ML20211G231
| ML20211G231 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/12/1997 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20211G216 | List: |
| References | |
| NUDOCS 9710020157 | |
| Download: ML20211G231 (168) | |
Text
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RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS Attachment C CHANGE PACKAGE
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I CTS 1.0
- Definitions" Marked Up Pages y
, n CTS 1.0' Definitions" Discussion of Changes 2 CTS 1.0
- Definitions
- No Significant Hazards 3 ITS 1.0 Definitions * - Revised Pages 4 CTS 3/4.8 " Rad Effluents" Marked Up Pages 5 CTS 3/4.8 ' Rad Effluents" Discussion of Changes 6 CTS 3/4.8 ' Rad Effluents" No Significant Hazards 7 CTS 3/4.8 ' Rad Effluents
- Revised Pages S STS 3.7.6 ' Main Condenser Offgas"- Marked Up Pages 9 STS 3.7.6 " Main condenser offgas"- Discussion of Devamrts 10 i STS 3.7.6 Bases Main Condenser Offgas" Marked Up Pages 11 STS 3.7.6 anses ' Main condensor offgas" Discussion o' oeviations12 CTS 4.0
- Rad Material Source'- Marked Up Pages 13 CTS 4.0 ' Rad Material Source'- Discussion of Changes 14 CTS 4.0 ' Rad Material Gource'- No Significant Hazards 15 iQ g CTS 5.0 ' Major Design Features * - Marked Up Pages 16 CTS 5.0
- Major Design Features'- Discussion of Changes 17 CTS 5.0 " Major Design Features *. No Significant Hazards 1S ITS 4.0 " Design Features"- Revised Pages f$
- Design Features * - Marked Up Ptges 20 STS 4.0 ' Design Features * - Discussion of Deviations 21 CTS 6.0 ' Administrative Controls"- Marked Up Pages 22 CTS 6.0 " Administrative Controls * - Discussion of Changes 23 CTS 6.0 ' Administrative Controls" No Significant Hazards 24 ITS 5.0
- Administrative Controls"- Revised Pages 25 STS 5.0 " Administrative Controls * - Marked Up Pages 26 STS 5.0
- Administrative Controts'- Discussion Of Deviations 27 CTS 7/8.0 ' RAD Environ Monitoring"- Marked Up Pages 28 OJ -
CTS 7/8.0 ' RAD Environ MonMoring* - Discussion of Changes 2$
CTS 7/8.0 ' RAD Environ Monitoring * - No Significant Hazards 30 31 NM
TABLE OF CONTENTS 1.0 DEFINITIONS 11 2.0 SAFETY LIMITS 21 2.1 - Safety Limits 21 2.2 Safety Limit Violation 21 BASES B21 LIMITING CONDITIONS FOR OPERA flON - SURVEILLANCE REQUIREMENTS 3.1 REACTOR PROTECTION SYSTEM 4.1 3/4.1-1 BASES B3/4.1 1 3.2 PROTECTIVE INSTRUMENTATION 4.2 3/4.2 1 A. Primary Containment isolation Functions A 3/4.2 1
- 8. Core and Containment Cooling Systems B 3/4.2 1 C.- Control Rod Block Actuation C 3/4.2 2 D. Radiation Monitoring Systems D 3/4.2 2 E. Drywell Leak Detection E 3/4.2 3 F. Surveillance information Readouts F 3/4.2 3 G. Recirculation Pump Trip / Alternate Rod G Insertion 3/4.2-4 H. Drywell Temperature H 3/4.25 BASES B3/4.21 3.3 RE: ACTIVITY CONTROL 4.3 3/4.3-1 Reactivity Limitations A 3/4.3-1 O A.C.B. Control Rods Scram insertion Times B 3/4.3-2 C 3/4.3-4 D. Control Rod Accumulators - D 3/4.3 5-E. Reactivity Anomalies E 3/4.3-6 F. Attemate Requirements 3/4.3-6 G. Scram Discharge Volume G 3/4.3-6 BASES B3/4.3 1 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 3/4.4-1 BASES B3/4.4-1 3.5 CORE AND CONTAINMENT COOLING 4.5 3/4.5-1 SYSTEMS A. Core Spray and LPCI Systems A 3/4.5-1 B. Containment Cooling System B 3/4.5-3 C. HPCI System C 3/4.5-4
- D. Reactor Core Isolation Cooling (RCIC) System D .3/4.5-5 E. Automatic Depressurization System (ADS) E 3/4.5-6 F. Minimum Low Pressure Cooling and Diesel F Generator Availability 3/4.5-7 G. (Deleted) G 3/4.5-8 H. Maintenance of Filled Discharge Pipe H 3/4.5-8 BASES B3/4.5-1 3
.(V PNPS i Amendment No.
- TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 3/4.6-1 A. Thermal and Pressurization Limitations A 3/4.6-1 B. - Coolant Chemistry B 3/4.6 3 C. Coolant Leakage C -3/4.6-4 D. Safety and Relief Valves D 3/4.6-6 E. Jct Pumps - E 3/4.6 7 F. Jet Pump Flow Mismatch F 3/4.6-8 G. - Structural Integnty G 3/4.6-8 H. Deleted H 3/4.6-8
- 1. Shock Suppressors (Snubbers) l 3/4.6-9 BASES B3/4.6-1
- 3.7 CONTAINMENT SYSTEMS . 4.7 3/4.7 1 ;
A. Primary Containment A 3/4.7 1 i B. Standby Gas Treatment System and Control B i Room High Efficiency Air Filtration System 3/4.7 11 C. Secondary Containment C 3/4.7-16 BASES B3/4.7-1 3.8 - PLANT SYSTEMS 4.8 3/4.8-1
.3.8.1 Main Condenser Offgas- 4.8.1 3!4.8-1 3.8.2 Mechanical Vacuum Pump 4.8.2 3/4.8-2 BASES B3/4.8-1 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 3/4.9-1 A. Auxiliary Electrical Equipment A 3/4.9 1 B. Operation with inoperable Equipment B 3/4.9-4 BASES B3/4.9-1 3.10 CORE ALTERATIONS 4.10 3/4.10-1 A. Refueling interlocks A- 3/4.10-1 B. Core Monitoring B 3/4,10-1 >
C, Spent Fuel Pool Water Level C 3/4.10-2 D; Multiple Control Rod Removal D 3/4.10-2 BASES B3/4.10-1 3.11 REACTOR FUEL ASSEMBLY 4.11 3/4.11 1 A. Average Planar Linear Heat Generation Rate A (APLHGR) 3/4.11-1
- B. Linear Heat Generation Rate (LHGR) B 3/4.11-2 C. Minimum Critical Power Ratio (MCPR) C 3/4,11-2 D. Power / Flow Relationship During Power D
- Operation 3/4.11-4 BASES B3/4.11-1 4
O PNPS- ii Amendment No.
- TABLE OF CONTENTS LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.12 FIRE PROTECTION 4.12 3/4.12 1 Alternate Shutdown Panels 3/4.12-1 BASES B3/4.121 3.13 INSERVICE CODE TESTING 4.13 3/4,13-1 A. Inservice Code Testing of Pumps and Valves 3/4.13-1 l BASES B3/4.131 4.0 -. DESIGN FEATURES 4.0-1 4.1 Site Location 4.0-1 4.2 Reactor Core 4.0 1 4.3 Fuel Storage 4.0-1 4.3.1 Criticality. 4.0-1 4.3.2 Drainage 4.0 2 4.3.3 Capacity 4.0-2 4.3.4 . Heavy Loads 4.0-2 5.0 ADMINISTRATIVE CONTROLS 5.0-1
. 5.1 Responsibility 5.0-1 5.2 Organization 5.0-2 5.3 Unit Staff Qualifications 5.0-4 5.4 Procedures .
5.0-5 5.5 Programs and Manuals 5.0-6 Reporting Requirements 5.0-10 0 5.6 5.7 High Radiation Area 5.0-13 O
PNPS iii Amendment No.
1.0 DEFINITIONS Th3 succe: ding frequ:ntly ussd ttrms are explicitly d:fintd so th:t a uniform inttrprIt: tion of the specifications may be achieved.
(] Safety Limit The safety limits are limits below which the reasonable maintenance of the ciedding and primary systems are assured. Exceeding such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences, but it indicates an operational deficiency subject to regulatory review.
@ Limitino Safety System Settino (LSSS)
, The limiting safety system settings are settings on instrumentation which initiate the automatic l
protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent margin with normal operation lying below these settings. The margin has been established so that'vith proper operation of the instrumentation the safety limits will never be exceeded.
g Limitino Conditions for Ooeration (LCO)
The limiting conditions for operation specify the minimum acceptable levels of system performance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and ebnormal situations can be safely controlled.
Aj h Core Operatino Limits Report ((COLk 3 + s Al A Lh" As8 The CORRCfEP^T'NG!!Kd5 R&BQB$ is a reload-cycle specific documenbus]
Q 4 J %demente end r=MencJthat provides core operating limits for the current operating reload cycle. These cycle specific core operating limits shall be determined for each reload cycle in accordance with SpecificationlW Plant operation within these operating limits is addressed in individual specifications. %
5.6.5 Ap
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Ew l Ooerable - Operability A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal or emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
>R Operatino Operating means that a system or component is performing its intended functions in its required manner.
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Amendment No. 133,470 4 4
1.0 - DEFINITIONS (Cont)
Ai E Immediate - Immediate means that the required action will be-initiated as soon as practicable considering the safe operation of O the unit and the importance of the required action.
Ai *l Reactor power coeration - Reactot power operation is any operation with the mode switch in the "Etartup' or "Run" position with the reactor critical and above it design power.
Ai *] Hot Standbv Condition - Hot standby condition means operation with coolant temperature greater than 212*F, system pressure less than 600 psig, the main steam isolation valves closed and the mode switch in startup.
A, Cold Condition - Reactor coolant temperature equal to or less than 212'F.
k li2d2 - The reactor mode is that which is established by the mode-selector-switch. The modes includ:lchutdcun, refuel, ;;;rtup and run A L. g hich ::: e: tined c; foll;;.;:
4 Startuo Mode - In this mode the reactor protection scram trip, initiated by main steam line isolation valve closure, is bypassed when reactor pressure is less than 600 psig, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with IRM neutron monitoring system trips and control rod withdrawal interlocks in service.
Ap % Run Mode - In this mode the reactor system pressure is at or above 785 psig and the reactor protection system is energized with APRM protection and RBM interlocks in service.
Ah' hl Shutdown Mode - The reactor is in the shutdown mode when the reactor mode switch is in the shutdown mode position and no core alterations are being performed,
- a. Ilot Shutdown means conditions as above with reactor coolant temperature greater than 212'F.
- b. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212'F.
g: Refuel Mode - The reactor is in the refuel mode when the mode
. switch is in the refuel mode position. When the mode switch is in the refuel position, the refueling interlocks are in service.
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I 1.0 DEFINITIONS (C:nt)
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L, Deslan Power. Design pow:r mtans a study stata powIr livil of 1998 thIrmal megawatts.
O A- Primary Containment Inteority Primary containment integrity means that the
, drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:
- 1. All manual containment isolation valves on lines connected to the reactor
! coolant system or containment which ars . lot required to be open during
, accident conditions are closed.
- 2. At least one door in each airlock is closed and sealed.
- 3. All blind flanges and manways are closed.
- 4. All automatic primary containment isolation valves and all instrument line i
flow check valves are operable or at least one containment isolation valve
- in each line having an inoperable valve shall be deactivated in the isolated condition.
- S. All containment isolation check valves are operable or at least one 4
containment valve in each line having an inoperable valve is secured in the isolated position, y Secondary Containment Inteority - Secondary containment integrity means that the reactor building is intact and the following conditions are met:
- 1. At least one door in each access opening is closed.
2, The standby gas treatment system is operable.
- 3. All automatic ventilation system isolation valves are operable or secured in the isolated position.
4
<fi Operatino Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
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c 4, Refuelino Outaae - Refueling outage is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant afte:
that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous refueling outage, the required surveillance g4 testino need not be oerformed until the next regularly scheduled outage.
(DefMien: U :nd V epp!y)
A-
% Refuelino Interval- Refueling interval applies only to ASME Code,Section XI lWP and IWV surveillance tests. For the purpose of designating frequency of these code tests, a refueling intarval shall
- mean at least once every 24 months.
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1.0 DEFINITIOMS (Cont)
A,s @ Core Alteration Core Alt:r; tion shall be the movement of any fu;l. sources, or r activdy control components, within the reactor vessel with the head removed, and fuelin the vessel. The following exceptions are not considered to be Core Alterations;
- a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement), and
- b. Control rod movement, provided there are no fuel assemblies in the associated core cell.
Suspension of Core Alterations shall not preclude completion of movement of a component to a safe position.
M Reactor Vessel Pressure Unless otherwise indicated, reactor vessel pressures listed in the
- Technical Specifications are those measured by the reactor vessel steam space detectors.
S. The-"al-pasemeters 4p gl' .
{ Minimum Critical Power Ratio (MCPR) the value of critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power.
-Aj M 3, l Transition Boilino - Transition boiling means the boiling r6gime between nucleate and film bolling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with ne.'ther type being completely stable.
Ap 3' l Total Peakino Factor The ratio of the fuel rod surface heat flux to the heat flux of an average too in an identical geometry fuel assembly operating at the core average bundle power.
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As 4, jn.strument Calibration - An instrument calibration means the adjustment of an instrument
'A[ signal output so that it corresponds, within acceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall <.. compass the entire instrument including actuation, alarm or trip ,
Ap ~
[T,] Instrument Channel- An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.
g l 4, l Instrument Functional Test - An instrument functional test means the injection of a simulated signal into the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action.
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4, lastrument Check - An instrument check is a qualitative determination of acceptable operability by observatiori of instrument behavior during operation. This determination shallinclude, where possible, comparison of the instrument with other independent instruments measuring the same variable.
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. - :. - . . . . . . i, 1.0- DEFINITIONS (Cont)
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4 lhl Loaic system Punctional Test - A logic system functional test means a test of all relays and contacts of a logic circuit from
_ v sensor to a :tivated device to . insure components are operable per design intent. Where practicable, action will go to coRpletiont (i.e., pumps will be started and valves opened).
Ag , l hl Trio System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.
g Protective Action - An action initiated by the protection system when a limit-is reached. A protective action can be at a cht.nnel or system level.
Ah lhl Protective Function - A system protective action which results from the protective action of the channels monitoring a particular ,
plant condition, g hl Simulaced Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.
O* Ag 5 Surveillance Precuency - Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.
The Surveillance Frequency establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g.,
transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance-intervals beyond that specified g for surveillances that ar ot performed during refueling outages. The.
- Ap p elimitation of'DefinitiorLasa is based on e qineering judgment and the recognition that the mos't probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements, This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly
. degraded beyond that obtained from the specified surveillance interval.
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1.0 PEFINIT10MB (Cont)
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surveillance Interval The surveillance interval is the calendar tima between surveillance teste, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the instrument, component, or A.s) system is not required to be operable, but the instrument, component, or system shall be tested prior to being declared eperable. The operating cycle interval is 24 months and the 25% tolerance given in efinition fot *svavt:1.tAwcr Aualis applicable. TheJ efueling_ interval is 24 months and the 5 87rWrNec .
7 tolerance criecified irddefinition,[248)is applicable. -tc As' An - lWe Fire Suoeression Water System - A fire suppression water system shall consist of: a water source (s)s gravity tank (s) or pump (s); and distribution paping with associated sectionalizing control or isolation valves. Such valves shall include hydrant post indicator valves and the first valse ahead of the water flow alarm device on each sprinkler, hose str"<p1pe or spray system riser.
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stacaered +-et Basis A staggered-test basis shall consist oft (a) a test schedule for a systems, subsystems, trains, or other designated component. obtained by dividing the specified test interval into D equal subinte rvals (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval.
A@ Source check A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to e radioactive source.
