ML19345B424

Proposed Tech Specs,Sections 2.0 & 3.0 Re Single Loop Operation

ML19345B424
Person / Time
Site:	Pilgrim
Issue date:	11/21/1980
From:	BOSTON EDISON CO.
To:
Shared Package
ML19345B422	List: ML19345B421 ML19345B424
References
NUDOCS 8012010241
Download: ML19345B424 (13)

Text

e, ATTAC1?ENT B License and Technical Specification Changes Lic. Page 3 T.S. Pages 6 7

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- 3-B. Technical Specifications i The Technical Specifications contained in Appendices A and B, as revised through A=:nd=ent Me. =rs hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. . Records t

-- Boston Edison shall keep f acility operating records in accord-ance with the requirements of the Technical Specifications. -

D. Equalizer Valve Restriction The valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation.

E.

(Deleted) 4 F. Fire Pr tection  !

• The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.19 of the NRC's Fire Protection Saf ety Evaluation (SE), dated December 21, 1978 f or the f acility. These modifications will be completed in accordance with the schedule in Table 3.1.

In addition, the licensee shall submit the additional information identified in Table 3.2 of this SE in accordance with the schedule contained therein. In the event these dates for submittal cannot be met, the licensee shall submit a report, explaining the cir-cumstances, together with a revised schedule.

The licensee is required to implement che administrative controls identified in Section 6 of the SE. The administrative controls shall be in ef fect by December 31. 1978.

G. Security Plan The licensee shall maintain in effect and fully implement all provisions of the Commission-approved physical security plan including amendments and changes made pursuant to the authority of 10 CFR 50.54(p). The approved security plan consists of l documents withheld from public disclosure pursuant to 10 CFR 2.790, referred to as Pilgrim Nuclear Power Station Physical J

Plan, dated November 7,1977 with Revision 2 dated May 26, 1978 and Revision 3 dated January 8,1979. ,

4. This license is subject to the following condition for the protection -

of the environment: Boston Edison shall continue, for a period of five

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1.1 TAT 2*:T LIMIT 2.'I tb!ITIhG SAFETY SYS~Di SETTE 1.1 TUEL CLADDING INTECRIIT 2.1 FUEL CLADDING YNTICRITT Ame11embilitv: 1eelicability:

Applies to the interrelated Applies to trip settings of the variables associated with fuel instru=ents and devices which are ther=al behavior. provided to Prevent the reactor system eafety limits from being

. exceeded.

Objective: Objective:

To establish '4-4ts belov which To define the level of the process the integrity of the fuel variables at which automatic pro-cladding is preserved. tactive action is initiated to prevent the fuel cladding integrity safany limits from being exceeded.

Soecification: Specification:

A. Reactor Pressure > 800 psia and A. Neutron Flux Scram Core Flav >10t of Rated The existence of a =4-4 .::2 The li=iting saf ety system trip critical power ratio (MCFR) less settings shall be as specified than 1.07 shall constitute vio- below: ,

lation.of 'the fuel cladding integrity s?.fety 14 d t. A MCPR 1. Nuetron 71ux Tric Settines of 1.07 is hereinafter referred to as the Safety Limit MCPR.

a. AFEM 71ux Scram Trie B. Core Thernal Power Linit (Reactor Setting (Run Mode)

Pressure 1800 psia and/or Core Flov (10%)

• When the Mode Switch is

, in the FIN position, the iThen the reactor pressure is 1800 AYEM flux scram trip psia or core flov is less than or setting shall be:

equal to 10% of rated, the steady S4.65W + 55% 2 loop state cord thermal power shall not S2.65W + 51.7% 1 loop exceed 25% of design thermal power.

