ML19262C710
| ML19262C710 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 02/04/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19262C708 | List: |
| References | |
| NUDOCS 8002150471 | |
| Download: ML19262C710 (13) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION y
g 3
- < E WASHINGTON. D. C. 20556 4
,o BOSTON EDISON COMPANY DOCKET N0. 50-293 PILGRIM NUCLEAR POWER STATION UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 40 License No. DPR-35 1.
The Nuclear Regulatory Commission (the Comission).has found that:
A.
The portions of the applicaticns for amendment by Boston Edison Company (the licensee) dated July 6 and September 27, 1979, for which an amendment is hereby granted, comply with the standards and requirements of the Atomic Energy Act of 1954, asmended (the Act), and the Commission's rules and regulationT set forth in 10 CFR Chapter I; B.
The facility will coerate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authoriled by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this anendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
1 L
80021I'0
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2-2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility License No. OPR-35 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 40, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR _ REGULATORY COMMISSION
..f) Yg
. (, d j Thomas AY ppolito, Gief Operating Reactors Branch #3,.
Division of Operating React 6rs~
Attachment:
Changes to the Technical Speci#ications Date of Issuance:
February 4, 1980
ATTACHMENT TO LICENSE AMENDMENT N0. 40 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Revise Appendix A as follows:
Remove Inse rt 32 32 49 49 87 87 137c 137c 137d 137d 137e 137e 160 160 208 208 208a 210 210 212 212 e:~
~
TABI.E 4.1.2 NEACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT cal.I!1 RATION HINIMMM cal.IllRATION FREQUENCIES FOR REACTOR PROTECTION INSTittlHENT CilANNELS Instrument Channel Group '1)
Calibration Test (5)
Minimum Frequency (2)
IRH liigh Flux C
Comparison to APRM on Contr>lled Note (4)
Shutdowns Full Calibration oncehperating cycle APRM 11igh Fc ux 1
Output Signal B
lleat Balance Once evnry 3 Days Flow Bias Signal B
Internal Power and Flow Test Each P_I;ueling Outage LPRH Signal B
TIP System Traverse Every 1000 Effective Full Power llours liigh Reactor Prer.sure A
Standard Pressure Source Every 3 Honths liigh Drywell Pressure A
Standard Pressure Source Every 3 Hontha Reactor Low Water I.evel A
Pressure Standard Every 3 Honths liigh Water Level in Scram Discharge Volume A
Note (6)
Note (6)
Turbine Condenser I.ow Vacuum A
Standard Vacuum Source Every 3 Honths Main Steam Line Isolation Valve Closure A
Note (6)
Note (6)
Hain Steam Line High Radiation B
Standard Current Source (3)
Every 3 Honths Turbine First Stage Pressure Permissive A
Standard Pressure Source Every 6 Honths Turbine Control Valve Fast Closure A
Standard Pressure Source Every 3 Honths Turbine Stop Valve Closure A
Note (6)
Note (6)
Reactor Pressure Permissive A
Standard Pressure Source Every 6 Honths r%
fills TABLE 3.2.B (Cont'd)
INSTElHEriTATIOff Tl!AT INITIATES OR C0!TfEOLS D!E CORE AIID COTITAINMENT COOLI!1G SYSTDG Minimum // of Operable Instrument Cliannels Ier Trip System (1)
Trip Funct. ion Trip 1.cvel "M t ing Pemarka 1
Core Spray Iump Start In conjunction with loss of Timer o <t el sec.
power initiates sequential 1
LPCI Pump Start Timer 4 et <6 :.ec.
starting of CSCS pumps 1
LPCI Pump Start Tirer 9 <-t e11 sec.
1 Auto Blowdown Timer
>90,
<120 sec.
In conjunction with low Low Feactor Water Level, Ilich Dry-well Pressure and LPCI or Core Spray Ibmp runnin,; interlock, initiates Auto Blowdown.