- 1 av eff;it: Ocar c 1cu?.; tion 9;nu;I (ODCM; Ar Offait; do;; c;1culation
=;nual (ODCM) -h:11 b; ; menuel em;.t;ining ttr curr;nt : thodcic;y _nd
.y 7;;;=;;;r; to b; ;;;d f cr th; calculuti;n of of f;it; so;;; du; to k 'i endice.;tiv; 3;;;;u; nd liqui? ;!!!rtr.t e, t h; calculat ic: c! 3;;;su: :nd liquie ;f flu;nt- ;_nitcr.ng in;trum;nt ticr clerc/t rip ;;;pcint;, cad th; conduct of t h; "ediclogical Enviscc ccat:1 Monituring . rogr c.
ly -+hhh Action - Action shall be thit part of a specification which prescribes remedial measures required under designated conditions. BB. 9:cieri;; cf th: Pub Mc+ 9;;ter(s) ci th; public :Pril includ; ; M per;cn; who ;;; not c;cupat ionally ;;;;;ist;f uitt the piar' . Thi;
.g- eetf< cry dc;; not -includ; ;;picy;;r of th; utility, it; cent *cctor;, or 7t w 4sh r; . A1;c ;;clud;d f rc; thi; c;t:3 cry cr; p;;;cn; who ; .tcr th; e W--tc ;;rvic; ; qui; m .t or tc m;h- driiverier. *his cat; , cry do;;
4ftehrd p;;;cn; who u;; portion; cf th; ;it; for 7;;rcoticn 1, 0; cup; tion;I cr other purpo;;; not ;;;;;ist;d with t h; citev l C+r Cit: Scunderv *h; : : bcundary ;; chcun i:. Figu:; .5 '
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1.0 DEF2NITIONS (Cont) W Dadun tr 'rcat cnt Cvotem n b Gaaro:cr ^3dwart e
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i cynter ir that-eye::: ident ified -ir rimerc 4.0 2. h Mou44-He4*Meeeme-Evr' r- Thc liquid reducetc ;rcat: cat eye;cm i; ihat syst-cr i devt i f i cd i , r i* pin 0 . 0 .
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A, F5q Automatic primary containn qt Isolation valves - Are primary containment isolation valves which receive an automatic primary containment group isolation cignal. A *ff s Pressure Boundsrv Leakaae - Pressure boundary leakage shall be leakage I through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall. A, p& Identified Lea,taae - Identified leakage shall bei
- 1. Reactor coolant leakage into drywell collection syttems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or
- 2. Reactor coolant leakage into the drywell atmosphere from sources which are both specifically located and known either not to interfere with the operation of the leakage detection system 8 or not to be Pressure Boundary Leakage.
A, Mr~ Unidentified Leakace - Unidentified leakage shall be all reactor coolant p leakage which is not Identified Leakage. C O v A v gg, - - - - . -. _ A No. 00, 113, 120, 130 ' l _ mend = cat
DISCUSSION OF CHANGES CTS 1.0: DEFINITIONS O 6DMINISTRATIVE CHANGES j Ai This change proposes to reformat Section 1.0 to be consistent with the BWR/4 l Standtird Technical Specifications (STS) NUREG 1433. Specifically, the alpha j designations for the definitions are Joad, all the defined terms have been capitaliznd and are in alphabetical order, and footers have been revised. During )' this reformatting and renumbering process, no technical changes (either actual or interpretational) to the TS were made, unless they were identified and justified. f A This change proposes editorial rewording (either adding or deleting) to be more j consistent with the NUREG 1433. This change will not change any technical j l requirements.
)
I A3 This proposed changJ replaces Specification 0.9.A.4 with 5.6.5. This change results from moving the Administrativa Controls Section 6.0 to Section 5.0 and ; l reformatting to be consistent with NUREG 1433. This change will not change
- any technical requirements. 1 A4 This change proposes to delete ..,"(Definitions U and V apply)"...from the
- definition of Refueling Outage. The definition defines what constitutes a Refueling outage, not how often they occur With respect to applying the maximum allowable extension of 25 percent of the specified surveillance interval j allowed by Definitions U and V, there is no specified interval for a Refueling j Outage. Definitions U and V contain sufficient guidance to ensure surveillances
! and tests are performed within their specified intervals. 4 ! As This change proposes to incorporate the definition for the Offsite Dose i Calculations Manual (ODCM) into Section 5.0, " Programs and Manuals",
- subsection 5.5.1 of the proposed PNPS improved TS. The requirements 1
contained in the DEFINITION are carried forward, with minor editorial rewording to be consistent with NUREG 1433, and result in no technical changes. A. This change p;oposes to delete the footnote because the definition that refers to this footnote is proposed to be relocated outside TS. TECHNICAL CHANGES - MORE RESTRICTIVE None O PNPS-D.O.C-1.0 1 9/12/97
DISCUSSION OF CHANGES CTS 1.0: DEFINITIONS O TECHNICAL CHANGES RELOCATIONS R, This change propnses to relocate the definitions for Member (s) of the Public, Site Boundary, and Radwaste Treatment System to the ODOM because the requirements corresponding to the specifications where these definitions are presently used are proposed to be relocated to the ODCM. Although considered a less-restrictive change, the change from existing restrictions on plant operations is brought about by TS changes discusses elsewhere (changes to section 3/4.8) and omission of these definitions from improved TS does not, by itself, reduce existing restrictions on plant operation. TECHNICAL CHANGES LESS RESTRICTIVE None O PNPS-D.O.C-1.0 2 9/12/97
NO SIGNIFICANT HAZARDS CONSIDERATION CTS 1.0 DdFINITIONS ADMINISTRATIVE CHANGES (Ai , A , A 3, A 4, As and A., Labeled Discussion of Changes for CTS 1.0) These proposed changes involve reformatting, renumbering, and rewording of the Technical Specifications (TS). These changes, since they do not involve technical changes to the Technical Specifications, are administrative, All of the administrative changes contained in the Discussion of Changes for this chapter are addressed by this evaluation. BECo has evaluated this proposed change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change insolve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: Definitions perform a supporting function for other sections of the TS. The proposed change to incorporate the definition for the Offsite Dose Calculations Manual (ODCM) into Section 5.0,
- Programs and Manuals", subsection 5.5.1 of the proposed TS will carry forward the requirements contained in the DEFINITION, with minor editorial O
s rewording to be consistent with NUREG 1433, and result in no technical changes. Since the requirements will remain, the impact on initiators of analyzed events or the assumptions assumed in the mitigation of accidents or transient events will not change. Editorial rewording (either adding or deleting) and reformatting is proposed to provide clarity and does not change any technical requirements.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously Ovaluated because of the following: These proposed changes do not involve a physical alterat.on of the plant (no new or different type of equipment will be installed) or changes in methods goveming normal plant operation. The proposed change will not impose any new or different requirements or eliminate any existing requirements. 3 Does this change involve a significant reduction in a margin of safety? Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: Definitions perform a supporting function for other sections of the TS and the proposed editing, omission or relocation of definitions associated with this change will not, by itself, reduce existing restrictions on plant operations. PNPS-N.S.H.C-1.0 1 9/12/97
(: @ NO SIGNIFICANT HAZARDS CONSIDERATION CTS 1.0 DEFINITIONS T_ECHtJICAL CHA.RGES MORE RESTRICTIVE None TECHNICAL CHANGES RELOCATIONS (R$ Labeled Discussion of Changes for CTS 1.0) The proposed changes relocate the definitions for Member (s) of the Public, Site Boundary, and Redwaste Treatment System to the ODCM. BECo has evaluated this proposed change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluetion is provided for the three categories of the \ significant hazards consideration standards: 1, Does the change involve a significant increas2 in the probability or consequences of an accident previously evaluated? ( Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously k evaluated because of the following: The definitions being proposed for relocation do not impact reactor operation, identify a parameter which is an initial condition assumption for a DBA or transient, identify a 9 significant abnormal degradction of the reactor coolant pressure boundary, and do not provide any mitigation of a design basis event.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: Relocating these definitions will not alter the plant configuration (no new or different type of equipment will be installed) or change methods governing normal plant operation. Relocating requirements will not impose different requirements and adequate control of information will be maintained. Relocating these definitions will not alter assumptions made in the safety analysis and licensing basis.
- 3. Does this change involve a significant reduction in a margin of safety?
The operation of PNPS in accordance with the proposed change will not involve a Ognificant reduction in a margin of safety because of the following: The definitions to be transposed from the Technical Specifications to the ODCM are the same as the existing Technical Specifications. Future changes to the ODCM will be controlled in accordance with proposed technical specification 5.5.1 "Offsite Dose Calculation Manual (ODCM)". O PN PS-N.S.H.C-1.0 2 9/12/97
m ini u NO SIGNIFICANT HAZARDS CONSIDERATION CTS 1.0 DEFINITIONS V TECHNICAL CHANGES LESS RESTRICTIVE Ncne O PNPS-N.S.H.C-1.0 3 9/12/97
1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved. ACTION ACTION shall be that part of a specification which presenbes remedial measures required under designated conditions, AUTOMATIC PRIMARY Are primary containment isolation valves which receive an CONTAINMENT automatic phmary containment group isolation signal. ISOLATION VALVES CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel, The following exceptions are not considered to be CORE ALTERATIONS:
- a. Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special movable detectors (including undervessel replacement); and
- b. Control rod movement, provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position. CORE OPERATING The CORE OPERATING LIMITS REPORT is a reload-cycle LIMITS REPORT (COLR) specific document that provides core operating limits for the current operating reload cycle. These cycle specific core O operating limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these operating limits is addressed in individual specifications. COLD CONDITION Reactor coolant temperature equal to or less than 212'F. DESIGN POWER DESIGN POWER means a steady state power level of 1998 thermal megawatts. FIRE SUPPRESSION A FIRE SUPPRESSION WATER SYSTEM shall consist of: a WATER SYSTEM water source (s); gravity tank (s) or pump (s); and distribution piping with associated sectionalizing control or isolation valves Such valves shallinclude hydrant post indicator valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser. HOT STANDBY HOT STANDBY CONDITION means operation with coolant CONDITION temperature greater than 212'F, system pressure less than 600 psig, the main steam isolation valves closed and the mode switch in startup. IMMEDIATE IMMEDIATE means that the required action will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action. O PNPS 11 Amendment No.
1.0 DEFINITIONS (Cont) INSTRUMENT An INSTRUMENT CAllBRATION means the adjustment of an O CAllBRATION instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter which the instrument monitors. Calibration shall encompass the entire instrument including actuation, alarm or trip. INSTRUMENT CHANNEL An INSTRUMENT CHANNEL means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a sing e trip signal related to the plant parameter monitored by that instrument channel. INSTRUMENT CHECK An INSTRUMENT CHECK is a determination of acceptable operability by observation of instrument behavior during operation. This determination shallinclude, where possible, comparison of the instrument with other independent instruments measuring the same variable. INSTRUMENT An INSTRUMENT FUNCTIONAL TEST means the injection of a FUNCTIONAL TEST simulated s!gnalinto the instrument primary sensor to verify the proper instrument channel response, alarm and/or initiating action. LEAKAGE a. Identified LEAKAGE:
- 1. Reactor coolant LEAKAGE into drywell collection systems, such as pump seal or valve packing leaks, that is ceptured and conducted to a sump or collecting tank, O or
- 2. Reactor coolant LEAKAGE into the drywell atmosphere from sources which are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be Pressure Boundary Leakage,
- b. Unidentified LEAKAGE:
Unidentified LEAKAGE shall be all reactor coolant leakage which is not Identified Leakage.
- c. Pressure Boundary LEAKAGE Pressure Boundary LEAKAGE shall be leakage through a non. Isolable fault in a reactor coolant system component body, pipewall or vessel wall.
LIMITING SAFETY The LIMITING SAFETY SYSTEM SETTINGS are settings on SYSTEM SETTING _ instrumentation which initiate the automatic protective action at a (LSSS) level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin with normal operation lying below these settings. The margin has been established so that with proper operation of the O instrumentation the safety limits will never be exceeded. PNPS 12 Amendment No.
1.0 DEFINITIONS LIMITING CONDITIONS The LIMITING CONDITIONS FOR OPERATION specify the p FOR OPERATION (LCO) minimum acceptable levels of system performance necessary to Q assure safe startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled. LOGIC SYSTEM A LOGIC SYSTEM FUNCTIONAL TEST means a test of all FUNCTIONAL TEST relays and contacts of a logic circuit from sensor to activated device to insure components are operable per design intent. Where practicable, action will go to completion (i.e., pumps will be started and valves opened). MINIMUM CRITICAL The value of critical power ratio associated with the most limiting POWER RATIO (MCPR) assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to cause some point in the assembly to experience boiling transition, to the actual assembly operating power. MODE The reactor MODE is that which is established by the mode-selector switch. The MODES include: Startup MODE in this MODE the reactor protection scram trip, initiated by main steam line isolation valve closure, is bypassed when reactor pressure is less than 600 psig, the low pressure main steam line isolation valve closure trip is bypassed, the reactor protection system is energized with IRM neutron monitoring system trips and y control rod withdrawal interlocks in service. J Run MODE In this MODE the reactor system pressure is at or above 785 psig and the reactor protection system is energized with APRM protection and RBM interlocks in service. Shutdown MODE The reactor is in the shutdown MODE when the reactor mode switch is in the shutdown mode position and no core alterations are being performed,
- a. Hot Shutdown means conditions as above with reactor coolant temperature greater than 212'F.
- b. Cold Shutdown means conditions as above with reactor coolant temperature equal to or less than 212'F.
Refuel MODE The reactor is in the refuel MODE when the mode switch is in the refuel mode position. When the mode switch is in the refuel position, the refueling interlocks are in service.
\
PNPS 13 Amendment No.
1.0 DEFINITIONS (C nt's _ OPERABLE- A system, subsystem, division, component, or device shall be I OPERABILITY OPERABLE or have OPERABILITY when it is capable of performing its specified safety function (s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function (s) are also capable of performing their related suppori function (s). OPERATING OPERATING means that a system or component is performing its intended functions in its required manner. OPERATING CYCLE Interval between the end of one refueling outage and the end of the next subsequent refueling outage. PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY means that the drywell INTEGRITY and pressure suppression chamber are intact and all of the following conditions are satisfied:
- 1. All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.
- 2. At least one door in each airlock is closed and sealed p
d
- 3. All blind flanges and manways are closed.
- 4. All automatic primary containment isolation valves and all instrument line check valves are operable or at least one containment isolation valve in each line having an inoperable valve shall be deactivated in the isolated condition.
- 5. All containment isolation check valves are operable or at least one containment valve in each line having an inoperable valve is secured in the isolated position.
PROTECTIVE ACTION An action initiated by the protection system when a limit is reached. A PROTECTIVE ACTION can be at a channel or system level. PROTECTIVE FUNCTION A system PROTECTIVE ACTION which results from the PROTECTIVE ACTION of the channels monitoring a particular plant condition. REACTOR POWER REACTOR POWER OPERATION is any operation with the mode OPERATION switch in the "Startup" or "Run" position with the reactor critical and above 1% design power. REACTOR VESSEL Unless otherwise indicated, REACTOR VESSEL PRESSURES
,Q PRESSURE listed in the Technical Specifications are those measured by the
() reactor vessel steam space detectors. PNPS 1-4 Amendment No.
f 1.0 DEFINITIONS (c ntinu:d) REFUELING INTERVAL REFUELING INTERVAL applies only to ASME Code, Section XI IWP and l\W surveillance tests. For the purpose of designating O frequency of these code tests, a REFUELING INTERVAL shall mean at least once every 24 months. REFUELING OUTAGE REFUELING OUTAGE is the period of time between the shutdown of the unit prior to a refueling and the startup of the plant after that refueling. For the purpose of designating frequency of testing and surveillance, a REFUELING OUTAGE shall mean a regularly scheduled outage; however, where such outages occur within 11 months of completion of the previous REFUELING OUTAGE, the required surveillance testing need not be performed until the next regularly scheduled outage. SAFETY LIMIT The SAFETY LIMITS are limits below which the reasonable maintenance of the cladding and primary systems are assured. Exceeding such a limit is cause for unit shutdown and review by the Nuclear Re0ulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences, but it indicates an operational deficiency subject to regulatory review. SECONDARY SECONDARY CONTAINMENT INTEGRITY means that the CONTAINMENT reactor building is intact and the following conditions are met: INTEGRITY
- 1. At least one door in each access opening is closed.
- 2. The standby gas treatment system is operable.
- 3. All automatic ventilation system isolation valves are operable or secured in the isolated position.
