ML20090H476

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Forwards Util 810512 Application for Amend to License DPR-35 Re Single Loop Operation,In Response to 831011 Request
ML20090H476
Person / Time
Site: 05000000, Pilgrim
Issue date: 10/24/1983
From: Vassallo D
Office of Nuclear Reactor Regulation
To: Doherty J
- No Known Affiliation
Shared Package
ML17139C189 List:
References
FOIA-84-105 NUDOCS 8311080317
Download: ML20090H476 (1)
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'e UNITED STATES 5. b' 'kI g NUCLEAR REGULATORY COMMISSION e

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October 24, 1983 Mr. Juim F. 00hercy

,318 Sumit Avenue Apt. r3 Brighton, Massachusetts 02135

Dear Mr. Doherty:

Re:

Pilgrim Station, Application for Amendment dated May 12, 1981 By letter dated October 11,'1983 you requested a copy of the Application for Amendment to DPR-35, Pilgrim Station dated May 12, 1981 which was mentioned in the Federal Register of August 31, 1983 at page 39,538.

Enclosed is the application you requested.

Sincerely, 8'

Domenic B. Vassallo, Chief Operating Reactors Bhnch #2 Division of Licensing

Enclosure:

As stated 54-2 %

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May.12, 1981 BEco. Ltr. #81-94--~

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Proposed

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'r. Thomas A. Ippolito, Chief

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..S. Nuclear Regulatory Ccmission

!ashington, D. C.

20555 C, [gg(C License No.

OPR-35 l

Docket No.

50-293 Revised Request for Technical Specification Changes Ccncerning Sinole Looo Oceration Reference (a)

BEco letter #80-295 (J.E. Howard) to NRC (T.A. Ippolito)

" Proposed Technical Specification Change Concerning Single Loop Operation", dated November 21, 1980 Jear Sir:

I acility Operating License No. DPR-35 for P'ilgrim Nuclear Power Station Unit #1 (PNPS-1) requires that the plant be shutdcwn if an idle recirculation loop can-I not be returned to service within'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Doston Edison requests that this license provision be revised to allow operation with one recirculation loop out of service.

It is also requested that this pro-posed change supercede in it's entirety, BECo. letter 480-295 dated November 21,1980,(ref. (a))which included provisions for 50". reactor power with one re-circulation loop out of service.

The loss of a single recirculation pump has occurred at several operating BWR's and is not, therefore, an improbable event. While the time required to procure necessary parts and to repair the loop depends on the nature of the failure, any loss of operating capacity would have a significant economic effect.

Modifying the Technical Snecifications as requested will reduce this potential economic i pact without reducing the safety of plant operation.

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s k!O ThN EDICON CCMPANY Mr.-Thomas A. Ippolito, Chief May 12,1981 Page Two The evaluation of this proposed mode of operation provided to Boston Edison by GE and described in Attachment (C), supports the conclusion that this mode of oper-atinn will nnt reduca cafaty marnins.

Schedule of Chance.

This change will be put into effejt upon receipt of approval frcm the Ccmission.

Fee Consideration No fee is proposed, as this change supercedes the Reference (a) letter which included payment pursuant to 10 CFR 170.12 and is still under review by your office.

Should you have any questions on this subject, please do not hesitate to contact us.

Very truly yours, JphtnY.

W./

Attachments (A) - Modified Rod Block Equation (B) - License & Technical Specification Changes (C) - Pilgrim Nuclear Power Station Single Loop Operation, NED0-24268, June 1980 with Errata and Addenda Sheet No. 1, Sept. 1980 3 signed originals and 37 copies

' Commonwealth of Massachusetts)

County of Suffolk

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Then personally appeared before me J. Edward Howard, who, being-duly sworn, did. state that he is Vice President - Nuclear of Boston Edison. Company, the applicant herein, and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf 'of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.

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Attachment A-Modified Rod Block Ecuation i

This' attachment describes the procedure for modifying the rod block equation for s

one-pump operation.

a.

The two-;;=r rcA block equerian in the existing Technical Specification is of the form:

.RB =-(mW + K)%

(1)

I where RB = power at' rod block in %

m = flow' reference slope for the rod block monitor (RBM)

W = drive flow in % of rated' K = power at rod block in " when W = 0.

For the case of top level rod block at 100* flow, denoted RB I

100 RB

= m(100) + K 100 or K = R3

- m(100) 100 Substituting for K in Equation (1), the two ' pump equation becomc.s:

R3=mW+(R3

-m(100]

(2) 100 b.

Next, the core-flow -(F )' versus drive flow (W) curves are determined for the two-pump and one-pump cases.

