam line break conditions and is consistent with the safety analysis.
In addition, the maximum drywell average air temperature is also the limiting initial condition used to determine the maximum negative differential pressure acting on the drywell and suppression chamber following inadvertent actuation of the containment sprays.
3/4. 6.1.7 PRIMARY .CONTAINMENT PURGE SYSTEM The 14-inch drywell and 12-inch suppression chamber supply and exhaust valves are limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months of use per 365 days during purge or vent operations in CONDITIONS 1, 2, and 3 to meet the requirements of Branch Technical 'PERATIONAL Position CSB 6-4 for valves greater than 8 inches in diameter. The requirement to limit the opening of 2CPS"AOV105, 2CPS"AOV107, 2CPS"AOV109, and 2CPS*AOV110 to 70 degrees, and 2CPS"AOVlll to 60 degrees ensures these valves will close during a LOCA or steam line break accident, and therefore, the site boundary dose guidelines of 10 CfR 100 would not be exceeded in the event of an accident during purging or venting operations.
NINE MILE POINT - UNIT 2 B3/4 6-2
p 136 CONTAINMENT SYSTEMS BASES PRIMARY CONTAINMENT PRIMARY CONTAINMENT PURGE SYSTEM 3/4. 6. 1. 7 (Continued)
Leakage integrity tests with a maximum allowable leakage rate for purge supply and exhaust isolation valves will provide early indication of resilient mate-rial seal degradation and will: allow the opportunity for repai~ before gross leakage failure develops. The leakage limit shall not be exceeded when the leakage rates are determined to be less than or equal to 4.38 scf/hour per 14-inch valve and 3.75 scf/hour per 12-inch valve+here pressurized to 0'l. lS er 40.0 psig> e s e ppLi c.+bLe, w ken, 3/4. 6. 2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment press-ure will not exceed the design. pressure of'5 psig during primary system blowdown from full operating pressure.
The suppression pool water provides the heat sink for the reactor coolant sys-tems energy release following a postulated rupture of the system. The suppres-sion pool water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1040 psig. Because all of the gases in the drywell are purged into the suppression pool air space during a LOCA, the pressure of the liquid must not exceed 45 psig, the suppres-
~
sion chamber maximum pressure..'he design volume of the suppression chamber (water and air).was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume
.is purged to the suppression chamber.
Using the minimum or maximum water volumes given in this specification, con-tainment pressure during the design-basis accident is approximately 40 psig, which is below the design pres'sure of 45 psig. Maximum water volume of 154,794 cubic feet results in a downcomer submergence of ll feet 0 inch, and the minimum volume of 145,495 cubic feet results in a submergence approximately 18 inches less. The majority 'of the Bodega Bay tests were run with a submerged length of 4 feet and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170 F, and this is conservatively taken to be the limit for complete con-densation of the reactor coolant, although condensation would occur for temperatures above 170'F.
I Should it be
~
necessary to make'he suppression chamber inoperable, this shall
~
detailed in Specification 3.5.3.
~ ~ ~ ~ ~
only be done as ~ ~
~
I NINE MILE POINT - UNIT 2 B3/4 6-3
137 PLANT SYSTEMS BASES CONTROL ROOM OUTDOOR AIR SPECIAL FILTER TRAIN SYSTEM 3/4. 7. 3 (Continued) each 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and high-efficiency particulate air (HEPA) filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of GOC 19 of Appendix A to 10 CFR 50.
3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM
. The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling system (ECCS) equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig. This pressure is substantially below that for which the RCIC system can provide adequate core cooling for events requiring the RCIC system.
The RCIC system specifications are applicable during OPERATIONAL CONOITIONS 1, 2, and 3, when reactor vessel pressure exceeds 150 psig because RCIC is the pri-mary non-ECCS source of emergency core cooling when the reactor is pressurized.
I Mith the RCIC system inoperable, adequate core cooling is assured by the OPERA-BILITY of the HPCS system and justifies the specified 14-day out-of-service peri od.

The Surveillance Requirements provide adequate assurance that RCIC will be OPERABLE when required. -@%hough &11 active components are testable and full flow can be demonstrated by recirculation during reactor operationQ~m~m

~nc&en~tes~eqvQ~wactor shutdown The pump discharge piping is main-tained full to prevent water hammer damage.