SIMULATED AUTOMATIC SIMULATED AUTOMATIC ACTUATION means applying a ACTUATION simulated signal to the sensor to actuate the circuit in question. SOURCE CHECK A SOURCti CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source. SURVEILt.ANCE Each Surveillance Requirement shall be performed within the FREQUENCY specified SURVEILLANCE INTERVAL with a maximum allowable extension not to exceed 25 percent of the specified SURVEILLANCE INTERVAL. The SURVEILLANCE FREQUENCY establishes the limit for which the specified time interval for Surveillance Requirements may be extended it permits an allowable extension of the normal surseillance interval to facilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience
] to extend surveillance intervals beyond that specified for surveillances that are not performed during refueling outages.
PNPS 15 Amendment No.
1.0 DEFINITIONS (Ccnt) SURVEILLANCE This limitation of this definition is based on engineering judgment ,l U FREQUENCY (continued) and the recognition that the most probable result of any particular surveillance being performed is the venfication of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval. SURVEILLANCE The SURVEILLANCE INTERVAL is the calendar time between INTERVAL surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the instrument, component, or system is not required to be operable, but the instrument, component, or system shall be tested prior to being declared operable. The operating cycle intervalis 24 months and the 25% tolerance of the definition of " SURVEILLANCE FREQUENCY"is applicable. The refueling intervalis D months and the 25% tolerance specified in the definiticn of
" SURVEILLANCE FREQUENCY"is applicabis.
STAGGERED TEST A STAGGERED TEST BASIS shall consist of: (a) a test BASIS schedule for D systems, subsystems, trains, or other designated components obtained by dividing the specified test intervalinto a equal subintervals; (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval.
~
TRANSITION BOILING TRANSITION BOILING means the boiling regime between nucleate and film boiling. TRANSITION BOILING is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable. TOTAL PEAKING The ratio of the fuel rod surface heat flux to the heat flux of an FACTOR average rod in an identical geometry fuel assembly operating at the core average bundle power. TRIP SYSTEM A TRIP SYS TEM means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A TRIP SYSTEM may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems. q
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PNPS 1-6 Amendment No.
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As LIMITING CONDITIONS FOR ORKEAIlQti Yg . W CE RI MIREMENTS 1.AN SYSITMS 3.8 4RA&MAGT4VfHWF4,yENTE.-(ContT - ^ .8 RfDMACT"M: SP"4rUliWTfr-(Cont )
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a-G___>,n__.s__. R
- 1. {he ;caccr.t rat-4cn of hyd:cg .
. th; cug-:nted offgar t(r-t=ent ;y;;; shall bc l i=X. than ^g cr qual R-I tc0(cdto1:: by voir=7 st--t-he em m.., _ b. .t._ . ..
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fee-i :q..: c:c on Tr a 1p; 3.0 0.
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~ ' ' ' ',_ ,__ Not Required to be perforened until I 31daysafteranySJAEinoperation.j y
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_s I G Main Condenser $ faan =AS systc; d c..ga:d,rcct;;.b'X__ grcc'.._,. 1 .than ? cent by __ M ; rifir t'is- . gamma - Lg:
..z.y y eq.1y~tc . .percent , . The gross Pentebetivityi teete l unc, rc:; tor; th; 4 l r-' / : ;-- - ) scic;;; irate or ncentretica-cf hydr cn-he noble gasesjfr;; th: ; team-.get
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s_ s.____>__s. hour; cr be i,a ccid to be uithin the 14mit of e+mtdcur condition uithin M Cp ificatic- 3.0.0.1 :t the hcur;. . felicuing 'requencic; by ' R6 perfcr-ing . i;; topic :.n:1y;i
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Main condenser Anolicability :
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At all times when steam is t h; stc;- j t air cjcctor
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d.;c..crg:,;l s 0 ,' ' 0 available to the air ejectors. A.i pCi/seen S_necification:
- 1. The grosI['G3 activity l-Wea a. At least once per 31 days.
enc /c: ac. :c) reinasc ime vs noble cases measured at IFe"']+-- I em
- LI t.___ 2.. _ _ _2__ Ma' I b. Once within 4 hw aner a 2 50%
i-. ; , .. -; _.+efshall be limited to 500,000 pCi/sec increase in the nominal steady staic
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..,.,2...a,,
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.,,..m PNPS 3/4.8 1 Amendment No.
J.IMITING CONDITI. OMS FOR 01tfAT1,9H YEILLAECK_RECtUIREMENTS l.AN SYSil.Ms 3,8 h.lJO!OACTIVC CPPLt'C"TC (Cont D - ^-- .8 7AANCACTI"U CPIhl8BNTfr-(Cont) O. "cin =0cnden;e r (Ocnt d [C . ".c i - Conde nc e ;- ; Cont; l o gamma Lt
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<B. Required Action and associated Completion Time l) not met.
< 1. Isolate SJAE within 12 hours OR
[ 2.1 Be in Hot Shutdown within 12 hours. AND 2.2 Be in Cold Shutdown within 36 hours. O (
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GED e aD . anical VNPumo chanicalYm Pum Acolicability: NOTE - I Whenever any main steam isolation valve When a channelis placed in an inoperabl h status solely for the performance of is open with steam flowing.. I required Surveillances, entry into d
.Sptgification: associated Conditione and Required Actions may be delayed for up to 6 hours Four channels of the Main Steam Line } provided the at,sociated Function
, Radiation Monitoring System Radiation - maintains mechanical vacuum pump ! High function for the mechanical vacuum isolation capacity. i pump shall be OPERABLE. - - -- - ! 69319E 1. l l - NOTE-
/
Perform 12 hours. a CHANNEL CHECK e 3 Separate Condition Entry is allowed for 2. Calibrate the trip units every each channel 92 days. ( 3. Perform a Channel CALIBRATION / $ A. One or more required channels every 24 months. The allowable trip 2 inoperable, value shall be s 5.5 x normal f background. 3 l 1. Restore channel to OPERABLE status within 24 hours. 4. Perform a LOGIC SYSTEM FUNCTIONAL TEST including QB t isolation valve actuation every
- 2. . NOTE 24 months.
p i Not Applicable ifinoperable channelis the result of an inoperable isolation valve. l Place channel or associated trip
- ' system in trip within 24 hours.
- B. Required Action and associated Completion Time of Condition A not met, t QB g
Mechanical vacuum pump isolation - capability not maintained. I
- 1. Isolate mechanical vacuum within 12 hours.
M
- 2. Isolate Main Steam Lines within f 12 hours, M
O' 3. Be in Hot Shutdown within 12 hours.
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Main CondIns;r Offgas lasertJ 3/4.8-1 Page I of J
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B 3/4.8 PLANT SYSTEMS O 8 3'4.8.4 u ia comaem er otr a - BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the condenser. Air and noncondensible gases are collected in the condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Augmented Offgas System. The offgas from the main condenser normally includes radioactive gases. Restricting the gross gamma activity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to a member of the public at and beyond the site boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. The Augmented Offgas System has been incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombinar to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser, the water and condensibles are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is normally monitored downstream of the moisture separator prior to entering the holdup line, in the event that the Augmented Offgas System is not in service, the activity rate of noble gases can be monitored at the Steam Jet Air Ejector Offgas Sampling System. APPLICABLE The main condenser offgas gross gamma activity rate is an SAFET( ANALYSES initial condition of the Main Condenser Offgas System failure event, discussed in the FSAR, Section 14.5.6 "Radwaste System Accidents", (Ref.1).This section analyzes the effects, which would result from the failure of the offgas system piping. The gross gamma activity rate is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits of 10 CFR Part 100 (Ref. 2). The main condenser offgas limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). O
Main Cond:nssr Offgas aert 3/4.8-1 Page 2 of 3 O s^ses APPLICABILITY The Specification is applicable when steam is being exhausted to the main condenser and the resulting noncondensibles are being processed via the main condenser offgas system. SPECIFICATION To ensure compliance with the assumptions of the main condenser offgas system failure event (Ref.1), the fission product release rate prior to entering the treatment, adsorption and delay systems should provide reasonable assurance that the potential total body accident dose to an individual at the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR Part 100. The limit for gross gamma activity rate of noble gases was derived using the guidance provided in NUREG 0133 (Ref. 3) and documented in reference 4. ACTIONS AJ If the offgas radioactivity rate limit is exceeded,72 hours is allowed to O < tore tae aro a === otivitv < te to -itaia the ii it. Tae 72 aov< Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins associated with permissible dose and exposure limits, and the low probability of a main condenser offgas system rupture. B.1. B.2.1. and B.2.2 If the gross gamma activity rate is not restored to within the limits in the associated Completion Time, the SJAE must be isolated. This isolates the main condenser offgas system from the source of the radioactive steam. The 12 hour Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems. An attemative to Required Actions B.1 is to place the unit in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least Hot Shutdown within 12 hours and in Cold Shutdown within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. O
taert 3 Main CondInstr Offgts Page J of 3 3/4.8 1 BASES SURVEILLANCE SR 4.8.1.1 REQUIREMENT 3 This SR, on a 31 day Frequency, requires an isotopic analysis of an offgas sample to ensure that the required limits are satisfied. The noble gases to be sampled are Xe-133, Xe-135, Xe-138, Kr-85m, Kr-87, and Kr 88. These are the isotopes important for monitoring fuel integrity, as recommended by the fuel supplier. If the measured rete of radioactivity increases significantly (by 2 50% after correcting for expected increases due to changes in THERMAL POWER or hydrogen injection), an isotopic analysis is also performed within 4 hours after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the fission gas release rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on operating experience. This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after the SJAE is in operation. Only in this condition can radioactive fission gases be in the main condenser offgas system at significant rates. O REFERENCES 1. FSAR, Section 14 5.6.
- 2. 10 CFR Part 100,
- 3. NUREG 0133
- 4. Calculation # PNPS-1-ERHS-Xil.D-8.
BASES: PLANT SYSTEMS 3 / 4 . 8 k'.'.'" C.'.'"rl'J E ""'."Cr0 (Cont) Main Condenser (C -) o air ejector off-ga monitors are provide nd when their tr point is ached, cause an isol ion of the air eject off-gas line. Isola 'on is initiated when th instruments reac their high trip int or one ha. an upscale trip and e other a downscale trip. There is fifteen minu. delay before the at ejector off-gas is- ation valve is losed. This d y is accounted for
- the 30-minute ho up time of the o -gas before it released to the sta ,
Both in ruments are requ ed for trip but the nstruments are designed t t any instrument ilure gives a down ale trip. The rip settings of t . instruments are .t so that the inst taneous stack release rate li - given in Specif tion 3.8 is not e . eded. H. Mechanical Vacuum Pumn e purpose of isolati the mechanical y uum pump line is o limit the re se of activity from e main condenser. During a Contro od Drop Accide fission products ld be transporte from the reactor brough the main .am lines to the co .nser4 The fiss n product radioactivity ould be sensed by e main steam lin radioactivity monitors, initia ng isolation of th mechanical vacu pump, insert 4 O A> r wiaica ir O\ ?. ;ndm;nt "c. M, 15 '. 0 3/'.0 3
Mrchtnical Vccuum Pump ls:l: tion Instrum:ntation Page of 6 - '2 B 3/4.8 PLANT SYSTEMS _O a 3'4.8.2 u ca aic i v ce = = e = m a i i ti a ia tr# at tioa BASES BACKGROUND The mechanical vacuum pump isolation instrumentation initiates a trip of the mechanical vacuum pump and isolation of the associated isolation valve following events in which main steam radiation exceeds predetermined values. Tripping and isolating the mechanical vacuum pump limits the offsite doses in the event of a control rod drop accident (CRDA). The mechanical vacuum pump isolation instrumentation (Ref.1) includes sensors, relays and switches that are necessary to cause initiation of an mechanicM vacuum pump isolation. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an isolation signal to the mechanical vacuum pump isolation logic.- The isolation logic consists of two independent trip systems, with two channels of Main Steam Tunnel Radiation-High in each trip system. Each trip system is a one-out-of-two logic for this Function. Thus, O l either channel of Main Steam Tunnel Radiation-High in each trip _ system are needed to trip a trip system.- The outputs of the channels in a trip system are combined in a logic so that both tip systems must trip to result in an isolation signal. There is one isolation valve associated with this function. APPLICABLE The mechanical vacuum pumo isolation is assumed in the safety SAFETY ANALYSES analysis for the CRDA. The mechanical vacuum pump isolation instrumentation in i tiates an isolation of the mechanical vacuum pump - to limit offsite doses resulting from fuel cladding failure in a CRDA (Ref. 2). The mechanical vacuum pump isolation satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). APPLICABILITY ' The mechanical vacuum pump isolation is required to be OPERABLE only if a main steam line is unisolated with steam flowing. If the main steam lines are isolated, the monitors could not detect high radiation due to the monitor's location downstream of the main steam isolation valves.
Mechanical Vacuum Pump Isolation Instrumsntation Page2of6 3/4.8.2 BASES O SPECIFICATION The OPERABILITY of the mechanical vacuum pump isolation is dependent on the OPERABILITY of the Individual Main Steam Line Radiation-High instrumentation channels, which must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 4.8.2.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated isolation valve. Allowable Values are specified for the mechanical vacuum pump isolation Function specified in the Specification. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CAllBRATIONS. Operation with a trip setpoint less consentative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (i.e., Main Steam Line Radiation-High), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined ac. counting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. ACTIONS A Note has been provided to modify the ACTIONS related to mechanical vacuum pump iso;ation instrumentation channels, The Required Actions for inoperable mechanical vacuum pump isolation instrumentation channels provic'e appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable mechanical vacuum pump isolation instrumentation channel. A.1 and A.2 With one or more channels inoperable, but with mechanical vacuum pump isolation capability maintained (refer to Required Actions B.1,8.2, and B.3 Bases), the mechanical vacuum pump isolation instrumentation is capable of performing the intended function. However, the reliability and redundancy of the mechanical vacuum O
Mtchanical Vccuum Pump Isolation Instrumtntttion
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g ,,,,,G:O ACTIONS A.1 and A.2 (continued) pump isolation instrumentation is reduced, such that a single failure in one of the remaining channels could result in the inability of the mechanical vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the low probability of extensive numbers of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of mechanical vacuum pump isolation,24 hours has been shown to be acceptable to permit restoration of any inoperable channel to OPERABLE status. (Required Action A.1). Altemately, the inoperable channel, or associated trip system, may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable isolation valve, since this may not adequately compensate for the inoperable valve (e.g., the valve may be inoperable such that it will not isolate). If it is not desired to place the channel in trip (e.g., as in the case where placing the Inoperable channel would result in loss of mechanical vacuum), or if the inoperable channel is the result of an inoperable valve, Condition B must be entered and its Required Actions taken. B.1. B.2 and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or other specified condition in which the Specification does not apply. To achieve this status, the plant must be brought to at least Hot Shutdown within 12 hours (Required Action B.3). Altemately, the associated mechanical vacuum pump may be removed from - service since this performs the intended function of the instrumentation (Required Action 8.1). An additional option is provided to isolate the main steam lines (Required Action B.2), which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser. Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the Function not maintaining mechanical vacuum pump isolation capa'.>ility. The Function is considered to be maintaining mechanical vacuum pump isolation capability when sufficient channels are OPERABLE or in trip such that the mechanical vacuum pump isolation instruments will generate a trip ragnal from a valid Main S,eam Line - High signal, and the isolation valve will close. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the
- mechanical vacuum pump isolation valve to be OPERABLE.
Mtchtnical Vccuum Pump Isolation Instrum:ntition 3/4.8.2 age O Bases ACTIONS A.1 and A.2 (continued) pump isolation instrumentation is rewced, such that a singis failure in one of the remaining channels could result in the inability of the mechanical vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE statut . Because of the low probability of extensive numbers of inoperabilities affecting multiple channels, and the low probability of en event requiring the initiation of mechanical vacuum pump isolation,24 hours has been shown to be acceptable to permit restoration of any inoperable channel to OPERABLE status. (Required Action A.1). Attemately, the inoperable channel, or associated trip system, may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable isolation valve, since this may not adequately compensate for the inoperable valve (e.g., the valve may be inoperable such that it will not isolate). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in loss of mechanical vacuum), or if the inoperable channel is the result of an inoperable valve, Condition B must be entered and its Required O ^=o" t * "- B.1. B.2. and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or other specified condition in which the Specification does not apply. To achieve this status, the plant must be brought to at least Hot Shutdown within 12 hours (Required Action B.3). Attemately, the associated mechanical vacuum pump may be removed from service since this performs the intended function of the instrumentation (Required Action B.1).- An additional option is provided to isolate the main l steam lines (Required Action B.2), which may allow operation to continue. Isolating the main steam lines effectively provides an equivalent level of protection by precluding fission product transport to the condenser, Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the Function not maintainin0 mechanical vacuum pump isolation capability. The Function is considered to be maintaining mechanical vacuum pump isolation capability when sufficient channels are OPERABLE or in trip such that the mechanical vacuum pump isolation instruments will generate a trip signal from a valid Main Steam Linel-High signal, and the isolation valve will close. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the mechanical vacuum pump isolation valve to be OPERABLE.