For the two-pump case the core flow and drive flow are derived by measuring the differential pressures in the jet pumps and recirculation loop, respectively.

Core flow for one pump operation must be corrected for the backflow.through the inactive jet pumps thus:

Actual core flow (one pump) = Active -jet pump flow - inactive jet pump flow.

Both c.he active and inactive flows are derived from the jet pump differential pressutss.- The drive flow is derived from the differ-ential pressure measurement in the active recirculation loop.

These-two curves are plotted in Figure 1.

The maximum difference between the one-pump and two-pump core flow is-' determined graphically.

This occurs at about 35% drive flow which is denoted W.

page 1 of 3~

c..Next, a horizontal line is drawn from the 35% drive flow point on the one l

pump curve to the two pump curve and the corresponding flow, W, is 2

W' i

. determined.

Thus, AW = W g

2 The rod block equation corrected for one pump flow is:

bR

- m(100b - ARB u3 = -tJ +

[ tuu j

.h re A R3 = R3 - R3 ~ #Aw 1

2 R3 = mW + R3

- m(100 + AW)

(3) 100 d.

The constants from the Technical Specification are:

= = 0.65 RB

= 107 100 From Figure 2:

W = 35 - 3d = 5 AW=W g

2 Evaluating in Equation (3),the one-pump rod block equation becomes:

g R3 = 0.65W + 107 - 0.65(100+3) = 0.65W + 38.7 (4)

This line is depicted in Figure 1 as the corrected rod, block line for one-pump operation.

APP.M TRIP SETTINC The A? W. trip settings.are flow biased in the same manner as the rod block monitor trip setting.

Therefore, the APdM rod block and scram trip settings are subject to the same procedural changes as the rod bicek monitor trip setting discussed above.-

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ATTACIO!ENT 3 License and Technical Specification Changes Lic.

Page 3 T.S.

Pagee 6 7

8

.9 11 12 13 14 15 18 20 21 27 54 127 127A 127A-1 205A 205B 205C 205C-2 205C-3 205C-4 205C-5 205C-6 205D 205E-1 205E-2 205E-3 205E-4 205E-5 sqAr,-

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3-3.

The Technical Specifications contained in Appendices A and 3, as revised through Amendment No. 48 are hereby incorporated in the license.

The

'.icensee shall operate the facility in accordance with the Technical Specifications.

C.

Records Boston Edison shall keep facility operating records in accordance with

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co,..f raraa t= n f rha Tachnical Specifications.

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

E.

(Deleted)

F.

Fire Protection i

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 Safety Evaluation (SE), dated December 21, 1978 for the facility.

Th es e 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 licansee shall sub-mit a report, explaining the circumstances, together with a revised schedule.

The licensee is required to implecent the administrative controls identified in Section 6 of the SE.

The adminLstrative controls shall be in effect by December 31, 1978.

G.

Physical Protection The licensee shall fully implement and maintain in ef fect all provisions of i

the following Commission approved documents, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p):

(1)

" Security Plan for the Pilgrim Nuclear Power Station", dated Novmeber 7, 1977 wich Revision 2 dated May 26, 1978 and Revision 3 dated Januar S,

1979.

(2)

" Pilgrim Nuclesr Power Station Safeguards Contingency Plan", dated April 5,1979 and revised by letters dated December 20, 1979 and April 22, 1980, submitted pursuant to 10 CFR 73.40.

The Contingency, Plan shall be fully implemented, in'accordance with 10 CFR 73.40(b), within 30 days of the approval by the Commission.

I (3)

" Pilgrim Nuclear Power Station Guard Training and Qualificatian Plan",

l Revision 3, dated October 1980 includes pages dated August 18, 19 79, May 28,19EO, and October 1,1980. This Plan shall be followed in accordance with 10 CFR 73.55(b)(4), 60 days after approval by the Commis-sion.

All security personnel, as required in the above plans, shall be l

qualified within two years of this approval. ~The licensee may mak2 changee i

to this plan without prior Commission approval if the changes do not decrease the safeguards ef fectiveness of the plan.

The licensee shall maintain records of and submit re@Srts concerning such changes in the

1. k SA. MT LDf1T 2/1 1$MITIhG SAFETY SYSTDi SETTIy; 1.1 FUII CIADDING INTEGRITT 2.1 FUEL CLADDING INTIORITT fcelicabilitv:

A elicability:

Applies to the interrelated Applies to trip settings of the variables associated with fuel i=stru=ents and devices which are thernal behavior.

provided to prevent the reactor syste= cafety id s ts from being exceeded.