2 RHR (LPCI) Ibrep 1)ischarge l'io + 10 r::ig Defers ADS actuation pending con-Itesrure Interlock firmation of low Ivessure core cooling system operation. (LPCI 2
Core Spray Pump Discharge 150 2 10 psig or Core Spray Pump running inter-l'ressure Interlock lock.
20-2' L of rated 1.
Permits closure of the Diesel 2
Ebergency Bus Voltage f
f Relay
) voltage resets Generator to an unloaded at less than 50%
emergency bus.
2.
Fermits startirs: of TS I4 kV motors.
3.3 and 4.3 BASES:
A.
Reactivity Limitation 1.
The core reactivity limitation is a restriction to be applied principally to the design of new fuel which may be loaded in the core or into a particular refueling pattern.
Satisfaction of the limitation can only be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle. The generalized form is that the reactivity of the core loading will be limited so the core can be made subcritical by at least R + 0.25% a k at the time of the test, with the strongest control rod fully withdrawn and all others fully inserted.
The value of R in %4k is the amount by which the core reactivity, at any time in the operating cycle, is calculated to be greater than at the time of the check; i.e.,
the initial loading.
R must be a positive quantity or zero.
A core which contains te=porary control or other burn-able neutron absorbers may have a reactivity characteristic which increases with core lif e-time, goes through a maximum and then decreases thereafter.
'"~
The value of R is the dif ference between the cal-culated core reactivity at the beginning of the operating cycle and the calculated value of core reactivity any time later in the cycle where it would be greater than at the beginning.
The value of R shall include the potential shutdown =argin loss assuming full B C settling in all inverted poison 4
tubes present in the core.
A new value of R cust be determined for each full cycle.
The 0.25%Ak in the expression R + 0.25%sk is provided as a finite, demonstrable, suberiticality margin.
This margin is demonstrated by full with-drawal of the strongest rod and partial withdrawal of an adjacent rod to a position calculated to in-sert at least R + 0.25 ak in reactivity, or by an ins equence, xenon-free cold critical measurement to demonstrate at least R & 0.25% Ak in reactivity with the most reactive control rod fully withdrawn.
Observation of suberiticality in this condition assures suberiticality with not only the strongest rod fully withdrawn but at lea s t an R + 0.25% Ak margin beyond this.
I 87 i
Table 3.6.I SAFFTY RELATED S110CK SUPPRESSORS (SNUBBERS)
Snubber No, location Elevation Snubber in liigh Snubbers Snubbers Snubbers ihadiation Area Espeelully inaccesutble Acceselble During Shutdown Difficult to During Normal During Normal Remove Operation Operation 11 SS-1-10-1 Main Steam 1.ine 42' X (Drywell)
SS-1-10-2 Main Steam Line 42' X (Drywell)
SS-1-10-3 Main Steam Line 42' X (Drywell)
S S-1-10-4 Main Steam Line 42' X (Drywell)
SS-1-10-5 Main Steam Line 42' X (Drywell)
SS-1-10-6 Main Steam Line 42' X (Drywell)
SS-1-10-7 Main Steam Line 42' X (Drywell)
SS-1-10 8 Ma in Steam Line 42' X (Drywell)
SS-1-10-9 Main Steam Line 42' X (Drywell)
SS-1-10-10 Main Steam 1.tne 42' X (Drywell)
SS-1-10-11 Malu Steam 1.tne 42' X (Drywell)
SS-1-10-12 Main Stenm 1.tne 42' X (Drywell)
SS-6-10-6 Feetw. iter Sy8.
41' X (Drywel1)
SS-6-10-7 Fecilwat er Syn.
41' X (Drywell)
SS-6-10-8 Feed ater Sys.
44' X (Drywell)
SS-6-10-9 Feet. water Syn.
41' X (Drywell) i SS-6-10-10 Feedwater Syu.