M:ch:nical Vccuum Pump Isolation Instrum:nt: tion 3/4.8.2 sort -- :- -Page4oit ! BASES' . ACTIONS- B.1. B.2. and B.3 (continued) The allowed Completion Time of 12 hours is reasonable, based on operating
- experience, to reach Hot Shutdown from full power conditions, or to remove the condenser pump from service, or to isolate the main steam lines, in an orderly manner and without challenging plant systems.
l SURVEILLANCE The Surveillances are modified by a Note to indicate that when a-REQUIREMENTS channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains mechanical vacuum pump isolation trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be retumed to OPERABLE status or the applicable Condition entered and
. Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform channel Surveillance.
That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the mechanical vacuum pumps will isolate when necessary. _O SURVEILLANCE SR 4.8.2.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a corpparison of the parameter indicated on one channel to a similar-parameter on other channelsJ It is based on the assumption that instrument 1 channels monitoring the same parameter should read approximately the same value._ Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure;. thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CAllBRATION. Agreement criteria are determined by the plant staff based on a combination of the channelinstrument uncertainties, including indication and readability, if a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO. O
M:ch:nical Vacuum Pump Isol: tion Instrum:nt: tion 3/4.8.2 n eert h Page 8 of BASES SURVEILLANCE SR 4,8.2.2 REQUIREMENTS (continued) Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.8.2.3. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 3. SR 4.8.2,3 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy CHANNEL CAllBRATION leaves the channel adjusted to account for instrument drift between successive calibrations consistent with the plant specific setpoint 0'4 methodology. The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 4.8.2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker is incapable of operating, the associated instrument channel (s) would be inoperable. The 14 month Frequency is based on the need to perform this Surveillance under the conditions that apply dunng a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
M;chinical Vccuum Pump Isolation Instrum:nt'; tion 0'2 Insert 4 Page6of6 (v i BASES REFERENCES 1. FSAR Section 7,12.1
- 2. FSAR Section 14.5.1
- 3. BECo Calculation I N1 104, " Review Setpoint Calculation for RM 1705-2A,2B,2C, and 2D - Main Steam High Radiation"
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DISCUSSION OF CHANGES CTS 3/4.6 RADIOACTIVE MATERIALS U ADMINISTRATIVE CHANGES Ai All editorial rewording (either adding or deleting) and renumbering is proposed to clanfy and adjust the specifications remaining, after removal of those l Technical Specifications addressed by Generic Letter 89-01, and does not change any technical requirements. Ed!torial rewording (either adding or deleting) is proposed to bs consistent with l NUREG-1433 to the extent possible, based on the PNPS specific licensing i basis. These proposed changes will result in no technical changes (either actual ! or interpretational) to the Technical Specifications. A This proposed change adds a note to clearly indicate that the first sample for noble gas activity is not required for 31 days after the SJAE is placed in operation. TECHNICAL CHANGES - MORE RESTRICTIVE Mi The proposed change delctos "(referenced to 30 minute holJup)". The limit of 500,000 pCi/sec as calculated in accordance with NUREG 0133 did not take credit for treatment, adsorption or delay. Although the limit for gross noble gas activity is not affected by this deletion, not taking credit for a 30 minute delay is considered a more restrictive change. O ij
' Mr This proposed change revises existing Specification 4.8.G.b with a requirement to verify the gross gamma activity of noble gases within 4 hours after a 2 50%
increase in the nominal steady state fission gas release. Current requirements do not stipulate a time limit. Current requirements also relate the increase to that of the previous day whereas the new requirements do not. The requirement was also editorially changed to be consistent with NUREG-1433 and allowances added for factoring out increases due to hydrogen injection. The addition of a time limit is considered a more restrictive change. M3 This proposed change replaces the existing requirements for actions to be taken when the gross gamma gas activity rate of the noble gases cannot be restored to allowable limits within the required Completion Time. The proposed actions are more restrictive than current requirements; therefore, this is a more restrictive change. M4 This proposed change replaces the current Limiting Conditions for Operation and Surveillance Requirements for the Mechanical Vacuum Pump with requirements for the instrumentation which monitors the main steam line radiatien. Although the existing specification implies the isolation signalis derived from the Main Steam Line Radiation Monitoring System, the proposed actions and surveillance requirements will more appropriately address instrumentation operability. The addition of new requirements is a more restrictivo change. O PNPS - D.O.C - 3/4.8 1 9/12/97
DISCUSSION OF CHANGES CTS 3/4.8 RADIOACTIVE MATERIALS
/^'s TECHNICAL CHANGES - RELOCATIONS Ri This proposed change affects those sections of Technical Specifications dealing with the control of radiological effluents, including those sections dealing with gaseous effluents, liquid effluents, and environmental monitoring. Associated Technical Specifications addressing definitions, administrative controls and reporting are also affected. The purpose of these specifications is to assure compliance with regulatory requirements goveming radioactive effluents, including 10 CFR 20.1302,40 CFR Part 190,10 CFR 50.36a, and Appendix I to 10 CFR 50.
In accordance witn the guidance of Generic Letter 89-01, this proposed change adds new programmatic requirements goveming radioactive effluents and radiological environmental monitoring to the Administrative Controls section of the Technical Specifications. Existing Technical Specifications containing procedural details on radioactive effluents, environmental monitoring, definitions and associated reporting requirements are being deleted. The deleted procedural details are being incorporated into the Offsite Dose Calculation Manual (ODCM). R; This change proposes to relocate the requirements (3/4.8.F.1; Table 3.8.2 and Table 4.8.4 Instrument # 4; and the notes for the Tables) for the main condenser offgas treatment system explosive gas monitoring system to licensee
/Q controlled documents (i.e., FSAR). This instrumentation is used to detect V hydrogen in the main condenser offgas treatment system to ensure that hydrogen concentrations are maintained below the flammability limit. The offgas system is designed to confine detonations without affecting safety-related equipment. The concentration of hydrogen in the offgas stream does not constitute a threat to the public health and safety and is not an initial assumption of any DBA or transient analyus.
R3 This change proposes to relocate details on how to determine that the limit on noble gas activity is met and where to take the sample to the BASES. Any changes to these details will be controlled in accordance with Specification 5.5.6; " Technical Specifications (TS) Bases Control Program
- TECHNICAL CHANGES - LESS RESTRICTIVE L3 This change proposes to revise the specification to specify only measurement of gross gamma activity of noble gases instead of beta and/or gamma activity.
Although this change would appear more restrictive, it is being proposed as less restrictive due to changes in technology which have allowed separation of close energy gamma emitters. Evaluations based solely on gamma activity are now faster and more representative of the whole body dose that would be received by an individual at the site boundary should a release occur.
/N U
PNPS - D.O.C - 3/4.8 2 9/12/97
l DISCUSSION OF CHANGES CTS 3/4.8 RADIOACTIVE MATERIALS _ TECHNICAL CHANGES iLESS RESTRICTIVE La This change proposes to revise the Specification and Surveillance Requirements to allow taking the sample used for isotopic analysis from either pretreatment monitor station. . PNPS has the capability of obtaining a sample either upstream of the offgas recombiner or down stream of the recombiner. The current sampling location (upstream of the recombiner)is subject to hydrogen ignition as discussed in General Electric Co. Service Information Letter (SIL) No.150 and presents a personnel safety issue. - Use of the sample location downstream of the recombiners as the normal sample point will - alleviate this personnel safety issue without affecting the objective of obtaining a sample prior to treatment, adsorption, or delay, This is considered a less 2 restrictive change only because station personnel will now have a choice of where to sample. O O _ PNPS - D.O.C - 3/4.8 3 9/12/97
NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3/4.8
.t',DIOACTIVE MATERIALS O
V ADMINISTRATIVE CHANGAS (Ai and A Labeled 2 Discussio n of Changes for CTS 3/4.8) The proposed changes involve renumbering, and rewording of the Technical Specifications. These changes, since they do not involve technical changes to the Technical Specifications are administrative All of the administrative changes contained in the Discussion of Changes for this chapter are addressed by this evaluation. BECo has evaluated these proposed changes and has determined that they involve no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the o aposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: All editorial rewording (either adding or deleting) and renumbering is made to restructure the section accounting for the requirements relocated in accordance with Generic Letter 89-01. During the editorial rewording and renumbering of the improved A () Technical Specifications, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. Adding a note to clearly indi:: ate that the first sample for noble gas activity is not required for 31 days after SJAE is placed in operation has always been considered the intent of this surveillance requirement. This allowance is consistent with the frequency for the required surveillance and allows time for concentrations of longer lived isotopes to reach equilibrium, in addition, other instrumentation continuously monitors the offgas to alert operators of significant increases in radioactivity.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create ihe possibility of a new or different kind of accident from any accident previously evaluated because of the following: The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change will not impose any new or different requirements or eliminate any existing requirements. p PNPS - N.S.H.C. - 3/4.8 1 9/12/97 l l
NO SIGNIFICANT HAZARDS CONSIDERATIONS ' SECTION 3/4.8
- RADIOACTIVE MATERIALS -
b - ADMINISTRATIVE CHANGES (continued)
- 3. Does this change involve a significant reduction in a margin of safety?
Operation of PNPS in accorJance with the proposed change will not involve a significant reduction in a margin of safety because of the following: The change is administrative in nature and does not involve any technical changes. The proposed change will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Also, because the change is administrative in nature, no question of safety is involved. O x O PNPS - N.S.H.C. - 3/4.8 2 9/12/97
NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3/4.8 RADIOACTIVE MATERIALS f k TECHNICAL CHANGES - MORE RESTRICTIVE (Mi, Ma, Mi and M4 Labeled Discussion of Changes for CTS 3/4.8) This particular No Significant Hazards Considerations Determination is for the changes labeled
" Technical Changes More Restrictive". These changes incorporate more restrictive changes into the current Technical Specifications by either making current requirements more stringent or adding new requirements which currently do not exist. All of the more restrictive changes contained in the Discussion of Changes for this chapter are addressed by this evaluation.
BECo has evaluated these proposed changes and has determined that they involve no significant hazards consideration. This determination has been performed in accordance with the enteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probabiliiy or consequences of an accident previously evaluated because of the following: The proposed change provides more stringent requirements than previously existed in the Technical Specifications. The more stringent requirements will not result in p operation that willincrease the probability of initiating an analyzed event if anything, Q the new requirements may decrease the probability or consequences of an analyzed event by incorporating the more restrictive changes discussed above. The change wil' not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements will not alter the operation of process variables, structures, systems, or components as described in the safety analyses.
- 2. Does the change create the possibility of a new or different kind of acciaent from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: Making existing requirements more restrictive and adding more restrictive requirements to the Technical Specifications will not alter the plant configuration (no new or different type of equipment will be installed) or change methods governing normal plant operation. These changes are cer.sistent with current design bases, licensing bases or assumptions made in the safety analysis.
- 3. Does this change involve a significant reduction in a matgin of safety?
Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: 4 U Adding these new requirements and making existing ones more restrictive does not affect any safety ana!ysis assumptions. As such, no question of safety is involved. PNPS - N.S.H.C. - 3/4.8 3 9/12/97 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3/4.8 RADIOACTIVE MATERIALS t) IECHNICAL CHANGES - RELOCATIONS (Rn R 2and R3 Labeled Discussion of Changes for CTS 3/4.8) These proposed changes relocate requirements from the Technical Specifications to a licensee controlled document. These changes are labeled " Technical Changes Relocations." These changes are listed below. All of the relocation changes contained in the Discutsion of Changes for this chapter are addressed by this evaluation. BECo has evaluated this proposed change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: These proposed changes relocate requirements from the Technical Specifications to the T. S. BASES, FSAR, or ODCM. The licensee controlled document containing the relocated requirements will be maintained using the provisions of 10 CFR 50.59 or a change contu process in the Administrative Controls Section of the Technical (n") Specifications. Since any changes to these licensee controlled documents will be evaluated per an NRC approved change control process, no increase in the probability or consequences of an accident previously evaluated will be allowed.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident pruiously evaluated because of the following: These changes do not alter the plant configuration (no new or different type of equipment will be installed) or methods goveming normal plant operation. These changes will not impose different requirements and adequate control of information will be maintained. These changes do r.ot alter assumptions made in the safety analysis and licensing basis.
- 3. Does this change involve a significant reduction in a margin of safety?
Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: The requirements to be relocated from the Technical Specifications to the FSAR , (S T S. BASES, or ODCM are the same as the existing Technical Specifications and any () future changes to this licensee controlled document will be evaluated per an NRC approved change control process. PNPS - N.S.H.C. - 3/4.8 4 9/12/97 l
NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3/4.8 RADIDACTIVE MATERIALS n b TECHNICAL CHANGES - LESS RESTRICTIVE (Li Labeled Discussion of Changes for CTS 3/4.8) This change proposes to revise the specification to specify only the measurement of gross gamma activity rate of noble gases instead of beta and/or gamma activity. Although this change would appear more restrictive, it is being proposed as less restrictive due to changes in technology which have allowed separation of close energy gamma emitters. Evaluations based solely on gamma activity are now faster and more representative of the whole body dose that would be received by an individual at the site boundary ,hould a release occur. BECo has evaluated this proposed change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards: 1 Does the change involve a significant increase in the probability or consequences of an accicient previously evaluated? Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: Basing the potential fission product release rate on gross gamma activity rate is more p) ( representative of tho whole body dose that would be received by an individual at the site boundary should a release occur. Therefore, reasonable assurance that the potential whole body accident dose to an individual at the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR Part 100 is maintaincd.
- 2. Does the change create the possibility of a naw or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: The proposed change will not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods goveming normal plant operation. Operation of the plant will not be altered by this change. This change will not place the plant in any new condition or introduce any mode of operation not previously analyzed.
- 3. Does this change involve a significant redaction in a margin of safety?
Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: Specifying a release rate based only on gamma activity is more representative of the g whole body dose that would be received by an individual at the site boundary should a T release occur. The actual margin of safety could be increased because potential errors in converting beta activity to whole body exposures are eliminated. PNPS - N.S.H.C. - 3/4.8 5 9/12/97
NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3/u RADIDACTIVE M ATERIAl.S TECHNICAL CHANGES llSS RESTRICTIVE (La Ltbeled Discussion of Changes for CTS 3/4.8) This change proposes to revise the Specification and Surveillance Requirements to allow taking the sample used for Isotopic analysis from either pretreatment monitor station. PNPS has the capability of obtaining a sample either upstream of the offgas recombiner or down stream of the recombiner. The current sampling location (upstream of the recombiner)is subject to hydrogen ignition as discussed in General Electric Service Information Letter (SIL) No.150 and presents a personnel safety issue. Use of the sample location downstream of the recombiners as the normal sample point will alleviate this personnel safety issue without affecting the objective of obtaining a sample prior to treatment, adsorption, or delay. This is considered a less restrictive change only because station personnel will now have a choice of where to sample. BECo has evaluated this proposed change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or censequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a
, significant increase in the probability or consequences of an acident previously \ evaluated because of the following:
Allowing the sample to be taken from either pretreatment monitor station will have no effect on the objective of assuring that the potential whole body accident doss tc an individual at the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR Part 100, because both monitor stations are prior to treatment, adsorption, or delay of the noble gases.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: The proposed thange will not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods goveming normal pl?nt operation. Operation of the plant will not be altered by this change. This change we not place the plant in any new condition or introduce any mode of operation not previously analyzed. O PNPS N.S.H.C, - 3/4.8 6 9/12/97
NO SIGNiflCANT llAZARDS CONSIDERATIONS SECTION 3/4 8 RADIDACTIVE MATERI ALS ( TECHNICAL CHANGES - LESS RESTRICTIVE (La Labeled Discussion of Changes for CTS 3/4.8) (continued)
- 3. Does this change involve a significant reduction in a margin of safety?
Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: The sample used to determine the gaseous activity rate will continue to be taken prior to treatment, adsorption, or delay of the noble gases O O PNPS - N.S.H.C. - 3/4.8 7 9/12/97
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 PLANT SYSTEMS 4.8 ELl,A_NT SYSTEMS
- 1. Main _ Condenser Offans 1. Main Condenser Offaas Aeolicability: 1. -- ---N OT E -
Not Required to be performed At all times when F, team is available to until 31 days after any SJAE in the air ejectors- operation. Specification: Venfy the gross gamma activity rate of the noble gases is The gross gamma activity rate of noble s 500,000 p Ci/second: gases measured at a main condenser protreatment monitor station shall be a. At least once per 31 days. limited to 500,000 p Ci/second. g Action: b. Once within 4 hours after a
- 50% increase in the A. With the gross gamma activity rate nominal steady state fission of the noble gases not within limits; gas release after factoring out increase due to
- 1. Restore the gross gamma chariges in THERMAL activity rate of the noble gases POWER level or hydrogen to within the limit within 72 injection.
hours. B. Required Action and associated O Completion Time not met.
- 1. Isolate SJAE within 12 hours.
QB 2.1 Bo in Hot Shutdown within 12 hours. 8.fiQ 2.2 Be in Cold Ghutdown within 36 hours. O PNPS 3/4.8.1 Amendment No.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 PLANT SYSTEMS (CONT) 48 PLANT SYSTEMS (CONI)
- 2. Mechanical Vacuum PumJ 2. MeJhanical Vacuum Pumo J
Applicabjlh -- --- - N OT E .------ .---.. When a channelis placed in an Whenever any main steam isolation inoperable status solely for the valve is open with steam flowing performance of requireo Surveillances, entry into associated Specification: Conditions and Required Actions may be delayed for up to 6 hours Four channels of the Main Steam Line provided the associated Function Radiation Monitonng System Radiation . maintains mechanical vacuum High function for the mechanical vacuum pump isolation capacity, pump shall be OPERABLE. ~~~~~~~~~~~~~~~~~~~~~~~~ Action:
.. .._...- N O T E -----.---.-. 1. Perform a CHANNEL CHECK Separate Condition Entry is allowed for every 12 hours.
each channel
. ..... .... ...... .... 2. Calibtate the trip units eyery A. One or more required channels 92 days.
inoperable'
- 3. Perform a Channel
- 1. Restore channel to OPERABLE CAllBRATION every 24 status within 24 hours. months. The allowable trip E value shall be s 5.5 x normal p
g 2. . -NOTE ~~ background. Not Applicable if inoperable 4. Perform a LOGIC SYSTEM channelis the result of an FUNCTIONAL TEST including inoperable isolation valve, isolation valve actuation every
. . . . - . . _. . 24 months.
Place channel or associated trip system in trip within 24 hours. B. Required Action and associated Completion Time of Condition A not met. M Mechanical vacuum pump isolation capability not maintained.
- 1. Isolate mechanical vacuum within 12 hours.
M
- 2. Isolate Main Steam Lines within 12 hours.
M
- 3. Be in Hot Shutdown within O 12 hours.
U PNPS 3/4.8.2 Amendment No.
B 3/4.8 PLANT SYSTEMS B 3/4.8.1 Main Condenser Offgas O BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the condenser. Air and noncondensible gases are collected in the condenser, then exhausted through the steam jet air ejectors (SJAEs) to the augmented offgas system. The offgas from the main condenser normally includes radioactive gases. Restricting the gross gamma activity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to a member of the public at and beyond the site boundary will not exceed a small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. The augmented offgas system has been incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and condensables are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is normally measured downstream of the moisture separator prior to O entering the holdup line in the event that the augment 7d offgas system is not in service, the activity rate of noble gases can be measured at the steam jet air ejector offgas sampling system. APPLICABLE The main condenser offgas gross gamma activity rate is an initial SAFETY ANALYSES condition of the main condenser offgas system failure event, discusseo in the FSAR, Section 14.5.6 "Radwaste System Accidents", (Ref.1). This section analyzes the effects, which would result from the failure of the offgas system piping. The gross gamma activity rate is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits of 10 CFR 100 (Ref. 2). The main condenser offgas limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). O (continued) PNPS B3/4.8-1 Amendment No. 1
M:in Cond:ns:r Offgas 3/4.8 1 BASES APPLICABILITY The Specification is applicable when steam is being exhausted to the main condenser and the resulting noncondensables are being processed via the main condenser offgas system. SPECIFICATION To ensure compliance with the assumptions of the main condenser offgas system failure event (Ref.1), the fission product release rate prior to entering the treatment, adsorption, and delay systems should provide reasonable assurance that the potential total body accident dose to an individual at the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR 100. The limit for gross gamma activity rate of noble gases was derived using the guidance provided in NUREG 0133 (Ref 3) and documented in reference 4. [ P ACTIONS AJ If the offgas radioactivity rate limit is exceeded,72 hours is allowed to restore the gross gamma activity rate to within the limit. The 72 hour Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins associated with permissible dose and exposure limits, and the low probability of a main condenser offgas system rupture. B.1. B.2.1. and B.2.2 If the gross gamma activity rate is not restored to within the limits in the associated Completion Time, the SJAE must be isolated. This isolates the main condenser offgas system from the source of the radioactive steam. The 12 hour Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems. An attemative to Required Actions B.1 is to place the unit in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least Hot Shutdown within 12 hours and in Cold Shutdown within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. O (continued) PNPS B3/4.8-2 Amendment No.
! Main Cond:nser Offg:s 3/4.8 1 BASES l SURVEILLANCE SR 4 81.1 REQUIREMJ sTS This SR, on a 31 day Frequency, requires an isotopic analysis of an
- offgas sample to ensure that the required limits are satisfied. The j noble gases to be sampled are Xe 133, Xe 135, Xe 138, Kr 85m, i Kr 87, and Kr 88. These are the isotopes impor1 ant for monitonng l
} fuelintegrity as recommended by the fuel supplier. If the measured l rate of radioactivity in reases significantly (by 2 50% after correcting , ! for expected increases due to changes in THERMAL POWER or i [ 1 hydrogen injection), an isotopic analysis is also performed within 4 hours after the increase is noted, to ensure that the increase la not l indicative of a sustained increase in the fission gas release rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on j operating experience. This SR is modified by a Note indicating that the SR is not required to i be performed until 31 days after the 3JAE is in operation. Only in this l condition can radioactive fission gases be in the main condenser j offgas system at significant rates. l ! REFERENCES 1. FSAR, Section 14.5.6
- 2. 10 CFR Part 100
! 3. NUREG 0133 i
- 4. Calculation # PNPS 1 ERHS XII.D-8 i
a 4 i t t 4 O i
- .PNPS B3/4.8-3 Amendment No.
B 3/4.0 PLANT SYSTEMS B 3/4.8.2 Mechanical Vacuum Pump Isolation Instrumentation n V _0_ASES BACKGROUND The mechanical vacuum pump isolation instrumentation initiates a trip of the mechanical vacuum pump and isolation of the associated isolation valve following events in which main f. team radiation exceeds predetermined values. Tripping and isolating the mechanical vacuum pump limits the offsite doses in the event of a control rod drop accident (CRDA). The mechanical vacuum pump isolation instrumentation (Ref.1) includes sensors, relays, and switches that are necessary to cause initiation of a mechanical vacuum pump isolation. The channels include electronic equipment that compares measured input signals with pre established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an isolation signal to the mechanical vacuum pump isolation logic. The isolation logic consists of two independent trip systems, with two channels of Main Steam Tunnel Radiation High in each trip system. Each trip system is a one-out of two logic for this Functiort Thus, either channel of Main Steam Tunnel Radiation High in each trip system are needed to trip a trip system. The outputs of the channels in a trip system are combined in a logic so that both trip systems must trip to result in an isolation signal. There is one isolation valve associated with this function. APPLICABLE The mechanical vacuum pump isolation is assumed in the safety SAFETY ANALYSES analysis for the CRDA. The mechanical vacuum pump isolation instrumentation initiates an isolation of the mechanical vacuum pump to limit offsite doses resulting from fuel cladding failure in a CRDA (Ref. 2). The mechanical vacuum pump isolation satisfies Criterion 3 of i 10 CFR 50.36(c)(2)(li). l l APPLICABILITY The mechanical vacuum pump isolation ;s required to be OPERABLE only if a main steam line is unisolated with steam flowing. If the main steam lines are isolated, the monitors could not detect high radiation due to the monitor's location downstream of the main steam isolation valves. O V (continued) PNPS B3/4.8-4 Amendment No.
Mcch:nical Vccuum Pump is:lition Instrum:nt: tion 3/4.8 2 BASES SPECIFICATION The OPERABILITY of the mechanical vacuum pump isolation is (- dependent on the OPERABILITY of the individual Main Steam Line Radiation High instrumr.ntation channels, which must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 4 0.2.3. The actual setpoint is cahbrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated isolation valve. Allowable Values are specified for the mechanical vacuum pump isolation Function cpecified in the Specification. Nominal trip setpoints are specified in the setpoint calculations. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CAllBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value,is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (i.e., Main Steam Line Radiation High), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. The trip setpoints are then determined, accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for. ACTIONS A Note has been provided to modify the ACTIONS related to mechanical vacuum pump isolation instrumentation channels. The Required Actions for inoperable mechanical vacuum pump isolation instrumentation channels provide appropriate compensatory measures for separate Inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable mechanical vacuum pump isolation instrumentation channel. A.1 and A.2 With one or more channels inoperable, but with mechanical vacuum pump isolation capability maintained (refer to Required Actions B.1, B.2, and B,3 Bases), the mechanical vacuum pump isolation instrumentation is capable of performing the intended function. However, the reliability and redundancy of the mechanical vacuum (continued) PNPS B3/4.8-5 Amendment No.
Mech:nical Vccuum Pump Isol:ti:n Instrum:nt: tion 3/4.8 2 BASES p ACTIONS A 1 and A 2 (continued) ( pump isolation instrumentation is reduced such that a single failure in one of the remaining channels could result in the inability of the mechanical vacuum pump isolation instrumentation to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the low _ probabilMy of extensive numbers of inoperabilities affecting multiple channels, and the low probability of an event requiring the initiation of mechanical vacuum pump isolation,24 hours has been shown to be acceptable to permit restoration of any inoperable channel to OPERABLE status. (Required Action A.1). Altemately, the inoperabia channel or associated trip system may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channelin trip with no further restrictions is not allowed if the inoperable channelis the result of an inoperable isolation valve, since this may not adequately compensate for the inoperable valve (e.g., the valve may be inoperable such that it will not isolate). If it is not desired to place the channel in trip (e.g , as in the case where placing the inoperable channel would result in loss of mechanical vacuum), or if the inoperable channel is the result of an inoperable valve, Condition B must be entered and its Required Actions taken. OV B.1. B 2. and B.3 With any Required Action and associated Completion Time of Condition A not met, the plant must be brought to a MODE or other specified condition in which the Specification does not apply. To achieve this status, the plant must be brought to at least Hot Shutdown within 12 hours (Required Action B.3). Altemately, the associated mechanical vacuum pump may be removed from service since this performs the intended function of the instrumentation (Required Action B.1). An additional option is provided to isolate the main steam lines (Required Action B.2), which may allow operation to continue. Isolating the main steam lines offectively provides an equivalent level of protection by precluding fission product transport to the condenser. Condition B is also intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the Function not maintaining mechanical vacuum pump isolation capability. The Function is considered to be maintaining mechanical vacuum pump isolation capability when sufficient channels are OPERABLE or in trip such that the mechanical vacuum pump isolation instruments will generate a trip signal from a valid Main Steam Line High signal, and the isolation valve will close. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the mechanical p d vacuum pump isolation valve to be OPERABLE. (continued) PNPS B3/4.8-6 Amendment No.
M:chanical Vacuum Pump Isolati:n instrum:nt; tion 3/4.8 2 BASES ACTIONS B 1. B 2 and B.3 (continued) e The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach Hot Shutdown from full power conditions, or to remove the condenser pump from service, or to it.olate the main steam lines in an orderiy manner and without challenging plant systems. SURVEILLANCE The Surveillances are modified by a Note to indicate that wh0n a REQUIREMENTS channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains mechanical vacuum pump isolation trip capability. Upon completion of the Surveillance or expiration of the 6 hour allowance, the channel must be retumed to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 3) assumption of the average time required to perform a channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the mechanical vacuum pumps willisolate when necessary. SR 4.8.2.1
!3 O Performance of the CHANNEL CHECK once every 12 hours ensures that a Oross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to venfying the instrumentation continues to operate properly between each CHANNEL CAllBRATION.
Agreement enteria are determined by the plant staff based on a combination of the channelinstrument uncert9inties, including indication and readability. If a channelis outside the criteria, it may be an indication that the instrument has drifted outside its limit. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent checks of channels during normal operational use of the displays associated with the required channels of this LCO. O (continued) PNPS B3/4.8-7 Amendment No.
M:ch nical Vacuum Pump Isol: tion Instrum:ntation 3/4.8 2 BASES SURVEILLANCE SR 4,8 2 2 REQUIREMENTS (continued) Calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inogorable if the trip setting is discovered to be less conservative than the Allowable Value specified in SR 3.8 2.3. If the trip setting is discovered to be less conservative than the setting accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 3. SR 4 8 iL;} A CHANNEL CALIBRATION 17 a complete check of the instrument , loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CAllBRATION leaves the channel adjusted to account for instrument drift between successive calibrations, consistent with the plant specific setpoint methodology. The Frequency is based upon the assumption of a 24 month
\ calibration intervalin tha determination of the magnitude of equipment dnft in the setpoint analysis.
SR 4.8 2.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function.
.Therefore, if a breaker is incapable of operating, the associated instrument channel (s) would be inoperable.
The 24 month Frequency is based on the need to perform this , Surveillance under the conditions that apply during a plant outage and the potential for an unp!:anea transient if the Surveillance were performed with the reactor at power. O (continued) PNPS B3/4.8 8 Amendment No.
M:ch:nical Vccuum Pump Isol: tion instrum:ntati:n 3/4.8 2 BASES REFERENCES 1. FSAR Section 7.12.1 O 2. FSAR Section 14.5.1
- 3. BECo Calculation I N1 104,
- Review Setpoint Calculation for RM-1705 2A,28,2C, and 2D Main Steam Line High Radiation".
h (G V(~s, PNPS B3/4.8 9 Amendment No.
y Main Condenser Offges , 3. m
-h [, ym. j
( 8.1 ).
\ 3 PLA!1T SYSTEMS (J
,3. Main Condenser Offgas LCO 3. M The gross gamma activity rate of the noble gases measured at J' _fthe main condenser tv .cust-4cn sys cm pretreatment monitor l_a stationi shall be W40) cl/ semed-tef t er dcroy of 30 -inutc;}. %.__/- B' himitedto 500,000 HC/second i - -----'~
,, x~ _ _ _
APPLICABILITY:
\_ ._x___.-_-
HGDB-+r 4d At all times when steam is available to the air ejectors I
%__n MODCC-2-ond '!Hi~ h l Ei T+ r2r: D E [r N inciated T~iW
*t;;- jet -ir ejcctor %7AG4-4Mpera t : cr. ,
ACTIO!1S _ Col 1DIT1011 REQUIRED ACTIO!1 CCt4PLETIO!1 TIME A. Jross gamma activity A.1 Restore gross gamma 72 hours rate of the noble activity rate of the gases not within noble gases to within n limit. limit. (O' B. Required Action and :scic;c cli m;itt 12 hcute l' 0.1 associated Completion ntcar linc;. 3, G Time not met. , g
% \-
B.2 Isolate SJAE. 12 hours N 3 Hot Shutdow/~h' n P: B.3.1 12 hours Be in W DB-4.]
-.__7 hMD Cold Shutdown 4 p3 k
B.3.2 BeinEpp[j,j 36 hours P3 (n) Sh'", t CTS
\ "
Cc. ;, ^ t / 0 7 / '; 5 2.' - 2 6HM PNPS 3/4.8.1 Amendment 110.
l Main Condenser Offgas 1 3 . h6 p3 4-
~ . , / l (0.1 / l
( SURVEILLANCE REQUIREMENTS s Pg SURVEILIJJJCE FR EC)UENCY
< 4.8.1.1 SR 3.et.1--.}.................. NOTE....................