Ob3ective:

_O_ bjective:

To establish 1'M ts below which To define the level of the process the integrity of the fuel variables at which automatic pro-cladding is preserved.

tactive action is ""tiated to prevent the fuel cladding integrity safety limits frem being exceeded.

Soecification:

Soecification:

A.

Raaeter Pressure > 800 usia and A.

Neutron 71cx Scram C. re Flow >10% of Rated The existance of a nininum critical ne li=iting safety system trip power ratio (MCPR) less than 1.07 settings shall be as specified fo:: two recirculation loop opera belo.:

tion (1.08 for single-loop opera-tion) shall constitute violation of 1.

Nuetron 71ux Trio Settines the fuel cladding integrity safety 11 nit.

This value of the MCPR is hereinafter referred to as the Safety Linic MCPR.

a.

A m Flux Scram Tris 3.

Core Ther=al 7ever I.init (Reactor Setting (Run Mode)

Pressure 1300 esta and/or core Flev s10")

When the Mode Switch is in the RUN positien, the

'a~ces the reactor pressure is 1800

.APRM fluz scra= trip psia or core flow is less than or setting shall be:

equal to 10% of rated, the steady S4.65W + 55% 2 loco state core thernal power shall not f2.65W + 51.7% 1 loco ertud 25% of design ther=al power.

Where:

C.

?over Tre.nsient S = Setting in percent h e safety linit shall be assumed of rated thermal to'be exceeded when scram is k=own power (1998 Wt),

to have been acceeplished by a

=eans other than the expected W = Percent of drive scran signal unless analyses flow to produce de=onstrate that the fuel-a rated core flow cladding integrity safety of 69 M lb/hr.

l li=Its defined in Specifi-l cations 1.1A and 1.13 vere not exceeded during the actual transient.

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'1.1 SAFETY LIMIT 2.1 I,IMITING SAFETY SYSTDi SETTING D.

k'henever the reactor is in the In the event of operation vith a cold shutdown ccedition with maximum fraction of limiting power irradiated fuel in the reactor density (MTLF3) gre.ater than the vescel, the vete.r level shall not fraction of rated power (FIP),

be less than 12 in. above the top the setting shall be modified as of the normal active fuel zone.

follows:

3ny 2 Leon S f (0.65W + 55,*. )

MFLyn Sg(0.65W+51.7%)[

FRP 1 1 teop, L

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hiere, FRP = fraction of rated thermal power (1998 F.4t)

MTLPD = maximum fraction of limiting power density where the limiting power de=sity is 13.4 r4/ft for 8x1 ar2 78x8R fuel.

The ratio of TRP to MTL?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.

For no e. bination of loop recircula-tion flow rate and core ther=al power shall the APRM flur scram trip setting be allowed to exceed 120% of rated the ;al power.

b.

APRM T1ux Scram Trip Settine (Refuel or Start and Hot Standby Mode)

When the reactor mode switch is in the EZIU"L or STARIUP position, the A?RM scram shall be set at less than or equa:. to 15% of rated power.

c.

IRM The IRM flux scram setting shall be 1120/125 ef scale.

3.

APRM Rod Block Trip Setting The APRM rod block trip s.cting shall i

be:

8K3 6. O.65W + 42:

2 toon S

7 RB f 0.65W + 38.7% 1 Loop

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._ SATETY LIMIT 2.1 t.IMITING SAFETY SYSTDi SETTINO

htere, IR3 = Rod block setting in percent of rated ther=al power (1998 W t)

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

In the event of operating with a maxima fraction 161 ting power density (M?L7D) greater than the fraction of rated power (TRP), the setting shall be modified as fol1<ris:

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RB 4.(0.65 W + 42%)

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. MFTJD j R3 f (0.65 U + 38.7% ~ FRP 1 Loop _

MFLPD

htere, TRP = fraction of rated thermalpcuer 1

MTL?D= = art =um fraction of limiting power density where the limiti=5 power density is 13.4 r4/f t for, 8x8 and ?8xSR fuel.

Thu ratio of 7RP to MTLPD shall be set equal to 1.0 unless the actual operati=g 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 be2 9 in. on level I

instrunents.

D.

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

1 E.

Turbine control valve fast clorure setting shall be 2.150 psis coe.-

trol oil pressure at acceleration relay.

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Condenser lov vacuum scram setting shall be 2 23 in. Eg. vacuum.

Main steam isolation scram setting G.:

j shall be i 10 percent valve clo-I sure.

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. BASES _:

4 1.1 FUEL CLADDING INTEGRITY A.