3/4.7.5 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event that initiates dynamic loads.,
Snubbers excluded from this inspection program are those installed on non-safety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
The visual inspection 'frequency is based upon maintaining a constant level of Therefore, the required inspection interval
~
snubber protection to systems.
varies inversely with the observed snubber failures and is determined by the
~
number of inoperable snubbers found during an inspection. Inspections performed NINE MILE POINT " UNIT 2 B3/4 7-4
138 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8. 1 3/4.8.2 8( 3/4.8.3 AC SOURCES DC SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the AC and DC power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety"related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant AC and DC power sources and distribution systems satisfy the requirements of GDC 17 of Appen-dix A to 10 CFR 50.
The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least Division I or II of the onsite AC and DC power sources and associated distribution systems OPERABLE during accident conditions coin-cident with an assumed loss of offsite power and single failure of the other onsite AC or DC source. Division III supplies the high-pressure core spray (HPCS) system only.
The AC and DC source allowable out-of-service times are based on RG 1.93, ~
"Availability of Electrical
~ ~ ~ ~ ~
Power Sources," December 1974. When diesel gener-ator ED'Division I) or EDGf3 (Division II) is inoperable, there is an
~ ~ ~ ~ ~ ~ ~
additional ACTION requirement to verify that all required systems, subsystems,
~ ~ ~
trains, components, and devices that depend on the remaining OPERABLE diesel
~ ~ ~
generator EDGkl or ED's a source of emergency power, are also OPERABLE.
This requirement is intended to provide assurance that a loss-of-offsite-power event will not result in a complete loss of safety function of critical sys-tems during the period diesel generator term "verify" ED'r EDGIER is inoperable.
as used in this context means to administratively check by exam-The ining logs or other information to determine if certain components are out of service for maintenance or other reasons. It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.
The OPERABILITY of the minimum specified AC and DC power sources and associated distribution systems during shutdown and refueling ensures that (1) tlie facil-ity can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient. instrumentation and control capability is available for monitoring and maintaining the unit status.
The Survyillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of RG 1.9, "Selection of Diesel Generator Set Capacity for Standby Power Supplies," December 1979; RG 1. 108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977; and RG 1. 137, "Fuel-Oil Systems for Standby Diesel Generators,"'evision 1, October 1979.
NINE MILE POINT " UNIT 2 B3/4 8-1
139 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4. 12. 1 MONITORING PROGRAM The Radiological Environmental Monitor ing Program required by this specification provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of. MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. After this period, program changes may be ini" tiated based on operational experience.
%, g..h-h The required detection capabilities for environmental sample analyses are tabu-lated in terms of e lower limits of detection (LLDs). The LLDs required by Table~1~ re considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before-the-fact limit representing the capability of a measurement system and not as an after the-fact limit for a particular measurement.
Detailed discussion of the LLD, and other detection limits, can be found in L. A. Currie, "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"
NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).
3/4. 12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at or beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by .the results of this census. The best information, such as from a door-to-door survey, from an aerial survey, or from consulting with local agricultural authorities, shall be used. This census satisfies the require-ments of Section IV.B.3 of Appendix I to 10 CFR 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in RG 1.109 for consumption by a child. To deter-mine this minimum 'garden size, the following assumptions were made: (1) 2(C of the garden was used for growing broad leaf vegatation (i. e., similar to lettuce and cabbage) and (2) the vegetation yield was 2 kg/m~.
A MILK SAMPLING LOCATION, as defined in Section 1.0, requires that at least 10 milking
~ ~
cows are present at a designated milk sample location. It has been NINE MILE POINT " UNIT 2 B3/4 12-1
140 ADMINISTRATIVE CONTROLS REPORTING RE UIREMENTS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORTS

6. 9.1.7 (Continued) previous environmental surveillance reports, and an assessment of. the observed impacts of the plant operation on the environment. The reports shall also include the results of the land use census required by Specification 3. 12.2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all-environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the OFFSITE DOSE CALCULATION MANUAL, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some indivi-dual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplemental report.