Where:

C. Power Transient *

- S = Setting in percent The safety li=lt shall be assumed of rated ther=al to be exceeded when scram is known power (1!,98 MWt) to have been accon:plished by a means other than the expected W = Percent of drive scras signal unless analyses flow to produce

[ de=enstrate that the fuel a rated core flov l cladding integrity safety of 69 H lb/hr.

lir.its defined in Specifi- .

cations 1.1A and 1.13 were not .

exceeded during the actual transient. ,

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1.1 . SAFETY LIMIT 2.1 LIMITING SAFETY SYSTN SEmNG D. Whenever the reactor is in the /

In the event of operation with a )

cold shutdown condition with max 1=u= fraction of limiting power '

irradiated fuel in the reactor density (MP.JD) greater than the  !

vessel, the veter level shall not fraction of rated pcVer (FRP),  ;

be less than 12 in. above the top the setting shall be modified as of the normal active fue.1 sone. follows:

pnp -

S f (0.65W + 55% ) . MFLPD ,

2b U S g (0.65W + 51.7%)

, FRP -

I toen l MFLPD Where, -

FRP = fraction of rated thermal power (1998 MWt)

MFLPD = =av4== fraction of limiting

~' power density where the

',~"' limiting power density is 13.4 IN/f t for 8x8 and 78x8R fuel.

The ratio of TRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value vill be used.

For no combination of loop recircula-tion flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 123% of rated

. thermal power.

b. APRM Tlux Scran Trip Settine (Refuel or Start and Hot Standbv Mode)

When the reactor mode switch is in the RE W EL or STARTUP position, the APEM scram shall be set at less than or equal to 151 of rated power.

c. IRM The IRM flux scram setting shall be s120/125 of scale.

B. APRM Rod Block' Trip Setting The APRM rod block trip setting shall be:

3RB f.O 65W + 42: 2 Loop 8

,, 7 RB f 0.65W + 38.7% 1 Loop

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1.1 SkrETY LIMIT 2.1 LIMITINO SAFETY SYSTEM SETTINO Where, IRB = Rod block entting in percent of rated ther=al power (1998 MWt)

W = Percent of drive flow required to produce a rated core fimr of 69M lb/hr. .

In the event of op rating with a maxinu= fraction limiting power ,

density (MFLPD) greater than the fraction of rated power (TEP), the setting shall be modif_ied as f_ollows:

S FRP RB f.(0.65 W + 42%) . MFLPD 9

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RB g (0.65 W 6 38.7%~ FRP 1 Loop MFLPD Where, FRP = fraction of rated thermalpover MFLPD=. max 1=u= fraction of li=iting power density where the limiting power density is 13.4 W/ft for 8x8 and PRx8R fuel.

f The ratio of TRP to MFL?D shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value vill be used. ,

C. Reactor low water level scra=

setting shall bel 9 in. on level instruments.

D. Turbine stop valve closure scram setting shall be s.10 percent valve closure.

E. Turbine control valve fast closure setting shall be }. 150 psig con-trol oil pressure at acceleration relay.

- F. Condenser low vacuum scram setting shall be 2 23 in. Eg. vacuum.

G. Main steam isolation scram setting shall be i 10 percent valve clo-sure.

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.-APR- _.M.Scr.am and Rod Block Trip Limiting Safety Systen l

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setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPF.4 system. As with the APFF. scra= trip setting,'the AFF.M rod block trip setting is adjusted downward if the maximum f raction of limiting power density exceeds the fractica of rated power,. thus preserving the APF.M rod block safety margin. Definition of single loop setpoints is given in Reference 2

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C. Reactor Water low level Scram Trip Setting (LL1)

The set point for lov level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. De results show that scram at this level adequately protects the fuel and the pressure barrier, because MCPR remains well above the safety limit MCPR in all cases,De and system scram -

setting pressure does not reach the safety valve settings. - ,

is approximately 25 in. below the normal operating range and is thus .

adequata to avoid spurious scrama.

D. Turbine Stop valve Closure Scram Trip Setting  !

j The turbine stop valve closure scram anticipates the pressure, neutron l fluz and beat flux increase that could result from rapid closure of I the turbine stop valves. With a scram trip setting ot i 10 percen't of i valve closure from full open, the resultant increase in surface heat fluz is limited such that MCPR re=ains above the safety limit MCPR

( even during the _vorst case transient that assumes the turbine bypass is closed. '

E. Turbine Control valve Tast Closure Seram Trip Setting n e turbine control valve fast closure scram saticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to Deload reactor rejection protection exceeding system theinitiates

- capability of the bypass valves.a scram when fast closure of the control valves is acceleration relay. B is setting and the fact that contrci valve closure ti=e is approximately twice as long as that for the stop valves means that resulting transients, while siwilar, are less severe than for stop valve closure. MCPR remains above the safety limit MCPR.

P. Main Condenser Lov Vacuum Scram Trip Setting

- To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine -

bypass valves.