Fuel Cladding Integrity Limit at Reactor Pressure 2 800 psia j

and Core Flow 2 10% of Rated The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage are not directly observable during reactor operation the thermal and hydraulic conditions resulting in a departure trom nucleate bullius hrvs been u cd tr rark cha Although it beginning of the region where fuel damage could occur.

is recognized that a departure from nucleat: 'seilin; vould not necessarily result in damage to BWR fuel rods, the critical power at which boiling-transition is calculated to occur has been adopted as However, the uncertainties in monitoring the a convenient limit.

core opertting state and in the procedure used to calculate the critical power result in an uncertainty in the value o~f the critical power.

Therefore the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which = ore than 99.9% of t.he fuel rods in the core are expected to avoid boiling tran-l sition considering the power distribution within the core and all un-certainties.

l The Safety Limit MCPR is generically determined in Reference 1 (page 13) for two recirculation loop operation.

This safety limit MCPR is increased by 0.01 for single loop operation as discussed in Reference 2 (page 13).

3.

Core Thermal Power Limit (Reactor Pressure <800 psig or Core Floe <10%

of Rated)

Since the pressure drop in the bypass region is essentially all elevation i

head which is 4.56 psi the core pressure drop at low power and all flows will always be greater than 4.56 psi.

Analyses show that with a flow of 28x103 lbs/hr bundle flow,. bundle pressure, drop is nearly independent of bundle power.and has a value of 3.5 psi.

Thus, the bundle flow with a 3 lbs/hr irrespective of 4.56 psi driving head will be greater than 28x10 l

total core flow and independent of bundle power for the range of bundle i

powers of concern.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assemoly critical power at this flow is approximately 3.35 MWt. With the design peaking factors the j

3.35 MWt bundle power corresponds to a core thermal power of more than 507..

Therefore,a core thermal power limit of 25% for reactor pressures below 800 psia, or core flow less than 10% is conservative.

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Fever Transient Plant safe:y analyses have shown that the scra=s caused by ex-ceeding any safety setting vill assure that the Saf ety Li=1: of Specification 1.1A or 1.13 vill not be exceeded.

Scram ti=es are checked periodically to assure the insertion thes are ad equat e.

The thersal pcver transient resulting when a scram is accomplished other than by the expected -scram-signal (e.g.,

scram from neutron flux following closures of the main turbine s:cp valves) does not necessarily cause fuel da= age.

Eevever,

~4 thi: :p::ift:ntien e ?=r.ey T.131: violation vill be assu=ed Gen a seras is nly accomplished by means of a backup fea:ure e

the plan: des p.

The concept of not approaching a Safety L...1: provided scram signals are operable 's supported by cne extensive plant safety analysis.

The computer-provided with Pilgria Uni: 1 has a sequence annunciation progra which will i=dicate the sequence in which events such as scran, AP3M trip initi.ation, pressure scram ini 1ation, etc. occur. This program also indicates when the e

scram serpoint is cleared. This will provide infor:ation on how long a ceram cendition exists and thus provide some measure of :he energy added during a transient.

D.

Reactor Vater tevel (Shutdove condition)

During periods when the reae:or is shutdown, censideracion cus:

also be given to water level require =ents due to the effect of decay heat. If reactor va:er level should drop below the l

=p of the active fuel during this ti=a, the abili:7 to cool the core is reduced. This redue:1on in core cooling capabili:7 c=uld lead to elevated cladding c peratures and clad perforacica.

The core can be cooled sufficiently should the vacer level be reduced to two-thirds the core height.

Establishment of the safecy 12 inches above.he top of the fuel provides adequate 11=1: a:

=argin.

his level vill be continuously monitored.

1 Rnferences 1.

" General Electrie. Boiling 'Jacer Reactor Generic Reloali Fuel Application", 'iEDE-2W11-P - A-1, July 1979.

2.

"P11 grin Nuclear Power Station Single-Loop Operation", NEDO-24268, June 1980.

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3ASES:

1.D ITING SAFETY S*.*STDi SETTI!CS RELATED TO FUEL CLADDING IN TEGRITY 2.1 FUIL CLA.DDING INTEGRITY The abnormal operational transients applicable to operation of the PNPS 1 Unic have been analy:ed throughout the spectrum of planned operating con-dicions up to the thermal power conditien of 1998 Wt.

The analyses were based upcn plant operation in accordance with the operating map given in 7f !nra '4.7-1 of the FSAR.

In addition.1998 We is the licensed maximum pcwer level of PNPS 1, and this represents the maximum steady-state power

Sich shall not knowingly be exceeded.

Transient analyses performed each reload a're given in Reference 1 (page 20).

Models and model conservatisms are also described in this reference.