Not required to be performed until 31 days af ter any [ main steam lim not isolated and) SJAE in operation. Verify the gross gamma activity rate of the 31 days ) noble gases is 5 -fMG)-mci /second (af ter decay of 3dWewt coi . IE2 , [y ~-,'4 500,000 88 Once within k#~~ 4 hours after a 2 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in fm THERMAL POWER
) level d iFhydrogen ,_._.]
4 anjection l L. s f P3 / P
,' N\
f) k m. m,, n,,, =
, , y 7+;q4 n n s
, c;j c pg PNPS 3/4.8.1 Amendment No.
i 1 DISCUSSION OF DEVIATIONS FROM NUREG 1433 Section 3.7.6 i Main Condenser Offgas i O BRACKETED PLANT SPEC',FIC CHANGES Bi This proposed deviation replaces the bracketed limit with PNPS specific values
; and deletes the bracketed reference to a 30 minute holdup since the methodology desentred in NUREG 0133 calculates the limit prior to treatment,
. adsorption, or holdup. i
] B2 This proposed deviation deletes the bracketed Action B.1 [ Isolate all main j steam lines within 12 hours). The option to isolate all main steam lines within 12 i hours is not adopted.
NON BRACKETED PLANT SPECIFIC CHANGES ! Pi This proposed deviation renumbers this LCO to 3.8.1. This LCO will remain in j section 3.8 until the conversion to Stan Jards is complete. l P This proposed Jeviation replaces applicability with current TS opplicability. j PNPS has not yet adopted Standard Technical Specification MODE terminology. 1 P Tit proposed change adds "or hydrogen injection" to also allow for excluding i the need for sampling if increase is due to initiation of hydrogen injection. O l t 4 i i } ) lO 1 1 NUREG 3.7.6 - D.O.D. 1 9/22/97
4
'A 8i Main Condenser - gas 1 y' ] B 3.h4 +
8 ' Restnoting the gross gemme aethrey rete of noble psaet from the mein T l t condenser provklos reasonable assurance that the total body exposure to pg. B 3. PLANT SYSTEMS e member of the publec et and beyond the sRe boundary will not er ed a small fraction of the timRs of 10 CFR 100 in the event this emuer v , j 3.h4 Main Condenser Offgas ' inadvertently discharged directly to the environment WRhout trol Wnt w .f l, BASES l
- BACKGROUND During unit operation, steam from the low pressure turbine l 1s exhausted directly into the condenser. Air and 4 noncondensible gases are collected in the condenser, then l
y; , , exhausted through the steam jet air ejectors (SJAEs) to the l Main Condenser Offgas System. The offgas from the main ! u condenser normally includes radioactive gases. j bDb. \ Thr ";!r .0;rd =;;r Offgas System has been incorporated into ? the Unit design to reduce the gaseous radwaste emission. !
- This system uses a catalytic recombiner to recombine ;
2 radiolytically dissociated hydrogen and oxygen. The gaseous
-pt mixture is cooled by the offgas condensers the water and
. condensibles are stripped out by the offgas condenser and i moisture separator. The radioactivity of the remaining mixture (i.e., the of fgas recorrbiner ef fluent) is I ' measW .-+ gaseous m itsr:dl downstream of the moisture e arator prior to entering the holdup line. lnggg g y p, p g 6 'Redweele System J APPLICABLE The main condenser offgas gross gamma activity rate is an l -SAFETY ANALYSES initial condition of the Main Condenser Offgas System failure event, discussed in the PSAR, Section (15.1.3 W - i j N (Ref. 1) s "h; . Or:1y;i; ;;;uc;; ; ;;;;; f;iltre i th; ".;in g - A ;nd:n;;r Off;;; Cy;ter wh;t r;; ult; in the rupturc Of th; f TNs section analyzes " in 0;nd=;;r Of fgar Cy;ter pr;;;ur; b;und ry. - The gross the effects, which gamma activity rate is controlled to ensure that, during the j would result from the event, the calculated offsite doses will be well within the failure of the offgas limits of 10 CFR 100 (Ref. 2) cr hc '!n0 ;t:f f ; Fir;V;d system piping , , _ _ _ _ , _ _ m _ ,19 ( The main condenser offgas limits satisfy Criterion 2 ef-the y n0 ?;1 icy Lt c nt. Q&iG,0 36(c)(6[ii)] 7, I o (SpecificMa les To ensure compliance with the assumptions of the Main 4 Condenser Offgas System failure event (Ref. 1), the fission l product release rat; rSculd bc ;;n;ist;nt with ; nobic ;;;
\ .- f;!c=; to Or ::;ctor reclant of 100- pip / 'i-:t ;;;;nd efter Qn..n sp=y a = - a= = . coc0ic =tsuana =e==tunh 1
(con;inued) k
- en/: cic r 2 . ' 1- nc; :, 0 /07/0nt:r0 i PNPS B3/4.8-1 Aruendment No.
_ . . . _ _ - ._ _ . , . _ _ _ _ _ . _ - _ . , _ _ , _ , . . _ , , _ . _ , _ . ~ _ _ . . _ , , _ _ _ - - . - _ - . . -
i a 3 1 i Insert (1) in the event that the Augmented Offgas System !s not in service, the activity rate of noble gases can be measured at the Steam Jet Air Ejector Offgas Sampling System. i i 1 J- ! neert(2) 1 i ! prior to entering the treatment, adsorption and delay systems should provide reasonable l 'j assurance that the potential total body accident dose to an indivicial at the exclusion area boundary will not exceed a small fraction of the limits specified in 10 CFR Part 100. The limit for gross gamma activity rate of noble gases was derived using the guidance provided in - NUREG 0133 (Ref, 3) and documented in Reference 4. lO 1 i l
- 2-f i
l i
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Main Condenser Offgas B 3.4,4
' *' ~* 81 BASES l
LCO thi; requiren nt 'l043?? "Nt :: 100 =01/n't ;; ; e..d - (ccati urd'; l0+0? :C1/;;;cndh peelficatRin~p !, APP 1,1CABI1 TTY The beO-is applicable when s eam is bait.g exhausted to the main condenser and the resulting noncondensibles are being processed via the Main Condenser Offgas System. T'.i; scrur; during "000 1, nd i:t:4ng "0000 0 and 3 with any l :1 ;;;;r
< Pe t +!.n; n;t i;;1;ted :.nd: th: CJ'J in p:: tion. 1:. "0000 t
- nd 5, ;;;; i; an being enh;u;ted t; th : 1 ;;nd:n:::
and th ::quir;;;nt; ::: ns; applic;t! . ACTIONS M l If the of fgas radioactivity rate liinit is exceed.d, 72 hours j is allowed to restore the gross gamma activity rate to ; within the limit. The 72 hour Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins associated with permissible dose and exposure limits, and the low probability of a Main Condenser Offgas System rupture. B.1, B.2. B.3.1, and B.3.2 If the gross gammo activity rate is not restored to within , the limits in the associated Completion Time, l:11 ::An L8 + et e;: line; crl the SJAE must be isolated. This isolates the Main Condenser Offgas System from the source of the radioactive steam. -The-=ain ::::: lie e-eee-eene4deced I;clated if ;; les:t on: : i.::::: i;;1; tion c;1 c- in ::ch 3 ;;in ;;;;- lin; i; c!;;;d, rad ;; 1 ::t ;,e ra;i ::::: 1;n drain calc; in : ch drain lin; is cic;cd. The 12 hour Cmpletion Tiine is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems. An alternative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LCO does not apply. To achieve this statub, the unit must be placed in at least MGDB-4 within 12 hours and in "000 within 36 hours. The h Hot~_eS , h V (Cold Shutdown ..
, (continued)
O'c?"/ t OTG B-ih4-2 R. 1, O t / 0 ~'/ 0 5 c.pg PNPS B3/4.6-2 Amendment No.
Main Condenser Offgss B 3.4 r4 y BASES j ACTIONS 3 1. B.2. B.1.1. and B 3.2 (continued) i' allowed Completion Times are reasonable, based or operating
- experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
c. re. t _
\-
' These are the isotopes important for rnontoring SURVEILIANCE SR 3.7.6 1 (m fuelintegtfly as recommended by the fuel REQUIREMENTS uppg- - #
/~ ,
This SR, on a 31 day Frequen'y, requires an isotopic / analysis of an offgas sample to ensure that the requi limits are satisfied. The n le. gases to be sampi are i Xe-1.13, Xe-135, Xe-138, Kr-85, Kr-87, and Kr-88. If the
- k. . . measured rate of radioactivity increases significantly (by
, 2 50% after correcting-for expected increases due to changes a
-or hydrogen ] in THERMAL POWE_R), an isotopic analysis is also performed d"M" # within 4 hours hfter the increase is noted, to ensure that the increase is not indicative of a sustained increase in
- the radioactivity rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on operating experience.
~
4 This SR is modified by a Note indicating that the SR is not j- required to be performed until 31 days after :n3 l::1 ;t::: {$' lin: i; n;; i;;1:ted and' the SJAE is in operation. Only in this condition can radioactive fission gases be in the Main , j Condenser Offgas System at significant rates. REFFRENCES 1. FSAR, Section (15.1.3 %
- 2. 10 CFR 100.
l l 3. NUREG 0133 \ i 946 3
}
- 4. CalculatH)n # PNPS 1 ERHS-Xil.D.8
% s
-- W ,
[O 0"n/ CTC x0 3. ' - 3 n:. 1, 0 /'W-96 l PNPS B3/4.8 Amendment No. l _-_____-____...__,___.,.__-.___._._.,___._,_,._-.m._.,~.__._._ - _ . _ . .
DISCUSSION OF DEVIATIONS FROM NUREG 1433 Section 3.7.6 Bas:s Main Condenser Offgas O BRACKETED PLANT EPECIFIC CHANGES Bi This proposed deviaSon ovietes the bracketed information and incorporates plant specific FSAR section and title. B2 This proposed deviation deletes the bracketed information (all main steam lines or) and the third sentence which clarifies what is considered main Heam isolation. The option to isolate all main steam lines within 12 hours in not adopted. NON-BRACKETED PLANT SPECIFIC CHANGES Pi This proposed deviation renumbers this LCO to 3.8.1 and deletes the header information. This LCO will remain in section 3.8 until the cc. terrion to Standards is complete. P2 Th.t proposed deviation adds information to clearly indicate that the purpose of restricting gross activity rate is to keep exposure at the site bounc try to a small fraction of the limits contained in 10 CFR 100 in the event cf an inadvertent discharge. O e. Thi n<ono ed deviation reniace *ording in NuReG with piani specific terminology. P4 Th!c proposed deviation replaces "menitored" with ' measured'. To moni:or implies that a r'3vice is used b indicate or record whereas the hient is .o determine (measure) the amount (concentration) of rad,os ,tivity in the offgas, i Ps This proposed deviation adds an optional sample location should the normal sample station be out of service. Pe This proposed deviation repaces NUREG wording with plant specific wording from the PNPS FSAR. Pr This proposed deviation replaces "the NRC Policy Statement" with 10 CFR 50.36(c)(2)(li). P. This oroposed deviation replaces NUREG wording with plant specific methodology for deriving the fission product release rate limit. PNPS limits are based on NUREG 0133 methodology O NUREG 3.7.6 Bases - D.O.D. 1 9/22/97
DISCUSSION OF DEVIATIONS FROM NUREG 1433 Section 3.7.6 B:s:s Main Condenser Offgas '
]' This proposed deviation deletes reference to MODES in the Applicability Po discussion and replaces MODE 3 and 4 with Wt Shutdown and Cold Shutdown in Actions B.3.1 and 83.2. PNPS has not yet adopted Standard Technical Specification MODE terminology. , Poi This proposed deviation replaces the isotope KR-85 with KR-85m and adds a statement to explain why the particuiar isctopos were chosen.
NON BRACKETED PLANT SPECIFIC CHANGES l Psi This proposed change adds "or hydrogen injection" to also allow for i excluding the need for sampling if increase is due to initiation of hydrogen injection. P i2 This proposed deviation adds additional references to identify methodology for determining release rate limit.
'(
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DISCUSSION OF CHANGES CTS 4.0 MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES ADMINISTRATIVE CHANGES None TECHNICAL CHANGES MORE RESTRICTIVE None TECHNICAL CHANGES RELOCATIONS Ri This change proposes that the requirements of section 4.0, *MISCELLAb'1EOUS RAOlOACTIVE MATERIALS SOURCES *, be relocated to the FSAR arv implementing procedures. The bases for relocating these requirements are that they duplicate other regulatory requirements contained in 10 CFR 20 and 10 CFR 30 and changes to the licensee controlled documents to which they are selocated are adequately controlled by other regulatory requirements. In accordance with these requirements, the FSAR and implementing procedures will contain adequate detail with respect to sealed source contamination, starveillance requirements, reports, an*1 records retention. Changes to the FSAR and implementing procedures will be ade;uptely controlled by the cited regulations and by the provisions of 40 CFR 50.59. TECHNICAL CHANGES LESS RESTRICTIVE O None 4 O PNPS - D.O.C. - 4.0 1 9/12/97
f NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 4.0: MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES ADMINISTRATIVE CHANQES None IgfHNICAL CHANGES - MORE RESTRICTIVE None TECHNICAL CHANGES - RELOCATIONS This proposed change relocates requirements from the Technical Specifications to a licensee controlled document. These changes are labeled " Technical Changes Relocations." These changes are listed below. (R, Labeled Discussion of Changes for CTS 4.0) Ri This change proposes that the eequirements of section 4.0,' MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES", be relocated to the FSAR. The bases for relocating these requirements are that they duplicate other regulatory requirements contained in 10 CFR 20 and 10 CFR 30 and changes to the licensee controlled documents to which they are relocated are adequately controlled by other regulatory requirements. In accordance with these requirements, the FSAR and implementing procedures will contain adequate G detail with respect to sealed source contaminatbn, surveillance requirements, V reports, and records retention. Changes to the FSAR and implementing procedures will be adequately controlled by the cited regulations and by the provisions of 10 CFR 50.59. BECo has evaluated this proposed change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92, The following evaluation h provided for the three categories of the significart hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: The requirements for miscellaneous radioactive materials do not impact reactor operation, identify a parameter which is an initial condition assumption for a DBA or transient, identify a significant abnormal degradation of the reactor coolant pressure boundary, and do not provide any mitigation of a design basis event. (continued) PNPS - N.S.H.C. - 4.0 1 9/12/97 i
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 4.0: MISCELLANEOUS RADIOACTIVE MATERIALS SOURCES , 4 TECHL, CAL CHANGES. RELOCATIONS (continued)
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The operation of PNPS in accordaace with the propoi ed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: Relocating these requirements will not alter the plant configuration (no new or different type of equipment will be installed) or change methods governing normal plant operation. Relocating te auirements will not impose different requirements and adequate control of information will be maintained. Relocating requirements does not alter assumptions made in the safety analysis and licensing basis, 3, Does this change involve a significant reduction in a margin of safety? The operation of PNPS in accor<1ance with the proposed change will not involve a significont reduction in a margin of safety because of the following: This change relocates requirements from the Technical Specifications to a licensee controlled docement.- This change will not reduce a margin of safety since it has no impact on any safety analysis assumptions. In addition, the requirements to be transposed from the Technical Specifications to the licensee controlled documents are the same as the existing Technical Specificaticns. Since any future changes to these O licensee controlled documents must be ovatetted per the cited regulations or requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed. TECHNICAL CHANGES LESS RESTRICT.JE
-None O
PNPS - N.S.H.C. - 4.0 2 9/12/97
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The reactor (center line) is located approximately 1800 feet from the nearest k{ 5.2 property boundary, REACTOR CORE The reactor vessel core design shall be as described in the CORF OPERATING LIMITS REPORT and shall be limited to these fuel assemblies which have been analyzed with NRC-approved codes and methods and approved by the NRC in its acceptance of Amendment 22 of GESTAR II.
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- a. K,ns 0.95 if fully flooded with water, which includes an allowance for uncenanties as described in Section 10.2.5 of the FSAR; As
( O b. K,n <0.90 when dry, which includes an allowance for uncenanties as descIjbed in Section)Q2.5 of the FSAR; and/
- Me - [ c. A nominal 6.60 inch center to center distance betw een fuel assemblics l I placed in storage racks.