To anticipate the transient.and automatic scram resulting from the closure of the turbine stop valves. Ice condenser vacuu=The low v initiates a scrs=.

\ a scram before the closure of the turbine stop valves is initiated.

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Transient and accident analyses demonstrate that these conditions result in adequate safety margins for the fuel.

References .

l'. Linford, R. B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NED0-10802, Feb., 1973.

2. BECo. letter (J. E. Howard) to NRC (T. A. Ippolito) titled " Proposed

. Technical Specification Change Concerning Single Loop Operation", dated November 31, 1980.

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Table 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTitUNENTATION REQUIREMENT ~

! Hina. Nomb2r Modeo in Which Function .

l Oparch10 Inst. Hunt En Opsrchie -

Channels per Trip Trip Function Trip Level Setting Refuel (7) Startup/ Hot Run Action (1)

! (1) System Standby 1 Mode Switch in Shutdown X Xi X i i 1 Hanual Scram X X. X A N

3 High Flux s120/125 of full scale X X' (5) A 3 Inoperative X X (5) A

APRM -

2 High Flux '

• (14) (15) (17) (17) X A or B 2 Inoperative 'X X(9) X A or 5 2 Downscale 2 2.5 Indicated on Scale .(11) (11) X(12) A or B 2 - High Flux (15%) 115% of Design Power X X (16) A or B 2 High Reactor Pressure $1085 psig X(10) X X A 2 High Drywell Pressure $2.5 psig 'X(8) X(8) X A

'. 2 Reactor Low Water Level 2 9 In. Indicated Level 'X X X A g-2 High Water Level in Scram Discharge Tank - 139 Callons .X(2) X '

X A 2 Turbine Condenser Low Vacuus 123 In. Hg Vacuum. X(3) X(3) X A or C 2 Main Steam Line High 17X Normal Full Power Radiation Background X X X A or C 4 Main Steam Line Isolation

. Valve Closure 110% Valve Closure X(3) (6) $((3) (6) X(6) A or C 2 Turb. Cont. Valve Fast ?150 psig Cont'rol 011 Closure Pressure at Acceleration Relay X(4) X(4), X(4) A or D

- 4 Turbine Stop Valve Closure $10% Valve Closure X(4) X(4) X(4) A or D U AAPRM high flux scram setpoint 1(.65W + 55) FRP ~ Two rectrc. pump operation

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PHPS TABLE 3.2.c '

• INSntIMENTATION mtAT INITIATES ROD BLOCKS Minimum I of Operable Instrument Trip Level Setting .

Instrument Channeta Per Trip Systems (1)_ ()

APRM Upscale (Flow Two Ioop (0.65W + 42) iift D"' .

2 Biased)

One loop (0.65W + 38.7) Mf' fn ..

2.5 indicated on scale APRM Downscale 2 .

_ TitP (

Two Loop (0.65W + 42) MFLPD Rod Block Monitor L. , -

1 (7)

(Flow Blased)

One Loop _(0.65W + 38.7) .

Rod Block Monitor 5/125 of full scale '

1 (7) Downscale 1RM Downscale (3) 5/125 of full scale ,,

3 (8) 1RM Detector not in

• 3 - Startup Position IRH Upscale <108/125 of full scale 3

SRM Detector not in (4)

$ 2 (5) Startup Position .

SRH Upecale :CIO counts /sec.

2 (5) (6) 4 m

LIMITINO CONDITIONS FOR OPERATION SURVEILLANCE PIQ"IREMENT

~/ 3.6.D Safety and Relief Valves (Cont'd)

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. pressure'shall be below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E. , Jet Pu=ps ,

E. Jet Pumps 1." Whenever the reactor is in'the Whenever there is r.ecirculation flow startup or run modes, all jet with the reactor in the startup or pu=ps shall be operable. If it is run modes, jet pump operability shall determined that a jet pump is be checked daily by verifying that inoperable, an orderly shutdown the following conditions do not oc-shall be initiated and the reactor cur simultaneously:

shall be-in a Cold Shutdown Condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1. The two recirculation loops have a flow imbalance of 15: .

or more when the pumps are operated at the same speed.