As discussed in Reference 2 (page 20), the core wide transient analyses for one recirulation tump operation is conservatively bounded by two-loop ?peration analyses and the flow-dependent rod block and scram setpoint equat:.ons are adjusted for one-pump operation.

Steady-state operation without forced tecirculation will not be permitted, except during startup testing.

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2. L BASES:'

I The basas for individual set points are discussed below:

Neutran F1"v Ser== Trip Ser::nga f

APRM The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads i

in percent of design power (1998 MWt).

Because fission chambers provide the basic input signals, the APRM, system responds directly to average neutron flux.

During transients, the instantaneous rate of heat transfer from the fuel (reactor charmal power) is less than the instantaneous neutron flux due to the time constant of the fuel.

Therefore, during abnormal operational transients, the thermal power i

of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses demonstrated that with a 120 percent scram trip setting, none of the abnormal operational transients j'

analyzed violate the fuel safety limit and there is a substantial margin from fuel damage.

Therefore, the use of flow referenced scram trip provides even additional margin.

The flow biased scram plotted on Figure 2.1.1 is based on recirculation i

i loop flow.

Figure 2.1.3, which shows the flow biased scram as a function of core flow, has also been included.

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l An increase in the APRM scram setting would decrease the margin present before the fuel cladding integrity safety limit is reached.

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The APRM scram setting was determined by an analysis of margins required ec provide a reasonable range for maneuvering during i

operation.

Reducing this operating margin would increase the I

frequency of spurious scrams, which have an adverse effect on reactor r,afety because of the resulting thermal seresses.

Thus, j

the APRM setting was selected because it provides adequate margin for the fuel cladding integrity safety limit yet allows operating t

margin that reduces the possibility of unnecessary scrams.

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10 20 30 40 50 60 70 80 50 100 110 120 0 --

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CORE COOLANT FLOW RATE (5 0F DESIGN) l

.#RM FLOW BIAS SCRAM VERSUS REACTOR CORE FLOW i

FIG. 2.1.3 Figure 2.1.3 above represents the APRM tvo loop flow bias scram with neutron flux plotted against core coolant flow race instead of recirculation loop flow as shown in Figure 2.1.1.

--N -

O

-r 1 BASTS _:

setting.

The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-ccre L?M systen. As with the A?M scram trip setting, the AyM rod block trip setting is adjusted downward if the aximum fraction of limiting power density exceeds the fraction of rated pcver, thus preserving the A?M rod block safety margin. Definition of sinale loco setpoints is given in Reference 3 (page 20).

~

C.

Rasetor Water M v Lcvt1 Scram Trip Setting (LLL) ne set point for lov leveJ. scram is above the bottom of the separator sk.irt.

This level has been used in transient analyses dealing with coolant inventory decrease. De results show that scran at this level adequately protects the fuel and the pressure barrier, because MCPR rs=ains well above the safety limit hCPR in all cases, and system ne scram' setting pressure does not reach the safety valve settings.

is approximately 25 in. below the normal operating range and is thus adequate to avoid spurious scrams.

D.

Turbine Seco Valve Closure Scram Trio Setting The turbine stop valve closure scram anticipates the prassure, neutron fluz and heat flux increase that could result frem rapid closure of the turbine stop valves.

With a scram trip setting of $ 10 percen't of valve closure from full open, the resultant increase in surface heat flux is li ited such that MCPR re=ains above he safety limit MCPR even during the vorst case transient that assumes the turbine bypass is closed.

Turbine Control Valve Tast Closure Sers= "rio Setting I.

che turbine control valve fast closure scram aqticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves.

n e resceor protection system initiates a scra= when fast closure of the control valves is initiated by the acceleration relay.

Bis setting and th,s fact that control valve closure time is approximately twice as long as that for the stop valves means that reaulting transients, while si=ilar, are less avere than for stop valve closure. MCP1 remains above the safety limit MCPR.

y.

Main Condenser lov Vacuum Scram Trip Setting To protect the main condenser against everpressure, a loss of condenser vacuum initistas automatic closure of the turbine stop valves and turbine To anticipate the transient and autonatic scram resulting, bypass valves.

f ro= the closure of the turbine stop valves, low condenser vacuum ne low vacuun scram set point it selected to initiate

)

initi. aces a scram.

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

\\

\\

a 18 l

.e

2.1 BASES

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

References 1.

"Supplerental Peload Licensing Submittal for Pilgrim Nuclear Power Station, i.", ygp2i. qw, a-i-a

' ~.

Tin:1 fafety Ant 1r**.s Report for Pilgrim Nuclear Power Station L* nit ill.