The reports shall also include the following: a summary description of the Radiological Environmemtal Monitoring Program; at least two legible maps" cover-ing all sampling locations keyed to a table giving distances and directions
~ ~ ~ ~ ~
from the centerline of one reactor; the results of licensee participation in
~
the Interlaboratory Comparison Program, required by Specification 3. 12.3; dis-
~ ~ ~ ~ ~ ~
~~~an
~
cussion of all deviations from the Sampling Schedule of Table
~ ~ ~
~ is-cussion of all analyses in which the LLD required by Table-4H~ was not
~ ~
achievable.
t
+ iX,t-l SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT""
6.9.1.8 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be sub-mitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date the plant achieves initial criticality.
The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste re-leased from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evalua-ting, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power
" One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
"" A single submittal may be made for a multiple unit site. The submittal should combine those sections that are common to all units at the site; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
NINE MILE POINT - UNIT 2 6-20

Subject:

Editoral changes to Technical Specifications The requested changes to Technical Specifications are enclosed. These items are editorial changes and are self explanatory.
INDEX 142 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS PAGE REACTIVITY CONTROL SYSTEMS (Continued) 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM........ 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Tank Volume vs. Concentration Requirements.. ~ ~ ~ ~ ~ ~ ~, ~ ~ ~ ~ ~ ~ ~ 3/4 1"22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..... ~....... .. ~ 3/4 2-1 Figure 3.2. 1-1 A
Maximum Average g P1 "EP Fuel Type BPBCRB219.........
E~'~~p Planar Linear Heat Generation Rate 3/4 2-2 Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate A g P1 Ep 1 '&W '~H~p Fuel Type PBCRB176........................................ 3/4 2-3 Figure 3.2. 1-3 Maximum Average Planar Linear Heat Generation Rate A
Fuel Type g P1 E P PE>>'1 P.6 PBCRB071.......................
~~PA 3/4 2"4 3/4 2.2 AVERAGE POWER RANGE MONITOR SETPOINTS..................... 3/4 2-5 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)....... 3/4 2-7 Figure 3.2.3-1 Minimum Critical Power Ratio vs. x at Rated Flow... 3/4 2-9 Figure 3.2.3-2 Kf as a Function of Percent Core Flow...... 3/4 2-10 3/4.2.4 LINEAR HEAT GENERATION RATE.............;....:... 3/4 2"11 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................. 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation... 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times.... 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements.................. 'I 3/4 3-7 3/4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION..'. 3/4 3-10 Table 3.3.2-1 Isolation Actuation Instrumentation . 3/4 3-12 NINE MILE POINT - UNIT 2 v JUN 25 1986
INDEX 143 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS PAGE 3/4.7 PLANT SYSTEM 3/4.7. 1 PLANT SERVICE WATER SYSTEMS Plant Service Water System - Operating.................... 3/4 7-1 Plant Service Water System - Shutdown...................... 3/4 7-4 3/4.7.2 Revetment-Ditch Structure...... 3/4 7-7 Table 3.7.2-1 Survey Points for Revetment-Ditch Structure.......... 3/4 7-9 3/4.7.3 CONTROL ROOM OUTDOOR AIR SPECIAL FILTER TRAIN SYSTEM...... 3/4 7-11 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM..................... 3/4 7-14 3/4.7.5 SNUBBERS....... 3/4 7-16 Figure 4.7.5-1 Sample Plan for Snubber Functional Test............: 3/4 7-21 3/4.7.6 SEALED SOURCE CONTAMINATION . 3/4 7-22 3/4.7. 7 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System.................... 3/4 7-24 Spray and/or Sprinkler Systems.. 3/4 7-27 lf C Oq Systems............................................... 3/4 7-29 Halon Systems.... 3/4 7-3/<
F sre II Hose Stations........................................
I J. J.
3/4 7-32 Table 3.7.7.5-1 Fire Hose Stations................................. 3/4 7-34 Yard Fire Hydrants and Hydrant Hose Houses................ 3/4 7-36 Table 3.7.7.6-1 Yard Fire Hydrants and Associated Hydrant Hose H ouseso ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-37 3/4.7.8 FIRE RATED ASSEMBLIES . 3/4 7-38 SYSTEM................................