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DISCUSSION OF CHANGES CTS 5.0 DESIGN FEATURES O. U 6DMINISTRATIVE CHANGES Ai All proposed reformatting and renumbering is intended to result in Technical Specifications (TSs) that are more readable, and therefore understandable, by plant operators as well as other users. During the reformatting and renumbering of the improved Technical Specifications, no technical changes (either actual or interpretational) to the TSs were made unless they were identified and justified. A2 Editorial rewording (either adding or deleting) is proposed to be consistent with NUREG-1433 to the extent possible, based on the PNPS specific licensing basis. These proposed changes resulted in no technical changes (either actual or interpretational) to the Technical Specifications. A3 This change proposes to delete the BASES for FUEL STORAGE It consists of Bases for only one section, Section 5.5.C. This Section addressed the Spent Fuel Storage which stipulates maximum K-infinity and U-235 enrichment which will result in a K.n s 0.95. FSAR Section 10.3.5 discusses spent fuel pool design for maintaining K.n for both normal and abnormal conditions. Changes to these requirements / assumptions will require a 10 CFR 50.59 evaluation. The Bases, by definition in 10 CFR 50.36, are not a part of the Technical Specifications. This change does not alter any design feature. This change is consistent with NUREG-1433. O G TECHNICAL CHANGES- MORE RESTRICTIVE Mi Additional information on new fuel storage rack center to center distance between fuel assemblies (FSAR 10.2.4) is being provided to be consistent with NUREG-1433. The addition of these new requirements makes the Technical Specifications more restrictive by adding requirements that did not previously exist in the Technical Specifications. M2 Additional information on the spent fuel storage pool drainage is added to identify the minimum level in the spent fuel pool that ensures safe storage d spent fuel. The addition of thest, ney! requirements makes the Technical Specifications more restrictive by adding requirements that did not previously exist in the Technical Specifications. D (v (continued) PNPS - N.S.H.C. - 5.0 1 9/12/97
DISCUSSION OF CHANGES CTS 5.0 DESIGN FEATURES TECHNICAL CHANGES - RELOCATIONS Ri The design parameters of the reactor vessel and containment remain detailed in FSAR Section 4.2 and Appendix M for the reactor vessel; FSAR Section 5.2 and Appendix L for the primary containment; and FSAR Sections 5.3 and 12.2 for secondary containment. Any changes to these design parameters or requirements must conform to the requirements of 10 CFR 50.59. Furthermore, sufficient detail relating to the reactor vessel and containment exists in LCOs to ensure any changes which may affect safety would require prior NRC review and approval. Since these features will continue to be described in the applicable sections of the FSAR and any changes will require review in accordance with 10 CFR 50.59, the details being relocated do not meet the criteria of 10 CFR 50.36(c)(4) for including as a design feature. Therefore, allowing the removal of these details from Technical Specifications, with their discussion in the FSAR, will not impact safe operation of the facility. R2 The seismic design requirements remain detailed in FSAR Sections 1.6.1.1.8, 2.5.3.4,12.2.3.1.5, and Appendix C. Any changes to these design parameters or requirements must conform to the requirements of 10 CFR 50.59. Since the features, if altered in accordance with 10 CFR 50.59, would not result in a significant effect on safety, the criteria of 10 CFR 50.36(c)(4) for including as a design feature are not met. Therefore, allowing the removal of these details from Technical Specifications, with their discussion in the FSAR, will not impact p safe operation of the facility. O TECHNICAL CHANGES - LESS RESTRICTIVE None b v CTS 5.0 D.O.C. 2 9/12/97 l l 1
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 5.0 DESIGN FEATURES / b ADMINISTRATIVE CHANGES (Ai, A , and A3 Labeled Discussion Of Changes for CTS 5.0) The proposed changes involve reformatting, renumbering, and rewording of the Technical Specifications and Bases. These changes, since they do not involve technical changes to the Technical Specifications are administrative. All of the administrative changes contained in the Discussion of Changes for this chapter are addressed by this evaluation. BECo has evaluated these proposed changes and has determined that they involve no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a significar,t increase in the probability or consequences of an accident previously evaluate i because of the following: The reformatting, renumbering, and rewording along with the other changes listed involve no technical changes to existing Technical Specifications. The proposed changes are administrative in nature and do not impact initiators or assumptions of A analyzed accidents or transient events. ( l
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods goveming normal plant operation. The proposed change will not impose any new or different requirements or eliminate any existing reouirements.
- 3. Does this change involve a significant reduction in a margin of safety?
Operation of PNPS in acce dance with the proposed change will not involve a significant reduction in a margin of safety because of the following: The changes are administrative in nature and do not involve any technical changes. The proposed changes do not impact initiators or assumptions of analyzed accidents or transient events. PNPS - N.S.H.C. - 5.0 1 9/12/97
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 5.0 DESIGN FEATURES O V TECHNICAL CHANGES MORE RESTRICTIVE (Mi and M2 Labeled Discussion Of Changes for CTS 5.0) These changes incorporate more restrictive changes into the current Technical Specifications by either making current requirements more stnngent or adding new requirements which currently do not exist. All of the more restrictiva changes contained in the Discussion of Changes for this chapter are addressed by this evaluation. BECo has evaluated these proposed changes and has determined that they involve no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:
- 1. Does the change involve a significant increase in the probabihty or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the propoced change will not invNve a significant increase in the probability or consequences of an accident previously evaluated because of the following: The proposed change provides more stringent requirements than previously existed in the Technica! Specifications. The more stringent requirements will not result in operation that will increase the probability of initiating an analyzed event, if anything, 'q the new requirements may decrease the probability or consequences of an analyzed g event by incorporating the more restrictive changes discussed above. The change will not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements will not alter the operation of process variables, structurbs, systems, or components as described in the safety analyses.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following: Making existing requirements more restrictive and adding more restrictive requirements to the Technical Specifications will not alter the plant configuration (no new or different
- type of equipment will be installed) or changes in methods goveming normal plant operation. The change does impose different requirements. However, the change is consistent with assumptions made in the safety analyses.
- 3. Does this change involve a significant reduction in a margin of safety?
Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: (] These new or more restrictive requirements are consistent with the current design and V licensing bases; therefore, a margin of safety is not affected. PNPS - N.S.H.C. - 5.0 2 9/12/97
NO SIGNIFICANT llAZARDS CONSIDERATIONS CTS 5.0 DESIGN FEATURES O U TECHNICAL QHANGES RELOCATIONS (Ri and R: Labeled Discussion Of Changes for CTS 5.0) These proposed changes relocate requirements from the Technical Specifications to a licensee controlled document. These changes are labeled " Technical Changes Relocations." All of the relocation changes contained in the Discussion of Changes for this chapter are addressed by this evaluation. BECo has evaluated these proposed changes and has determined that they involve no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three
- categories of the significant hazards consideration standards-
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Operation of PNPS in accordance with the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated because of the following: These proposed changes relocate requirements from the Technical Specifications to the FSAR. Since any changes to the FSAR must be evaluated per 10 CFR 50.59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed. V 2 Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? Operation of PNPS in accordance with the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated because of the following:
- .~
These changes relocate requirements to the FSAR. These changes do not alter the plant configuration (no new or differant type of equipment will be installed) or the methods governing normal plant operation. These changes do not impose different requirements and adequate control of information will be maintained. This change will not alter assumptions made in the safety analysis and licensing basis.
- 3. Does this change involve a significant reduction in a margin of safety?
Operation of PNPS in accordance with the proposed change will not involve a significant reduction in a margin of safety because of the following: These changes relocate requirements from the Technical Specifications to the FSAR. The requirements to be are the same as the existing Technical Specifications. Since any future changes to the FSAR must be evaluated per the requirements of 10 CFR 50.59, no reduction (significant or insignificant) in a margin of safety will be allowed. v E PNPS - N.S.H.C. - 5.0 3 9/12/97
p.mpism- ii.ii.ii,ni ..i. NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 5.0 DESIGN FEATURES l O 1ece~,cA<cN ~ees.ee<oc 1,0~e None O O PNPS - N.S.H.C. - 5.0 4 9/12/97
NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 5.0 DESIGN FEATURES O This proposed Technical Specification Change has been evaluated against the criteria for and identification of licensing and regulatory actions roquiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). The following is a discussion of how the proposed Technical Specification Change meets the criteria for categorical exclusion. 10 CFR 51.22 (c)(9): Although the proposed change involves changes to requirements with respect to inspection or surveillance requirements; (i) the proposed change involves no Significant Hazards Consideration (refer to the No Significant Hazards Consideration section of this Technical Specification Change Request), (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite since the proposed changes do not affect the generation of any radioactive effluents nor do they affect any of the permitted release paths, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Based on the aforementioned and pursuant to 10 CFR 51.22(b), no environmental assessment or environmental impact statement need be prepared in connection with issuance of an amendment to the Technical Specifications incorporating the proposed changes of this request. O PNPS - N.S.H.C. - 5.0 5 9/12/97
Design FC tur:s 4.0 4.0 DESIGN FEATURES 4.1 Site Location Pilgrim Nuclear Power Station is located on the western shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts and contains approximately 517 acres owned by Boston Edison Company as shown on FSAR Figures 2.2-1 and 2.2 2. The site boundary is posted and a pommeter security fence provides a distinct security boundary for the protected area of the station. The reactor (center line)is located approximately 1800 feet from the nearest property boundary. 4.2 Reactor Core The reactor vessel core design shall be as described in the CORE OPERATING LIMITS REPORT and shall be limited to those fuel assemblies which have been analyzed with NRC approved codes and methods and approved by the NRC in its acceptance of Amendment 22 of GESTAR 11. 4.3 Fuel Storage 4.3,1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintc!ned with:
- a. Fuel assemblies having a maximum k-infinity of 1.32 for standard core geometry, calculated at the bumup of maximum bundle reactivity, and an average U-235 enrichment of 4.6 %
averaged over the axial planar zone of highest average enrichment; and
- b. K.n 5 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 10.3.5 of the FSAR.
O (continued) LJ PNPS 4.0-1 Amendment No. i l 1
Design Fe:tur:s 4.0 4.0 DESIGN FEATURES 4.3 FuelStorage (continued) 4.3,1.2~ The new fuel storage racks are designed and shall be maintained with;
- a. K.n 50.95 if fully flooded with water, which includes an allowance for uncertainties as described in Section 10.2.5 of the FSAR;
- b. K.s 50.90 when dry, which includes an allowance for uncertainties as described in Section 10.2.5 of the FSAR; and
- c. A nominal 6.60 inch center to center distance between fuel assemblies placed in storage racks. !
4.3.2 Drainaoe The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 115 ft. 4.3.3 Capacity ( The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3859 fuel assemblies , 4.3.4 Heavy Loads
- a. Loads in excess of 2000 lb. shall be prohibited from travel over fuel assemblies in the spent fuel storage pool.
- b. No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the energy absorbing pad.
O PNPS TS 4.0-2 Amendment No.
D2 sign Facturas 4.0 (' (_)) 4.0 DES!GN FEATURES 4.1 Site Location Pilgrim Nuclear Power Station is located on the western shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts and contains approximately 517 acres owned by Boston Edison Company as shown on FSAR Figures 2.2-1 and 2.2-2. The site boundary is posted and a perimeter security fence provides a distinct security boundary for the protected area of the station. The reactor (center line) is located approximately 1800 feet from the nearest property boundary. 4.2 Reactor Core The reactor vessel core design shall be as described in the CORE OPERATING LIMITS REPORT and shall be limited to those fuel assemblies which have been analyzed with NRC approved codes and methods and approved by the NRC in its acceptance of Amendment 22 of GESTAR 11. G V 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: a, Fuel assemblies having a maximum k-infinity of 1,32 for standard core geometry, calculated at the burnup of ma>Jmum bundle reactivity, and an average U-235 enrichment of 4.6 % averaged over the axial planar zone of highest average enrichment; and
- b. K.n .< 0.95 if fully flooded with unborated water, which includes an ' allowance for uncertainties as described in Section 10.3.5 of the FSAR.
O (continued) U PNPS 4.0-1 Amendment No.
. . - - - - _ _ . - - - . - . - . - . . - - - - . . - -~
D: sign F :tur:s 4.0 4 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3.1.2 The new fuel storags racks are designed and shall be maintained with:
- a. K.n 50.95 if fully flooded with water, which includes an allowance for uncertainties as described in Section 10.2.5 of the FSAR;
- b. K.n s0.90 when dry, which includes an allowance for.
uncertainties as described in Section 10.2.5 of the FSAR; and
- c. A nominal 6.60 inch center to center distance between fuel assemblies placed in storage racks.
4.3.2 Drainaae ! The spent fuel storage poolis designed and shall be meintained to prevent inadvertent draining of the pool below elevation 115 ft. 4,3.3 Capacity i
- The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3859 fuel assemblies ,
4.3.4 Heavy Loads
- a. Loads in excess of 2000 lb. shall be prohibited from travel over fuel assemblies in the spent fuel storage pool.
- b. No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the energy absorbing pad.
l O 4.0-2 Amendment No. PNPS TS 1
Design Features 4.0 l e 4.0 DESIGN FEATURES 4.1 Site Location u
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4.3 Fuel Storage 4.3.1 Critica)ity . 4.3.1.1 The spent fuel storag esigned and shall be maintained withe for standard core geometry, calculated at the burr.up of
--+ maximum bundle reactivity, and an t-
- a. Fuel asse lies mum -Anf nity of 4 .3 .u_ _ _ _ _ ,
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- co d cond it ions' l average U.235 enrichment of 44.M weis'.t pc;ccat % averaged over the adal pianar zone of highest p average enrichment; and 6
- b. K,c 5 0.95 if fully floode d water, which includes an allowance for uncertainties as described in 4Section 4h+ of the PSAR4- and
\ i 10.3.5 Ba .
% (continued) n
- o. . n. , ,i m, 2n6 4.0 1 Re .. .., ,. n. ,i n, , i n, e.
PNPS [ Amendment N<}
Design Features 4.0 I Y
/7 t
4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)
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- 7-~ r 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
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a, r Pit 2:T._ " '/: _ ""' 'h. ~':f_T" ::::" ;"" " 3 m.s... m. p..,...c.-..,, h b. Twer 40.95 if fully flooded with u.bcratcd water, Bli which includes an allowance for uncertainties as f described in 4Section M+ of the PSARt ; 10.2.5
- e. V,r c 40.96 if modcratcd by squ;;us fccm, hich
^
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.. includes an allowance for uncertainties {gd Re described in 4Section M of the FSARt; and
+
10.2.5 g h d. A nominal 46.60t inch center to center distance between fuel assemblies placed in storage racks. o h w En 4.3.2 Drainace The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below clevation 4+0 f t) . 4.3.3 Canacity [BM The spent fuel storage pool is designed and shall be maintained witt a storage capacity limited to no taore than '2005} fuel acr.emblies . " 3859 E G;', 4.3.4 Heavy Loads j a Loads in excess of 2000 lb. shall be prohibited from travel over fuel assemblies in the spent fuel storage pool, b No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the energy absorbing pad JA M O CMn/: CTO 4.0-2 n:. 1, 0:/07/95 Qmendment Noj
SECTION 4.0 QNSERTD O Pilgrim Nuclear Power Station is located on the western shore of Cape Cod Bay in the Town of Plymouth, Plymouth County, Massachusetts and contains approximately 517 acres owned by Boston Edise.. Company as shown on FSAR Figures 2.2-1 and 2.2 2. The site boundary is posted and a perimeter security fence provides a distinct security boundary for the protected area of the station. The reactor (center line) is located approximately 1800 feet from the nearest property boundary. The reactor vessel core design shall be as described in the CORE OPERATING LIMITS REPORT and shall be limited to those fuel assemblies which have been analyzed with NRC approved codes siid methods and approved by the NRC in its acceptance of Amendment 22 of GESTAR 11. O V O
DISCUSSION OF DEVIATIONS FROM NUREG 1433 4 4.0 DESIGN FEATURES
/]
V BRACKETED PLANT SPECIFIC CHANGES Bi This proposed change deletes the brackets and incorporates the PNPS plant specific words from current Technical Specifications. Location cased on latitude and longitude was replaced with a wording taken from the FSAR to provide a more general description of location. 82 This proposed change deletes the brackets and incorporates the PNPS plant , specific words from current Technical Specifications as amanded by Amendment 155. Additional wording for core geometry and average enrichment, as described in Section 10.3.5 of the FSAR, were added for clanfication. 83 This proposed change deletes the brackets and inserts the FNPS plant specific FSAR reference. 84 This proposed change deletes the brackets and incorporates the PNPS plant specific value for K.n of the new feel storage vault for dry conditions from the current Technical Specifications. Bs This proposed change deletes the brackets and inserts the PNPS plant specific value for the center to center distance between fuel assemblies in the new fuel storage vault, as described in Section 10.2.4 of the FSAR. g Be This proposed change deletes the brackets and inserts the PNPS plant specific
; value for elevation, as calculated from drawing C-178, of the siphon-breakers on the fuel pool cooling retum lines. The purpose of these siphon breakers is to prevent inadvertent drain down of the spent fuel pool.