Attachment A, ' Modified Rod Block Equation" to 3Eco letter (J. E. Howard) 3.

to NRC (T. A. Ippelito) dated May 12., 1981.

I o

e e

9 e

o e

e e

s a iw.M a Operablo Icot.

Mutt 33 Opserblo Channelo per Trip Trip Fznction Trip Leval S:tting Ritt01 (7)

Jtcrtup/II t Ruf~~ Action (1)

(1) Sy tem S to idby 1

Hode Switch in Shutdown X

X X

h.

1 Hanual Scram X

X X

A IRH 3

Iligh Flux 1120/125 of full scale X

X (5)

A 3

Inoperative X

X (5)

A APRH 2

Iligh Flux (14) (15)

(17)

(17)

X A or B 2

Incperative X

X(9)

X A or 2

Downscal i 2 2.5 Indicated on Scale (11)

(11)

X(12)

A cr 4 Iligh Flux (15%)

515% of Design Power X

X (16)

A or B 2

2 liigh Reactor Pressure

$ 1085 polg X(10)

X X

A 2

Iligh Drywell Pressure

$ 2.5 psig X(8)

X(8)

X A

2 Reactor Lou Water Level 29 In. Indicated Level X

X X

A liigh Water Level in Scram 2

139 Callons X(2)

X X

A Discharge Tank i

2 Turbine Condenser Low Vacuum 223 In. lig Vacuum X(3)

X(3)

X A or C 2

l Main Steam Line liigh 17X llormal Full Power Radiation

Background

X

).

X A ot J 4

Hain Steam Line Isolation Valve Closure 110% Valve Closure X(3) (6)

X(3) (6)

X(6)

A or C 2

Turb. Cont. Valve Fast 1150 psig Control 011 Closure Pressure at Acceleration Relay X(4)

)(4),

X(4)

A or D 4

Turbine Stop Valve Closure

$10% Valve Closure X(4)

)(4)

X(4)

A or D

  • APluthighfluxacramnetpoint$(.65W+55){MFLPD FRP -

Two recirc. pump operation

[

I e ( 69w o s i W) { Fit P },, g, g,

,gp,g,ggg, ar

l PtiPS i

TABLE 3.2.C

.,[,

It4STHlillEtITATIC;I 11 TAT lillTIATES ROD BLOCKS Hininsusf of Operable Instrtement Trf.p Level Setting Instrument

' FRP

~

Channola Per Trip Systems (1)_

Two Loog (0.65W + 42)

HFLPD J,

APR11 Upscale (Flow One Inop_ (0.65W + 38.7) r vRe 2

Blased)

{HFLPD 2.1 indicated on scale APlut inwnscale 2

(2)

Two Loott (0.65t1 + 42)

H D

Rod Block Honitor (Flow Biayed) yp."LPD _

I 1 (7)

One Loop _(0.65W i 38.7) 5/125 of full scale Rod Block Honitor 1 (7)

Downscale 5/125 of full scale IRH Downscale (3) 3

  • c (8)

IRH Detector not in 3

Startup Position

<108/125 of full scale 1RH Upscale 3

l4)

SRH Detector not in 2 (5)

Startup Position

.,10 counts /sec.

SRH Upscale 2 (5) (6)

  • LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.D Safety and Relief Valves (Cont'd) 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 Pu=ps 1.

Whenever the reactor is in the Whenever there is recirculation flow startup or run = odes, all jet vich the reactor in the starcup or pu=ps shall be operable.

If it is run = odes, jet pu=p operability shall deter =ined that a jet pu=p is be checked daily by verifying that inoperable, an orderly shutdov=

the folleving conditions do not oc-shall be initiated and the reactor cur si=ultaneously:

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 flov i= balance of 15%

or =;re when the pu=ps are operated at the sa=e speed.

2.

The indicated value of core flow rate varies f:o= the value derived fro = loop flev

=casure=ents by = ore than 10%.

3.

The diffuser to lover plenu=

differential pressure t2ading on an individual jet pu=p varia.s fro = established ' jet pu=p P characteristics by

= ore 'than 10%.

F.

Jet Pu=? FJev Mis =atch F.

Jet Pu=p Flow Mis =atch 1.

Whenever both recirculation pu=ps Recirculation pu=p speeds shall be are in opera: ion, pu=p speeds shall checked and logged at least once be =aintained 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 ec al to 80%.

2.

Single loop reactor operation is not permittea for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the following designated ad-justments are made for APRM rod block and scram setpoints (Tech.

Spec. 2.1.A and B) RSM setpoint (Table 3.2.C), MCPR fuel cladding integrity safety limit and operating..

limits (Tech. Spec. 1.1.A and 3.11.C, respectively), and MAPLHGR (Tech Spec. 3.11.A).