'/4 3.4.7. 9 MAIN TURBINE BYPASS 7+X
+o NINE MILE POINT UNIT 2 X11 Jun Ss )ggg
INDEX 144 BASES FOR SECTIONS 3.0/4.0 PAGE REACTOR COOLANT SYSTEM (Continued) 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................... B3/4 4-3 Operational Leakage....,................. B3/4 4-3 3/4 ~ 4e 4 CHEMISTRYo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B3/4 4"4 3/4.4. 5 SPECIFIC ACTIVITY............... B3/4 4"4 3/4.4. 6 PRESSURE/TEMPERATURE LIMITS................. B3/4 4"5 .
Bases Table B 3/4.4.6-1 Limiting Reactor Vessel Toughness.......... B3/4 4-6 Bases Figure B3/4. 4. 6-1 Fast Neutron Fluence (E>l MeV) at > T as a Function of Service Life at 90K of RATED'HERMAL POWER and 90K Availability........... B3/4 4-7 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES.............. B3/4 4-8 3/4.4.8 STRUCTURAL INTEGRITY.... B3/4 4-8 3/4.4.9 RESIDUAL HEAT REMOVAL..... B3/4 4-8 3/4.5 EMERGENCY CORE COOLING SYSTEMS ECCS - OPERATING AND SHUTDOWN B3/4 5-1 3/4 5 2 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
3/4.5.3 SUPPRESSION POOL.............;........... B3/4 5-3 3/4.6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIMARY CONTAINMENT Primary Containment Integrity........... B3/4 6-1 Primary Containment Leakage.... B3/4 6-1 Primary Containment Air Locks..................... B3/4 6-1 Primary Containment Structural Integrity.................. B3/4 6-2 Drywell and Su'ppression Chamber Internal Pressure......... B3/4 6-2 Drywell Average Air Temperature B3/4 6-2 NINE MILE POINT " UNIT 2 XV111 JUN 25 1986
145 INDEX BASES PAGE CONTAINMENT SYSTEMS (Continued)
Primary Containment Purge System.......................... B3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...... .. .. ................. B3/4 6-3 3/4. 6. 3 PRIMARY CONTAINMENT ISOLATION VALVES...................... B3/4 6-4 3/4.6.4 SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS............. B3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT..................................... B3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL............... B3/4 6-6 3/4.7 PLANT SYSTEMS 3/4.7. 1 PLANT SERVICE WATER SYSTEMS..................... B3/4 7-1 3/4.7.2 REVETMENT-DITCH STRUCTURE................ B3/4 7-1 3/4.7. 3 CONTROL ROOM OUTDOOR AIR SPECIAL FILTER TRAIN SYSTEM...... B3/4 7-1 Bases Figure B3/4.7.2-1 Plan View - Revetment-Ditch Structure, Inservice Inspection Service Station Locations..... ... ~ B3/4 7-2 Bases Figure B3/4.7.2-2 Typical Section - Revetment-Ditch Structure, Inservice Inspection Service Station L ocatlons................oo....oo.....o..o...o..o.o....... B3/4 7-3 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM................. ~... B3/4 7-4 3/4.7.5 SNUBBERS....................................... B3/4 7-4 3/4. 7. 6 SEALED SOURCE CONTAMINATION............................... B3/4 7" 6 3/4. 7.7 FIRE SUPPRESSION SYSTEMS .............................. B3/4 7-6 3/4.7.8 FIRE RATED ASSEMBLIES..................................... B3/4 7-7 3/4.7.9 MAIN TURBINE BYPASS SYSTEM..................... B3/4 7-7 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 3/4.8.2 AC SOURCES, DC SOURCES, AND ONSITE POWER 3/4.8.3 DISTRIBUTION SYSTEMS...................................... B3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES.......... B3/4 8-3 NINE MILE POINT - UNIT 2 X1X JUN 25 >s86
146 INDEX BASES FOR SECTIONS 3.0/4.0 PAGE 3/4.9 REFUELING OPERATIONS 3/4.9. 1 REACTOR MODE SWITCH........................... 83/4 9-1 3/4. 9. 2 INSTRUMENTATION......... 83/4 9-1 3/4. 9. 3 CONTROL ROD POSITION...................................... 83/4 9-1 3/4.9.4 DECAY TIME................................................ 83/4 9-1 3/4.9.5 COMMUNICATIONS............................................ 83/4 9-1 3/F 9.6 REFUELING PLATFORM....,................. 83/4 9-1 3/4.9.7, CRANE TRAVEL - SPENT FUEL STORAGE POOL.... 83/4 9-2 3/4.9.8+ WATER LEVEL, REACTOR VESSEL AND WATER LEVEL, AND SPENT 3/4.9.9> FUEL STORAGE POOL 83/4 9-2 3/4.9. 10 CONTROL ROD REMOVAL....................................... 83/4 9-2.