44' X (Drywell)
I SS-10-30-1 RilR Syntem 52' X (Drywell)
SS-10-20-2 lulR Syutem 52' X (Drywell)
SS-10-20-3 lulu Syntem 52' X (Drywell)
SS-10-20-4
'UIR System 52' X (Drywell)
SS-10-30-5 RilR System 24' X (Drywell)
SS-10-30-6 RilR System 24' X (Drywell)
SS-10-20-7 RilR System 24' I,'
X (Drywell)
SS-10-20-8 RilR System 24' X (Drywell)
SS-10-3-9 RitR System 87' X (Drywell)
SS-10-3-10 RilR System 90' X (Drywell)
SS-2-20-1 Recir. System 42' X
X (Drywell)
SS-2-20-2 Recir. System 42' X
X (Drywell) 42' X
X (Drywell)
SS-2-20-3 Recir. System SS-2-20-4 Recir. System 42' X
X (Drywell)
'; S 10 - 5 Ite r I. Syntem IS' X
X (Drywel1)
SS-2-30-6 Recir. System IS' X
X (Drywell) 3S-2-10-7 It e c i r. System 15' X
X (Drywell)
C
Table 3.6.1 SAFETY RELATED S110CK SUPPRESSORS (SNUBBERS)
Snubber No.
Locatir,n Elevation Snubber in liigh Snubbers Snubbers Snubbers Radiation Area Especially Inaccessible Accessible During Shutdown Difficult to During Normal During Normal Remove Operation Operation SS-2-30-8 Recir. System 15' X
X (Drywell)
SS-2-30-9 Recir. System 11' X
X (Drywell)
SS-2-30-10 Recir. System 11' X
X (Drywell)
SS-2-10-Il Herir. Synsem 27' X
X (Drywoll)
SS-2-10-12 Itecir. Syntem 27' X
X (Drywell)
S S 10 - 1 1 Ite r l i. Syntem 27' X
X (Drywell)
SS-2-30-14 Recir. Syutem 27' X
X (11rywell)
SS-2-30-15 l<ecir. System 27' X
X (Drywell)
X (Drywell)
SS-2-30-16 Recir. System 27' X
X (ltrywell)
SS-2-20-19 Recir. System 16' X
i SS-2-20-20 Herir. Syntem 16' X
)
X (Drywell)
SS-2-20-21 Ite c i r. Syntem 19' X
i X (ilrywell)
SS-2-20-22 Iteci r. System 16' X
X (Drywell)
SS-2-50-21 Ractr. RF#t em 17' N
X (DryWell)
SS-2-20-24 Recir. System 18' X
X (Drywell)
SS-2-20-25 Recir. System 16' X
X (Drywell)
SS-2-50-26 Recir. System 16' X
X (Drywell) 0 d
c e
Table 3.6.1 SAFETY RELATED S110CK SUPPRESSORS (SNUBBERS) l Snubber No.
In cat ion Elevation Snubber in liigh Snubbers Snubbers Sr.ubbers Radiation Area Especially Inaccessible Accessible During Shutdown Difficult to During Normal During Normal Remove Operation Operation SS-6-10-1 Feedwater System 42' X (Drywell)
SS-6-10-2 Feedwater System 42' X (Drywell)
SS-6-10-3 Feedwater System 42' X (Drywell)
SS-6-10-4 Feedwater System 42' X (Drywell)
SS-6-10-5 Feedwater System 42' X (Drywell)
SS-13-3-1 RCIC 38' X (Drywell)
SS-13-3-2 RCIC 38' X (Drywell)
SS-14-3-1 Core Spray 65' X (Drywell)
SS-14-3-2 Core Spray 65' X (Drywell)
SS-14-3-3 Core Spray 65' X (Drywell)
SS-14-3-4 Core Spray 65' X (Drywell)
SS-23-10-1 II.P. C. L.
42' X (Drywell)
SS-23-10-2 li. P. C. I.
42' l
X (Drywell)
X ll.P.C.I. Quadrant S-23-3-30 ll. P. C. L.
-3'09" X li.P.C.I. Quadrant S-23-3-31 II.P. C. I.
-3'09" X li.P.C.I. Quadrant S-23-10-32 II.P.C.I.