129

LIMITING CONDITIONS FOR OPERATION SURUEILLANCE REOUlxDfENT G.

Structural Integrity G.

Structural Integrity 1.

The structural 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 i

level required by the ASME results obtained from compliance Soiler and Pressure Vessel with this specification will be Code,Section XI, " Rules evaluated after 5 years and the frr I~ier fica Inspectina M ennetiininna of this evaluation Nuclear Power Plant will be reviewed with AEC.

Components." 1974 O

h e

i 127A

1 5.C St-ueturst 1;, ent- / (Ccn't)

Edition (ASb2 Code, S;ctica XI).

In the intsrira until tha nucicar system piping inspection evaluarton icyc1 criteria of the ASE 1

Doiler and Pressure Vessel Code,Section XI,1974 Edition, are completed, the applicabic evaluation level provisions of the ATE toiler and Pressure Vessel Code,Section XI, 1971 Sirmser Addenda shall be used in the Insewico Inspection of nucicar piping.

Components of the primary systen boundary whose in-scrvice.cx:mina-tion. reveals the absence of fisw indications not in excecs of the allowable indication standards of this code are acceptablo Joe continued service.

Phnt operation with components which have in-::rrice exanin.:ti:n fle.: indi:sti:n(:) in :n:c:: of the allo able indicatien standards of the code shall be subject to EC spproval.

a.

Components whose in-service examination reveals flaw

. indication (s)in excess of the allowable indication standards of the ASE Code,Section XI, are unsecep-c.61e for continued service unicss the fo11cwing requirements are met:

(i)

A. analysis and evaluation of the detected flaw

~

indication (s) shall be submitted to the !GC that de=enstrate that the cenponent structural intea,rity justifies continued service. The analysis and evaluation shall follow the procedures eutlined in Appendix A, 'Tvaluation of Fisu indications,"

of ASE Ccde, Section."I.

(ii)

Prior to the resu=ption 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 co=ponent be repsired or replaced.

b.

For conponents approved for continued service in accordance with paragraph s, above, seexamination of the arcs containing the fisu indication (s) shall be conducted during each scheduled successive in-se:vice inspection. An analysis and evaluation shall be submitted to the ?GC following each in-service inspection.

The analysis and evaluation shall follow the procedures outlined in Appendix A, 'tvaluatica of Fisw Indications," of ASE Codn,Section XI, and shall reference prior analyscu subraitted to the IGC to the extent applicable.

Prior to resumption of service follouing cach in-service inspection, the EC 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.

Repair or replace =cnt of components, including re-examinations, c.

shall conform with the requirements of the ASE Code,Section XI.

In the esse of repairs, flaws shall 1e either emoved or repaired to the extent necessary to meet the allsusbic indication standards specified in ASE Code,Section XI.

e t

127A-1

.~_. _ _ _ _ _

  • SURVEILLANCE REQUIRDENTS LD41 TING CONDITIONS FOR OPDATIOR 4.11 RIACTOR 7UEL ASSDGLT f

3.11 REACTOR FUEL ASSDGLT App 11eab111ti Aeplie ab111ty The surveillance Requirements the Limiting Conditions for Operation apply to the parameters which I

sesociated with the fuel rods apply the fuel rod operating condi-te.those parameters which monitor the s

ti:r.s.

fuel rod operating conditions.

Objective _.

t Objective.

The Objective of the Surveil-The Objective of the Limiting Condi-lance Requirements is to tions for Operation is to assure ti.e specify the type and frequency performance of the fusi rods.

of surveillance to be applied to the fuel rods.

i Specifications _

Specifications _

f Average 71anar Linear Heat A.

A.

Averate Planar Linear Heat Generation Rate (APGGR)

Generation Rate (A?LMGR) i The APLBCE for each type of During power operation with both.

fuel as a function of average recirculation pumps operating, the planar arposure shall be APuGR for each type of fusi as a determined daily during 1

function of svarage planar exposure reactor operation at j,> 25 :

shall not exceed the applicable rated thammi power.

l limiting value shown in Figures j

3.11-1 through 3.11-5.

The' top curves are applicable for core flow Areater than or equal to 90% of rated core flow.

When core flow is less than 90% of rated core flow, the lower curves shall be limiting.

For i

' greater than.24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operation wtth one l

recirculation pump, values from these curves are to be multiplied by 0.84 for 8x8 and 8x8R fuel.

If at any time during operation it is detemined by nomal surveillance that the limiting value for APLHGR is being exceeded.

action shall te initiated within 15 minutes to restore operation to with-in the prescribed limits.