3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............. 83/4 9"2 3/4. 10 SPECIAL TEST EXCEPTIONS 3/4. 10. 1 PRIMARY CONTAINMENT INTEGRITY..................'........... 83/4 10-1 3/4. 10.2 ROD SEQUENCE CONTROL SYSTEM............................... 83/4 10-1 3/4. 10. 3 SHUTDOWN MARGIN DEMONSTRATIONS............................ 83/4 10-1 3/4. 10. 4 RECIRCULATION LOOPS 83/4 10-1 3/4.10.5 OXYGEN CONCENTRATION..... 83/4 10" 1 3/4. 10. 6 TRAINING STARTUPS.. 83/4 10-1 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING............ 83/4 10"1 3/4. 11 RADIOACTIVE EFFLUENTS 3/4. 11. 1 LIQUID EFFLUENTS C oncentration.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 83/4 11-1 D ose.......... ~ ~ ~ ~ ~ ~ ~ 83/4 11-1 Liquid Radwaste Treatment System.. 83/4 11-2 Liquid Holdup Tanks......... ........
~ 83/4 11-2 NINE MILE POINT - UNIT 2 XX
147 INDEX BASES FOR SECTIONS 3.0/4.0 PAGE RADIOACTIVE EFFLUENTS (Continued) 3/4. 11.2 GASEOUS EFFLUENTS 0 ose Rate ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B3/4 11-2 0 ose - Noble Gases...................................... B3/4 11-3 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form................ B3/4 11-4 Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System. 83/4 11-5 Explosive Gas Mixture... B3/4 11-5 Main Condenser - Offgas................................. B3/4 VENTING or PURGING..................................;... 11'3/4 11-6 3/4, 11. 3 SOLID RADIOACTIVE WASTES. B3/4 11-6 3/4. 11. 4 TOTAL DOSE........................ B3/4 11-6 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12. 1 MONITORING PROGRAM.. 83/4 12-1 3/4.12.2 LAND USE CENSUS................... B3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM.............i........ B3/4 12-2

5. 0 DESIGN FEATURES 5 1

~ SITE ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ V J E xcluslon Area..................................... 5 1 Low Population Zone............................................ 5-1 Map Defining'Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
NINE MILE POINT - UNIT 2 xxi JUN 25 1986
148 INDEX ADMINSTRATIVE CONTROLS PAGE 6, 2 ORGANIZATION Offsite.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 1 U nit Staff.......,...,..............................,
Independent Safety Engineering Group F unction............................................. 6-3 C omposltlon. ~ ~ .... ....
~ ~ ~ ~ ~ ... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Responsibilities.................... 6-3 Figure 6.2.1-1 ~
Organization Chart........
'R~r~~hwmk
.i%at+en anagement M
6-4 Figure 6.2.2-1 Nine Mile Point Nuclear-S4a&~R - Site+peratlons-0 rganlzatlon............................. 6-5 Table 6.2.2-1 Minimum Shift. Crew Composition ...................... 6-6 R ecords ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
Assistant Station Shift Supervisor/Shift Technical Advisor..... 6-7

6. 3 FACILITY STAFF UALIFICATIONS......... ~........................ 6-7 6.4 TRAINING.... ~ ........... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

6. 5 REVIEW AND AUDIT Site Operations Review Committee F unction................................................ 6-8 Composition............ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6 8 A lternates.............................................. 6-8 M eetlng Frequency....................................... 6-8 Q uorum...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 Responsibilities........................... 6-8 0 utleS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

NINE MILE POINT - UNIT 2 XX111 JUN 2 5 tggj
149 3/4.3 INSTRUMENTATION 3/4.3. 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION reactor protection system instrumentation channels

3. 3. 1 shown As a minimum, in Table 3. 3. l-ltheshall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3. 1-2.

APPLICABILITY As shown in Table 3.3.1-1.