-3'09" X ll.P.C.I. Quadrant S-23-3-33 li. P. C. I.
-3' 09" X ll.P.C.I. Quadrant S-23-10-34 li. P. C. I.
-6' X li.P.C.I. Quadrant S-23-10-35 Il P.C.I.
-6' p
X li.P.C.I. Quadrant S-23-3-36 li. P. C. I.
-3'09" j
X ll.P.C.I. Quadrant S-23-3-37 II. P. C. I.
-3'09" X RilR Pump Room S-10-3-43 RilR
-3'06" X RilR Pump Room S-10-20-44 RilR
-3'06" X Reactor Building S-30-3-45 RBCCW 83'5" X Torus Compartment S-10-10-46 RilR 6"
d Kalifications to this Table due to changes in high radiation areno should be submitted to the NRC as part of t he next l icense amendment.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREME?TTS 3.7.D Primary Containment Isolation 4.7.D Primary Containment Isolation Valves Valves 1.
During reactor power operating 1.
The primary containment conditions, all isolation isolation valves surveillance valves listed in Table 3.7.1 shall be performed as follows:
and all instrument line flow check valves shall be operable a.
At least once per operating except as specified in 3.7.D.2.
cycle the operable isolation valves that are power operated and automatically initiated shall be tested for simulated automatic initiation and closure times.
b.
At least once per quarter:
(1) All normally open power operated isolation valves (except for the main steam line power-operated
~~~
isolation valves) shall be fully closed and reopened.
(2) Trip main steam isolation val.yes individually and verify closure time, c.
At least twice per week the main steam-line power-operated isolation valves shall be exercised by partial closure and subsequent reopening.
d.
At least once per operating cycle the operability of the reactor coolant system instrument line flow check valves shall be verified.
2.
In the event any isolation 2.
'.menever an isolation valve valve specified in Table 3.7.]
listed in table 3.7.1 is becomes inoperable, reactor inoperable, the position of at power operation nnr continte least or.e other valve in each provided at least one valve in line having an inoperable each line having an inoperable ialve shall be recorded dai13 val re is in tha mode corres-por. ding to the isolated condition.
3.
If Specification 3.7.D.1 and 3.7.D.2 cannot be net, an orderly shutdown shall be jnitiated and the reactor shall be in the Cold Shut-down condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
160
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY The Pilgrim Station Manager shall bz responsible for overall f aciliry In his absence the Assistant Station Manager or an alte nate operation.
desianated in writing by the Statirn Manager shall assume this responsibility.
6.2 ORGANIZATION A.
OFFSITE and technical The Company organization for statica management support shall be as shown on ?igure 6.2.1.
B.
FACILITY The Facility organization shall be as shown on Figure 6.2.2 and:
1.
Each on duty shift shall be ce= posed of at least the =ini=un shift crew composition shown in Table 6.2.1.
2.
At least one licensed Operatet hall be in the control reoc when fuel is in the reactor.
3.
At least two licensed Operators shall be present in the con:rol room during reactor startup, scheduled reactor shutdow: and during recovery f rom reacter trtps.
4.
An individual qualified in radiation protection proced:res shall be on site when fuel is in the reactor.
5.
ALL CORE ALTERATIONS perfo =ed while fuel is in the reactor vessel af ter the initial fuel loading shall be directly supervised by either a licensed Senior Reacter Operator or Senior Reactor 0;erater Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
6.
The organization of the ad=inistration of the Pilgrim : Fire Prevention and Protectics Program shall be as shown on Figure 6.2.3.
A Fire Brigade ef 5 members including the Fire Chief shall be maintained on site at all times. This excludes 3 menbers of the mini =um shift crew necessary for saf e shutdown and any persennel required for other essential f unctions during a fire energency.
6.3 FACILIIT STAFF OL'ALIFICATIONS The qualifications with regard to educational and experience backgrou:ds of the f acility staf f at the time cf appoint =ent to the active pcsitica shall meet the requirements as described in the American National Staniards Institute N18.1-1971, " Selection and Iraining of Personnel for N;: lear Pcwer Plants."