2. The indicated value of core flow rate varies from the value derivec from loop flow measurementr by more than 10%.

, 3. The diffuser to lower plenum differential pressure reading on an individual jet pump varies ' rom established ' jet pump P characteristics by

, more than 10%.

F. Jet Pu=p Flow Mis =atch F. Jet Pu=p Flow Mismatch

1. Whenever both recirculation pumps Recirculation pump speeds shall.be are in operation, pu=p speeds shall checked and logged at least once be maintained within 10* of each per day. ,

other when power level is greater than 80* and within 15% of each other when power level is less than or equal to 80%. .

2. When only one recirculation pump is

in opera tion, a) The idle recirculation pump shall be electrically disarmed until such time as it can be restarted in accordance with 3.6.A.4 and 3.6.F.1 b) The recirculation controls shall be -placed in the manual mode t

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r LIMITING CONDITTONS FOR OPERATION SURVEII. LANCE REOUIRD'E::T 3.6.F Jet Pump Flow Mismatch (Cont'd) c) Power level shall be limited to 507' d) Settings for the rod block monitor, 'APRM rod bloc.k trip and flow bias scram

- shall be modified for one

. loop operation in accordance with Specification 2.1 e) MAPLHGR limits shall be reduced as specified in 3.11A.

G. Structural Integrity G. Structural Integrity

1. The s'tructural integrity of 1. The nondestructive inspections the primary system boundary listed in Table 4.6.1 shall be shall be maintained at the performed as specified. The level required by the ASME results obtained from compliance Boiler and Pressure Vessel _w ith this specification will be Code,Section XI, " Rules evaluated after 5 years and the for Inservice Inspection of conclusions of this evaluation Nuclear Power Plant will be reviewed with AEC.

. Components," 1974 F

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.3 5.C St ructural Intenrity (Cen't)

Edition (ASE Code,Section XI) . In the interi: until the nuclear system piping inspection evaluation Icvel criteria of the ASE Boiler and Pressure Vessel Code,Section XI,1974 Editien, are completed, the applicabic cvaluation level provisions of the ASE Boilcr and Pressure Vessel Code,Section XI, 1971 Sumer Addenda shall be used in the Inservice Inspection of nucicar piping.

Components of the primary systen boundary whose in-service ,cxn=ina-tion revesis the absence of flaw indications not in excess of the allowable indication standards of this code are acceptabic for continued se wice. Plant operation with cocponents which have

. in-service examination flaw indication (s) in excess of the

' allevable indication standards of the Code shall be subject to

' NEC approval,

a. Components whose in-scrvice ext.mination reveals flaw indication (s)in excess of the allowable indication standards of the ASE Code,Section XI, are uriaccep-table for continued service unicss the following requirements are met: -

(i) An analysis and evaluation of the detected flaw

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indication (s) shall be submittcJ to the~ NRC that demonstrate that the cocponent structural integrity justifies continued service. The analysis and evaluation shall follow the procedures outlined in Appendix A, 'Tvaluation of Flaw Indications,"

of ASE Code,Section XI.

(ii) Prior to the resumption of service, the NRC shall review the analysis and evaluation and either approve resumption of plant operation with the affected components or require that the component be repaired or replaced.

b. For components approved for continued service in accordance with paragraph a, above, reexamination of the area containing the flaw indication (s) shall be conducted during each scheduled successive in-service inspection. An analysis and evaluation shall be submitted to the NRC following each in-service inspection. The analysis and evaluation shall follow the procedures outlined in Appendix A, . 'Tvaluation of Flaw Indications," of ASE Code,Section XI, and shall reference prior analyses submitted to the NRC to the extent applicable.

Prior to resumption of service following cach in-service inspection, the NRC shall review the analysis and evaluation and either approve resumption of plant operation with the affected components or require that the component be repaired or replaced.

c. Repair or replacement of components, including re-examinations, shall conform with the requiremants of the ASE Code, Section

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XI. In the case of repairs, flaws shall be either removed or repaired to the extent necessary to meet the allowable i indication standards specified in ASE Code,Section XI.