ACTION:

a. With the number of'PERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one Trip System, place the inoperable channel(s) and/or that Trip System in the tripped condition*

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The provisions of Specification 3.0.4 are not applicable.

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both Trip Systems, place at least one Trip System** in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3. 1-1.

SURVEILLANCE RE UIREMENTS 4.3. l. 1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
.4.3. 1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3. 1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3. 1-2 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per Trip System so that all channels are tested at least'once per N times 18 months, where N is the total number of redundant channels in a specific reactor Trip System,
" An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be r'estored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3. 1-1 for that Trip Function shall be taken.
"* The Trip System need not be placed in the tripped condition if this would cause the Trip Function to occur. When a Trip System can be placed in the tripped condition without causing the Trip Function to occur, place the Trip System with the most inoperable channels in the tripped condition.gf both systems have the same number of inoperable channels, place either Trip System in the tripped condition.
NINE MILE POINT " UNIT 2 3/4 3-1
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150 INSTRUMENTATION 3/4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION

3. 3. 2 The isolation actuation instrumentation channels shown in Table 3. 3. 2-1 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2-2 and with ISOLATION SYSTEM RESPONSE TIME shown in Table 3.3.2-3.

APPLICABILITY: As shown in Table 3.3.2-1.
ACTION:

a. With an isolation actuation instrumentation channel Trip Setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.2-2, declare the channel inoperable until the channel is re-stored to OPERABLE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one Trip System, place the inoperable channel(s) and/or that Trip System in the tripped condi-tion within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The provisions of Specification 3.0.4 are not applicable.

C. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both Trip Systems, place at least one Trip System"" in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and take the ACTION required by Table 3.3.2-1.
" An inoperable channel need not be placed in. the tripped condition if this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
"" The Trip System need not be placed in the tripped condition if this would cause the Trip Function to occur. When a Trip System can be placed in the tripped condition without causing the Trip Function to occur, place the Trip System with the most inoperable channels in the tripped condition, $ f both systems have the same number of inoperable channels, place either Trip System in the tripped condition.
NINE MILE POINT -, UNIT 2 3/4 3-10
(
JL
(
151
'ABLE 3. 3. 3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 33- With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .
~Not, Ms'.
ACTION 34 Vi-ter-thenumber-ofOPERABLE-charm%~-ess than-reqtm e&b~4e
~lelAEEE
-bus-power avaÃabA+ty-at~as-t once-per-12-hours-or-de~e tthhe-
~oc~ed-ECCS iwoperatAe ACTION 35- With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the associated AOS valve or ECCS inoperable.
ACTION 36- With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a. For one Trip System, place that Trip System in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months " or declare the HPCS system in-operable.

b. For both Trip Systems, dec'1are the HPCS system inoperable.

ACTION 37- With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within 1 hour" or declare the MPCS system inoperable.
ACTION 38- With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8. 1, 1 or 3.8. 1.2, as appropriate.
ACTION 39- With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months "; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST.
The provisions of Specification 3.0.4 are not applicable.
NINE MILE POINT - UNIT 2 3/4 3-34
1 ~
152 INSTRUMENTATION RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATMS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS 4.3.4. l. 1. Each ATMS-RPT System instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4. 1-1.
IG 4.3.4. 1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automated operation of all channels shall be performed at least once per 18 months.