In addition, the individ.;a; perfor=ing the function of Radiation Protection Manager shall zeet er exceed the qualifications of Rer:lat:ry Guide 1.8, September, 1975.
205
STATION MAN AGl~.R (ORC)
(NSRAC)
- STAFF ASSISTANT STATION MANACER (ORC) 1 CHIFF CHIEF Clll EF ClllEF HET110DS,
OPERATING HAINTtNANCE RADIOloCICAL TErtlNICAL SECUltiTY TRAINING &
EIMINEFR ENGINtER INGlhEER ENGINEFR SUPER'/ISOR COMPLIANCE (ORC) (SRO)
(OHC)
(ORC)
(ORC)
CROUP LEADER (ORC) 1 WAtrH l t's llra I t' Al.
11 t'llN 14:Al.
tNGlkttk(S)
STAFF STAFF (SRO) l I
OPERATlW' lit Al.lil WASIE REACTOH CinNICAL COtTLIANCE NUCLEAR OFFICE AD SUPERVISOR (s)
HAIN1tNANCE PilYSICS AllRA MANAGEMFNT ENGINEFR(S)
ENGINFER(S)
ENGINEER (S)
TRAINING SUPERVISOR L
(RO)
SUPERVISOR (S)
ENGINEER (S) kNCINEER(S)
ENGINEER (S)
SPECI ALIST (S)
OMS)
NilCI kAR Hl!CI EAR N f AI.Til PLANT Mt.CilA N IC( S )
PilYS ICS AIARA NUCitAR NilCl. EAR Cll NICAL NUCl. EAR UPERAluR(S)
ANI)
I t CilN IC I AN(S)
IkCllNIC I ANS l'I AN1 I LCllN!C I AN (S IECll ICIAN(S CONTROL CLERICAL (RO)
ATTENDANT (S)
AND Cl.tRKS ATTENDENT(S)
TECllNICI AN(S)
STAFF i
NUCl.FAR AUXil,l ARY OPERATOa(S) l'Oi)E : NShAC - MtHhER OF NSl4AC OllC
- IIMitt R Of 0144.
I'llCRIH I STATION OkCANITATFON 90
- HNC H4 Arloir Or t'14 A TOM I I rl N';r FIGURE 6.2.2 Sko
- N!.C St HlOk ktACI'OR Ol'FRAIOR I. ! C t'NSE g
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6.5 REVIEW AND AUDIT A.
OPERATIONS REVIEW COMMIITEE (ORC) 1.
FUNCTION The ORC shall function to advise the Pilgrim Station Manager on all matters related to nuclear safety.
2.
COMPOSITION The ORC shall be composed of the:
Chairman /
Station Manager Vice Chairman Assistant Station Manager Member:
Methods, Co=pliance & Training Group Leader Member:
Chief Operating Engineer Member:
Chief Technical Engineer Member:
Chief Maintenance Engineer Member:
Chief Radiological Engineer 3.
ALTERNATES Alternate me=bers shall be appointed in writing by the ORC Chairman to serve on a te=porary basis; Eowever, no more than two alternates shall participate in an ORC quorum at any one time.
4 MEETING FREQUENCY The ORC shall meet at least once per calendar month and'as convened by the ORC Chairman.
5.
OUORUM A quorum of the CRC shall consist of the Chairman /Vice Chair =an and two members including alternates.
6.
RESPONSIBILITIES The ORC shall be responsible for:
a.
Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto that affect nuclear safety.
b.
Review of all proposed tests and experiments that affect nuclear safety.
Review of all proposed changes to the Technical Specifications.
c.
d.
Review cf all proposed changes or modifications to plant systems or equipment that affect nuclear safety, Investigation of all violations of the Technical Specifications and e.
shall prepare and forward a report covering evaluation and recom-mendations to prevent recurrence to the Nuclear Operations Manager and to the NSRAC Chair =an.
212
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