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EUR'.'EILLAEE MQ*JIF DEN'S

'LIMITINO CO CITIONS TOE OPD.ATION 4.11 FIACTOF Wrt ASSDOLT 3.11 REAC~CR WEL ASSD'!* T Applicabi7iti Applic ability _

The surveillance Require:ects '

The Liciting Conditions for Operation apply to the pars:4ters which as sociated with the fuel rods apply the fuel rod operating condi-  ;

to those parameters which monitor the tions.

fuel rod ope, rating conditions. ,

Objective Objective _

The Obje:tive of the Surveil-The Objective of the Limiting Condi- lance Require:ents is to tions for Operation is to assare the specify the type and frequency perfor=ance of the fuel rods. of surveillance to be applied to the fuel rods.

l Specifications Specifications _

A. Average 77=nar Linear Heat A. Average Planar Linear Heat Generation Rate (APLEGR)

Generation Rate ( AFLHOR)

The APLE2 for each type c' .

During power operation with both fuel as a function of sverage recirculation pu=ps operating, the planar exposure shall be APLEGK for each type of fuel as a deter =ined daily during function of average planar exposure reactor operation at jt251 shall not exceed the applicable rated thermal power.

li=iting value shown in Figures 3.11-1 through 3.11-5. The top curves are applicable for core flow greater than or equal to 90% of rated core flow. When core flow is '

less than 90% cf rated core flow, the lower curves shall be limiting. For operation with one recirculation pump, values from these curves are to be multiplied by 0.65. If at any time during operation it is deter =ined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated within 15 .

minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2) hours ,

the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding ,

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action shall continue until reactor l operation is within the prescribed limits.

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, B ASES 3.11A Averste 71snar Line ar Erst Ceneratic9 bte (A?uS Q nis specifications assures that the peak cladding te=perature folleving the postuisted design basis Icss-of-coolant secident will net exceed the 11=1 specified in the 10 cr2 50, Appendix K.

De peak cladding te:perature (PCT) following a postulated ,

loss-of-coolant accide=t is pri=srily a function of the average heat gerieration rate of all the rods of a fuel asse=bly at any

- axial locatica snd is cely dependent, secondarily on the rod to red power distribution within an asse=bly. ne Peak clad te=perature is calculated assuning a UCR for the highest ~

Povered rod which is equal to er less than the design LHOR.

This LB;R ti=ce 1.02 is used i= the beat-up code along with the exposure dependent steady state gap conductance a=d

- rod-to-rod local peaking factors. ne limiting value for AFLEGR is this LEGR of the highest povered rod divided by its local peaking factor.

The calculational procedure used to establish the AP GGR limit for each fuel type is based on a loss-of-coolant accident analysis.

The emergency core cooling systen (ECCS) evaluation models which are e= ployed to deter =ine the ef fects of the loss of coolant accident (LOCA) in accordance with 10CTR50 and Appendix K are discussed in Reference 1. D e models are identified as IAMB, SCAT, SAFI, REFLOOD, and CHASTE. 'The 1AMS Code calculates the short ter: blevdown response and core flov, which are input i=to the SCAT code to calculate blevdown beat transfer coef ficients.

The SAFE cede is used to ceter_ine longer ter: syste= response and flows fro = the various ECC syste=s. Where appropriate, the output of SAFE is used in the REFLOOD code to calculate liquid levels. De results of these codes are used in the CEASTE code to calculate fuel clad te=peratures and maxix.:= average planar linear heat generation rates (MAFLHOR) for each fuel type.

The significant plant input parameters are given in Reference 2.

MAPLEGR's for the present fuel types were calculated The curves by in theFigures above procedure and are included in Reference 3.

3.11-1 through 3.11-5 were generated by multiplying the values in These multipliers Reference 3 by factors given in Reference 4.

vere developed assu=ing no core spray heat transfer credit in the LOCA analysis.

General Electric is performing analyses for single loop operation of Preliminary evaluation of these calculations performed according PNPS-1. indicates that a multiplier. of to the procudure outlined in NEDO-20566-2 0.83 should be applied to the MAPLHGR limits for single loop operation of the PNPS-1.

Further, GE has performed a large number of single loop analyses for similar plants; in no case has a multiplier of less than 0.70 been required. Because PNPS-1 does not have LPCI modification and because the limiting break is a suction lins break, the single loop MAPLHGRUntil multiplier the is expected to be significantly better than for most other BWE's. t is PNPS-1 calculations can be verified (as required by 10CFR.50 App. B) proposed that a multiplier of 0.65 be conservatively applied for single loop operation a t the PNPS-1. ,

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