NINE MILE POINT - UNIT 2 3/4 3-46
TABLE 3.6.3-1 (Continued)
PRIHARY CONTAINMENT ISOLATION VALVES ISOLATION VALVE ISOLATION MAXIMUM CLOSING VALVE NO. VALVE FUNCTION GROUP SIGNAL(a) TIME (SECONDS) 2IAS*SOV164 ADS Hdr A Nz supply Outside IV B,F,Z,RH 2IASASOV165 ADS Hdr B Nq supply Outside IV B,F,Z,RH 2IAS*SOV166 IAS DryweA Relief Valve Outside IV B,F,Z,RH 2IASASOV184 IAS Srywe-I+ Relief Valve Inside IV B,F,Z,RH 2 I ASASOV168 Inst. Air to Testable Check Outside IV B,F,Z,RH 2IASASOV180 Inst. Air to Testable Check Inside IV B,F,Z,RH 2IAS"SOV167 IAS to Test Ck. 8 Vac. Bkrs. Outside IV B,F,Z,RH 2IASASOV185 IAS to Test Ck. 8 Vac. Bkrs. Inside IV B,F,Z,RH 2HCS"HOVl A,B Hz Recombiners Sply to Supp. Chamber Outside IV's B,F,Z,RH 30 2HCS"HOV2 A,B Hz Recomb. Ret. from Supp. Chamber Outside IV's B,F,Z,RH 30 2HCS*HOV3 A,B Hz Recomb. Return from Drywell Outside IV's 8 B,F,Z,RH 30 2HCS"HOV4 A,B(n) Hz Recomb. Suply. to Supp. Chamber Inside IV's B,F,Z,RH 30 2HCS"HOV5 A,B(n) Hz Recomb. Ret. from Supp. Chamber Inside IV's B,F,Z,RH 30 2HCS"HOV6 A,B(n) Hz Recomb. Ret. from Drywell Inside IV's 30 2CPSASOV119 Containment Purge to Supp. Chamber Outside IV B,F,Y,Z,RH 2CPSASOV120 Containment Purge to Drywell Outside IV B,F,Y,Z,RH 2C PS" SOV121(n) Containment Purge to Supp. Chamber Inside IV B,F,Y,Z,RH 2CPS*SOV122(n) Containment Purge to Drywell Inside IV B,F,Y,Z,RH 2CMS*SOV24 A,B,C,D CHS from Drywell Inside 8 Outside IV's B,F,Z,RH 2CMS*SOV26 A,B,C,D CHS from SP Inside 8 Outside IV's B,F,Z,RH 2CHS"SOV32 A,B CHS to Drywell Outside IV's B,F,Z,RM 2CMS"SOV33 A,B(n) CHS to Drywell Inside IV's B,F,Z,RH 2CMS*SOV34 A,B(n) CMS to SP Inside IV's B,F,Z,RM 2CMS"SOV35 A,B CHS to SP Outside IV's B,F,Z,RH 2CMS"SOV60 A,B CMS to Drywell Outside IV's B,F,Z,RH 2CMS"SOV61 A,B(n) CHS to Drywell Inside IV's B,F,Z,RH 2CMS"SOV62 A,B CMS to Drywell Outside IV's B,F,Z,RM 2 CMS"SOV63 A,B(n) CMS to Drywell Inside IV's B,F,Z,RM
TABLE 3. (Continued)
PRIMARY CONTAINMENT ISOLATION VALVES ISOLATION VALVE ISOLATION MAXIMUM CLOSING VALVE NO. VALVE FUNCTION GROUP SIGNAL(a) TIME (SECONDS) 2ISC" EFV34 To 2ISC" FT478 2ISC"EFV35 To 2ISC"FT470 2ISC"EFV36 T 2ISC"FT47F 2ISC" EFV37 2ISC*EFV38
~To 2ISC*FT47S To 2ISC*FT47H 2ISC" EFV39 To 2ISC"FT47P 2ISC" EFV40 To 2ISC"FT488 2I SC" EF V41 ..., To 2ISC*FT47U 2ISC" EFV42 To 2ISC"FT47W,FT480 2ISC" EFV9 Containment Pressure 2ISC"PT15C, 168, 16D 21SCAEF V12 Containment Pressure 2ISC*PT158, 178, 17D 2 ISC*EF V16 Containment Pressure 2ISC"PT15A, 16A, 16C 2I SC" EF V19 Containment Pressure 2ISC" PT15D, 17A, 17C 2CHS" EFVlA To CHS"PTlA 2CHS" EFV18 To CHS"PT18 2CHS"EFV3A To CMS"PT2A 2CHS*EFV38 To CMS"PT28 2CHS*EFV5A ~
To CHS"PT7A 2CHS*EFV58 To CHS*PT78 2CHS*EFV6 To CHS-PT168 2CHS"EFVSA To CHS"LT9A, llA, 114 2Cl1S"EFVSB To CHS"LT98, 118, 105 2CHS"EFV9A To CHS"LT9A, llA, 114 2CHS"EFV98 To CHS*LT98, 118, 105 2CHS"EFV10 To CMS-PI173 CW 2ICS" EFVl 2ICSAEFV2 2DERAEFV31 To 2ICS*PDT167 To 2ICS*PDT167 To DER"PT134 gl
pl I
0
155 INSTRUMENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATMS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of study events in General
.Electric Company Topical Report NED0-10349, dated March 1971, NED0-24222, dated December 1979; and Section 15.8 of the FSAR.