87 This proposed change deletes the brackets and inserts the PNPS plant specific value for fuel assembly storage capacity from the current Technical Specifications. NON-BRACKETED PLANT SPECIFIC CHANGES Pi This proposed change replaces Specifications 4.2.1 and 4 2.2 with PNPS plant specific wording from the current Technical Specifications. Amendment 133 to the PNPS Technical Specifications removed all cycle-specific requirements and placed them in the Core Operating Limits Report (COLR)in accordance with the guidance provided by Generic Letter 8816. P2 This proposed change deletes Specification 4.3.1.1.c. The PNPS spent fuel storage rack design does not rely on center-to-center spacing for maintaining suberitacality. P3 This proposed change deletes Specification 4.3.1.2.a. The PNPS current design and licensing bases do not impose K-infinity limits for the new fuel storage vault. O O (continued) NUREG 1433 - D.O.D. - 4.0 1 9/12/97
DISCUSSION OF DEVIATIONS FROM NUREG 1433 4.0 DESIGN FEATURES O NON-BRACKETED PLANT SPECIFIC CHANGES (continued) U P. This propocod change replaces "if moderated with aqueous foam" with "when dry". PNPS does not have the capability to flood the new fuel storage vault with aqueous foam. Ps This proposed change incorporates Specification 4.3.4, " Heavy Loads", from the current PNPS Technical Specifications. These requirements were added to the PNPS licensing bases by the Safety Evaluation written in support of Amendment 33 [ Q O v NUREG 1433 - D.O.D. - 4.0 2 9/12/97
h 6.0 ADMINISTRATIVE CONTROLS g At 6.1 RESPONSIBILITY , S C 4 'T"5 The Station Director shall be' accounteMe for overall f acility operatio t 4
-4e i
\ hic abacncc, the Ctation Circctee sh Q1 designake7 in w itin thed~ndividue l to :;um; this responsibilityv@5 ring hEsMeng* nsert(l) Mt [succesTs n)
ZATION A A, f Ar Offsite and Onsite crosnizations Onsite and offsite organizations shall be established for unit operation and [5.2.1 corporate management, respectively. The onsite and offsite organizations
- shall include the positions for activities affecting the safety of the nuclear
- power plant. pliough$ b
---+ v
@. Lines of authority, respo sibility, and communication shall be established and defined fee-the highest management levels, through l
intermediate levels, te--and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the-ferm-ef--organization charts, functional descrip;ione i of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. i These requirements shall be documented in the Pilgrim Station Final i Safety Analysis Report.
@. The Station Director shall be responsible for overall ee h safe operation and shall have control over those onsite activities necessary N 3for safe 'peration and maintenance of the plant.
- of theplantf
.v c . The Senior Vice President - Nuclear shall have corporate responsibility , for overall plant nuclear saf ety and shall take any 'aoasures needed to
! ensure acceptable performance of the staff in operating, maintaining, 1 and providing technical s 4u port to the plant to ensure nuclear spfat .7 l @erform ) Qand j j
@. The individuals w o traTn the operating staf f and those who carry out health physics - quality assurance functions may report to the j appropriate onsite manager; however, t M shall have sufficient organizational freedom to ensure their Lndepend ce from operating ga7 p y pressures.
" { these individuals
,5M Unit staff D
- Q g e.
R > h Coch On du w shift ch;11 b; co;F;;;d of et lccet th; --in : = Chi n e 1 7 7 ;_. r g _ ,7 , _m _ y . As - The unit staff organization shall include the following;
- a. A non-licensed operator shall be on site w hen fuel is in the reactor and an additional non-licensed erational mode other than cold shutdown or! '
l
\ operator refueling. shall be assigned w hen the reactor is in anx_f op/
i X - 7 mN o l i l 1 j ' ECViCIOn 17'
' T.C ndmC n t NO, 20, 20, 40, 40, CO, 00, 00, 122, 125, 132 C1 n
i
5 Section 5.0 INSERTS 3 I h The Station Director or his designee shall approve, prior to implementation, each 4 proposed test, experiment or modification to systems or equipment that affect nuclear safety. 4 i i n i l 5.1.2 The Nucear Operations Supe. visor (NOS) shall be responsible for the control ! room command function. During any absence of the NOS from the control room - l while the unit is in an operational mode other than cold shutdown or refueling,
- -an individual with an active Senior Reactor Operator (SRO) license shall be -
designated to assume the control room command function. During any absence [ of the NOS from the control room while the unit is in cold shutdown or refueling, ! en individual with an active SRO license or Reactor Operator (RO) license shall { be designated to assume the control room command function. 4 i j'
~
- O i
3-l i-1 i. 4 i
- l 6
-.=-=wv- c -- .m-~- - , - - - - - - roo ri r . . . , .- -- 1--+,-%-n., , - e-4
.0 ADMINISTRATIVE CONTROLS (Cont) 6.2 ORGANIZATION (Cont) A8 5.22 pn y'M F' . Unit Staff (Cont) / atleastonelicensed j/ QSRO}
/~'N p b ecs b ---
- d 4 When the uni is in an operational mode other than co shutdown or
$5ddiEQ1Q refueling, a person hcidimf-e Senior Reactor Operator Liccnac shall be
% present in the control room at all timce. In addition tc thic Ocnior kQleast o Operatee,-e licensed th;
... . l._.; y when gye i s, Operator tye or Ocnicr_^pc3)c'.or
.[rcay_. shall be present at
- Mm thm .A : ~_
R.ecctor) (,, i Ai Lin the control room }./ 4 8
- w begmmng I
t least two hicensed Owrcto. shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery om reactor trips . d e. nsert (3) La ' [e"4 . individual qualified in radiation protection procedures shall be on A, site when fuel is in the reactor . % nsert (4) h 57 ALL 0000 'LTC"'.T10NO pcrfo mcd while fuc1 i; ir the rcc tcr vcacci cftc : g3 the initici fuci 1 ceding ;P;11 bc dircct'.y cupcrvised by cithcr a liccnacd Oc-ic nccctor Opcrctor cr Oceticr necctor Opcrator ;'mitri to g ~ . Tuc' j'- .dling ut- b r .c 7hcr n..;urrent ; - punuitiilitics during this s operstieer g lasert(5) , _ ,
.A h OCl;;;d g8 3
[ Operations n Department Manager [
} h. . The Chici ^pcrcting Enginccr, Nuclear Watch Engineers, and Nuclear IN Operations Supervisors shall hold a Senior Reactor Operator License.
The Nuclear Plant Operators shall hold a Reactor Operator License. S Insert (6)
.3 UNIT STAFF OUALIFICATIONS *~QQcced the minimum qualifications Thc qu-lification; with rcgsrd to cdu: tionci and cnpc:icnce background; cf
- f. 5.3.1 ,
4 t-heaunit staf f ct the timc of cppointment to th; cctive pccition shall meet "Ifach mem er etc rcquiscm:nta as described in the American National Standards Institute Cihe N1'Y1-1971, " Selection and Training of Personnel for Nuclear Power Plants." In addition, the individual performing the function of Radiation Protection Manager shall meet or exceed the qualifications of Regulatory Guide 1.8, September, 1975. 1.211NO g W __. ._ _m _ _ _ ,_. L_
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As n_7,;;;n ;,,
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. 10, 21, 30, 11, 10, CO, ' ' , 00, 122, 125, 122, it3 6- G l
L Section 5.0 INSERTS 0:
- d. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
- e. Higher grade licensed operators may take the place of lower grade -
licensed or unlicensed personnel. (nsert(4h The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position. O sert (5
- c. The amount of overtime worked by unit staff members performing safety-related functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12).
I. The Shift Control Room Engineer (SCRE) shall provide advisory technical support to the Nuclear Operations Supervisor (NOS) in the areas of engineering and accident assessment, l.1 addition, the SCRE shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift. A Shift Control Room Engineer with a Senior Reactor Operator license may . simultaneously serve as SCRE and SRO. O . i
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- l. .
- a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and ,
- b. The ODCM shall also contain the radioactive effluent controls and radiological l environmental monitoring activities and descriptions of the information that should i be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 and Specification 6.6.3.
- a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
- 1. sufficient information to support the change (s) together with tbs appropriate analyses or evaluations justifying the change (s), and
- 2. a determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302,40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix 1, and not adversely impact ,
- b. Shali become effective after review and acceptance by the Operations Review Committee and the approval of the Chemistry and Radiological Department Managers; and
- c. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made.
- b. Periodic visual inspection to identify and estimate leakage, 5.5.3 Post Accident Samplina This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment
- a. Training of personnel; i
- b. Procedures for sampling and analysis; and I l
- c. Provisions for maintenance of sampling and analysis equipment.
- a. Limitations on the functions l capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
- b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 CFR 20, Appendix B, Table 2, Column 2;
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in j accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
- Page 3 of
- 5.5.4 Radioactive Effluent Controls Proaram (continued)
- d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I; 4
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every
- 31 days; I f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radim.ctivity when the projected doses in a period of I 31 days would exceed 2% of the guidelines for the annual dose or dose
- g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the following:
- any organ.
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;
- i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and J. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources,
- conforming to 40 CFR 190.
- b. Licensees may gwu Wanges to Bases without prior NRC approval provided the changes do not irivous sither of the following:
- 2. a change to the updated FSAR or Bases that involves an unreviewed L safety question as defined in 10 CFR 50.59.
- A tabulation fThEmber'p of station, utility and other personnel (including contractors) receiving exposures ge;;tc:- than-100 mrem /yr and their associated man rem exposure according to work and job functions, (e g. reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (including a description), waste 4 processing, and refueling) sh:11 b; subritted rn annual b;;ia. This tabulation supplements the requirements ' "^ ' 10 CFR 20, The I'*h dose assignment to various duty functions may be estimategbased oon' Q
- a. Core operating limits shall be established .d ; mente ir E0d*"
- ARTS Improvement Program (3) ' Analyses for Pilgrim Nuclear Power Station', dated September 1907, (the approved version at the time the reload analy_ses are performed shall be identified in the C M CE5"w'JINd llj5i58]
- 1. Table 3.1.1 - APRM High Flux trip level setting
- 2. Table 3.2.C -APRM Upscale trip level setting
- 3. 3.11.A - Average Planar Linear Heat Generation Rate (APLHGR)
- 4. 3.11.B -l.inear Heat Generation Rate (LHGR)
- 5. 3.11.C -Minimum Critical Power Ratio (MCPR)
- 6. 3.11.D - Power / Flow Relationship During Power Operation
- 7. 4.2 - Reactor Core O
- c. The core operating limits stall be determined that all A > applicable limits (e.g., f.el thermal mechanical limits, core thermal-hydraulic limits, (ECCS) limit.s, nuclear limits such as shutdown margin, and transient are met *
- d. The CMG-C44AAT4HG-b4HMG-REM *T, including any mid cycle Re revisions or supplements-thereto, shall be provided upon i sua ..,
- euppic;catal scieet cortaining dos; casts; cnts for th prc.;cua
- nual nadicact;ve p f ; g _ _m nq7___ n _7_ _ ; ,
- radioactive liquid and gaseous effluents and solid waste released from the unit. The mat & rial provided shall be consistent with the objectives outlined in the ODCM and i process control procedures and in conformance with 10 CFR 50.36a and 10 CFR 50
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- ADMINISTRATIVE CONTROLS TECHNICA_L CHANGES MORE RESTRICTIVE (continued)
- Programs and Manuals", to the TS as follows:
- REACTOR STARTUP AND POWER TEST PROGRAM".
- Radioactive Materials." Editorial rewording (either adding or deleting) is proposed to be consistent with NUREG 1433.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 3. Does this change involve a significant reduction in a margin of safety?
- 1. Does the change involve a significant increase in the probability or consequences of an accidnnt previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 3. Does this change involve a significant reduction in a margin of safety?
- eduction in a margin of safety because of the following:
- 1. Does the change involve a significant increase in the prooability or consequencec of an accident previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 3. Does this change involve a significant reduction in a margin of safety?
- 3. (continued) invoked by the regulation. Since any changes to a licensee :ontrolled document must continue to meet the regulation, no increase (significant or it significant) in the probability or consequences of an accident previously evalu ated will be allowed.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 3. . Does this change involve a significant reduction in a margin of safety?
- 1. Does the change involve a significant increase la the probability or consequences of an accident previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 2. (continued) J This change proposes to relax the requirement to have an individual qualified in radiation protection procedures to be onsite when fuelis in the reactor. The proposed change will allow the position to be vacant for up to two hours in order to provide for unexpected absence. The proposed change will not create the possibility of an accident. This change will not physically alter the plant (no new or different type of equipment will be installed) or the methods of operation.
- 3. Does this change involve a significant reduction in a margin of safety?
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 3. Does this et nge involve a significant reduction in a margin of safety?
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- 3. . Does this change involve a significant reduction in a margin of safety?
- Operator (RO) license shall be designated to assume the control room command function.
- a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Pilgrim Station Final Safety Analysis Report (FSAR);
- d. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures 5.2.2 Unit Staff The unit staff organization shallincluce the following:
- a. A non-licensed operator shall be on site when fuel is in the reactor and an additional non-licensed operator shall be assigned when the reactor is in an operational mode other than Cold Shutdown or Refueling.
- b. At least one licensed Reactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is in an operational mode other than Cold Shutdown or Refueling, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room.
- c. At least two licensed ROs shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips .
- d. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 5.2.2.a and 5.2.2.1 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.
- e. Higher grade licensed operators may take the place of lower grade licensed or unlicensed personnel.
- f. An individual qualified in radiation protection procedures shall be on p site when fuelis in the reactor. The position may be vacant for not ty more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position,
- g. The amount of overtime worked by unit staff members performing safety-related functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12),
- h. The Operations Department Manager, Nuclear Watch Engineers, and Nuclear Operations Supervisors shall hold a Senior Reactor Operator License. The Nuclear Plant Operators shall hold a Reactor Operator License.
- i. The Shift Control Room Engineer (SCRE) shall provide advisory technical support to the Nuclear Operations Supervisor (NOS) in the areas of engineering and accident assessment. In addition, the SCRE shall meet the qualifications specified by the CommLsion Policy Statement on Engineering Expertise on Shift. A Shift Control Room Engineer with a Senior Reactor Operator license may simultaneously serve as SCRE and SRO.
- a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;
- b. The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG-0737 Supplement 1, as stated in Generic Letter 82 33;
- c. Quality assurance for effluent and environmental monitoring;
- d. Fire Protection Program implementation; and
- e. A.' programs specified in Specification 5.5.
- a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and
- b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release, reports required by Specification 5.6.2 and Specification 5.6.3.
- a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
- 2. a determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20,1302, 40 CFR 190,10 CFR 50.36a, and 10 CFR 50, Appendix 1, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;
- b. Shall become effective after review and acceptance by the Operations Review Committee and the approval of the Chemistry and Radiological Department Managers; and
- c. Shall be submitted to the NRC in the form of a cornplete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shallindicate the date (i.e.,
- a. Preventive maintenance; and
- b. Periodic visual inspection to identify and estimate leakage.
- a. Training of personnel;
- b. Procedures for sampling and analysis; and
- c. Provisions for maintenance of sampling and analysis equipment.
- a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; p (continued)
- b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 CFR 20, Appendix B. Table 2 Column 2;
- c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
- d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas, conforming to 10 CFR 50, Appendix I;
- e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;
- f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the h'v annual dose or dose commitment, conforming to 10 CFR 50, Appendix 1;
- g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the following:
- 1. For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and
- 2. For lodine-131, lodine-133, Tritium, and all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.
- h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; f3 (continued)
- i. Limitations on the annual and quarterly doses to a member of the public from lodine-131, lodine-133 Tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and J. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
- a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
- b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
- 1. a change in the TS incorporated in the license; or
- 2. a change to the updated FSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59.
- c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.
- d. Proposed changes that meet the criteria of Specification 5.5.6b above shall be reviewed and approved by the NRC prior}}