The end-of-cycle recirculation pump trip (EOC-RPT) system is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the,end of cycle. The physical phenomenon involved is that the void reactivity feedback from a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each EOC-RPT system trips both recirculation pumps., reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events for which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system;. a position switch from each of the other two stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast
.closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.
t Each EOC-RPT system may be manually bypassed by use of a switch which is administratively controlled by procedures. The manual bypasses and the auto-matic Operating Bypass at less than 30K of RATED THERMAL POMER are annunciated in the control room.
The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,
190 milliseconds. Included in this time are: the time from initial valve movement to reaching the Trip Setpoint, the response time of the sensor, the response time of the system logic, and the time alloted for breaker arc supression.
~p Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference NINE MILE POINT " UNIT 2, B3/4 3"3
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1 l~
156 INSTRUMENTATION

BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION 3/4. 3.4 (Continued) between each Trip Setpoint and the Allowable Value is an allowance for instru-,

ment drift specifically allocated for 'each trip in the safety analyses. The Trip Setpoint and Allowable Value also contain additional margin for instrument accuracy and calibration capability.
3/4.3.5 REACTOR CORE ISO ATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel.
Operation with a trip set less conservative than its Trip Setpoint but within
-. its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is an allowance for instru-ment drift specifically allocated for each trip in the safety analyses. The Trip Setpoint and Allowable Value also contain additional margin for instrument accuracy and calibration capability.
3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4. 1.4, Control Rod Program Controls, and Section 3/4.2, Power Distribution Limits. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the. difference between each Trip Setpoint and the Allowable Value is an allowance for instru-ment drift specifically allocated for each trip in the safety analyses. The Trip Setpoint and Allowable Value also contain additional margin for instrument accuracy and calibration capability. The setpoint is referenced to a scram discharge volume instrument zero level at elevation 263 feet 10 inches.
3/4. 3. 7 MONITORING INSTRUMENTATION 5crom bschaqa votre. eo4r lava,l -hi~)
3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels; (2) the alarm or automatic action is initiated when the radiation level Trip Setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to moni'tor and assess these variables following an accident. This capability is consistent with 10 CFR 50, Appen-dix A, General Design Criteria (GDC) 19, 41, 60, 61, 63, and 64.
NINE MILE POINT - UNIT 2 B3/4 3-4
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'j~s v it LI 4
0
157 REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads from temperature and pressure changes in the system.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 are derived from the fracture toughness requirements of 10 CFR 50, Appendix G, and ASME Code Section III, Appendix G. The curves are based on the RTNDT and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in FSAR Subsection 5.3. 1.5, "Fracture Toughness."
The reactor vessel materials have been tested to determine their initial RTNDT. The results of these tests are shown in Bases Table B3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence, phosphorus content, and copper content of the material in question, can be predicted using Bases Figure B3/4.4.6-1 and the recommenda-tions of RG 1.99, Revision 1, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."
-The actual shrift in RTNDT of the vessel material will be established periodi-cally during operation by removing and evaluating irradiated specimens in-stalled near the inside wall of the reactor vessel in the. core area~Since the neutron spectra at the specimens and vessel inside radius are essentially identical@+"(he irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figures 3.4.6. 1-1, 3.4.6. 1-2, and 3.4.6.1-3 shall be adjusted, as required, on the basis of the specimen data and recommendations of RG 1.99, Revision 1. Data determined from specimens removed at the end of the first cycle will be used to adjust the fluence of Bases Figure B3/4.4.6-1.
The pressure-temperature limit lines shown in Figures 3.4.6.1-1 and 3.4.6. 1-3, for inservice hydrostatic testing and leak testing and for reactor criticality have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.
The number of reactor vessel irradiation surveillance capsules and the frequen-I cies for removing and testing the specimens in these capsules are provided in Table 4.4.6. 1.3-1 to assure compliance with the requirements of Appendix H to 10 CFR 50.
NINE MILE POINT - UNIT 2 B3/4 4-5
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