If the APLHGR is not returned 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 correspond-ing cction shall continue until re-actor operation is within the pre-scribed'11mits.

w

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,_.-~-..__---..,m

, _ _ ~. - -

_-.p-v r_-__-

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-,,_www,-

LDtITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRDiEhIS C.

Minimum Critical Pover Ratio (MCPR)

C.

Miniman Critical Power Ratio O!cPR)

I During power operation MCPR for two MCPR shall be determined daily recirculation loop operation shall during reactor pcwer operation at

> 25lll rated thernal power and be A 1.35 for 8x8 and P8x8R fuel. If following any change in power l

at any time during operation it is level or distribution chat wculd detercined by nornal surveillance cause operation with a limiting that the limiting value for MCPR is control rod pattern as described being ac:.eded, etion th:11 tr in the bases ior Specification initiated within 15 minutes to re-3' '3*0*

store operation to within the prescribed limits.

If the steady state MCPR is not returned to with-in the prescribed limits within two 12) hocrs, 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 action shall continue until reactor operation is within the prescribed limits.

Tor core ilovs other than. rated the MC73, shall be.2 1.35 for 8x8 and t

PSx31 fuel cines L, where L is as

~

~

shown in 71gure 3 11-8.

As an alternative method providing equivalent thermal-hydraulic proteo-tion at core flows other than raced, the calculated MC:F1 may be divided

.by Kg, where Kg is as shown in figure 3.11-8.

I For one recirculation loop operation, of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the MCPR limits at rated flow are 0.01 higher than the comparable ^ two-loop values.

r

  • ee En

._.~

.i

.. LAs. u.-

Averste Planar Linear Best ceneration Rate (ArucR1 3.11A D is specifications assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendiz K.

The peak cladding temperature (PCT) following a postulated heat gederation rate of all the rods of a fuel assembly at any i

niel Imatina and.is only dependent, secondarily on the rod to rod power distribution within en assembly.

Tha pc:k clad tenparature is cateulated assuning a UCR for the highest powered red which is equal to or less than sne de.,1;;n' 2C1.

This LHCR times 1.02 is used in the heat-up code along with the exposure dependent steady state gay conductance and tod-to-rod local peaking factors.

De limiting value for l

. APuCR is this LEGR of the highest powered rod divided by i

its local peaking factor.

The calculational procedure used to establish the AFLEGR limit i

l for each fuel type is based on a loss-of-coolant accident analysis.

The emergency core cooling system (Eccs) evaluation models which i

i are a= ployed 'to determine the effects of the loss of coolant 1

accident (LOCA) in accordance with 10CTR50 and Appendix K are i

The models are identified as LAM 3, l

discussed in Reference 1.

SCAT, SA72, REFLOOD, and CEASTE. 'The LAM 3 Code calculates the 4

short term blowdown response and core flow, which are input into l

the SCAT ' code to calculate blevdown heat transfer coefficients.

[

The SAFE code is used to' deter =ine longer tera system responseWhere approp and flows from the various ECC systems.

output of SATI is used in the REFLOOD code to calculate liquid The results of these codes are used in the CIASTR code l

levels.

to calculate fuel clad tamperatures and animan ave; rage planar 7

f linear heat generation rates (NAFLEGR) for each fuel type.

The significant plant input parameters are given in Reference 2.

l MA?LHCR's for the present fuel types were calculated by the above The curier in figures i

procedure and are included in Reference 3.3.11-1 through i'

These multipliers Reference 3 by factors jiven in Reference %.were devel i

t LOCA analysis.

t Reduction factors for one recirculation loop operation were derived in Reference 5.

I

~

1

a l

RUDDGS I

1.

General Electric 3'a Generic Reload Tuel Application, NEDE-24011-P.

2.

Less of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDo-21696 August 1977.

3.

" Supplemental Raioad Licensing Submittal for Pilgrim Nuclear Power l

!!:ti 5et

  • 1 R*1and 4". MED0-24224 November 1979.

4

" Supplement 1 to supplemental Reload Licensing Sub sittal for Pilgrim f

Nuclear Power Station Unit 1 Reload 4" NEDO-24224-1 March lyou.

i 5.

" Pilgrim Nuclear Power Station Single Loop Operation", NEDO 24268, i

June 1980.

I l l

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1 l

l l

i l

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=

205c-2

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' 3:

3211C MINDc:M CRinCAL Pcvtx RAno (Mc71)

Operattu timit NC7R yor any abnorr.a1 operating transtant analysis evaluation with the initial condition of the :uctor being at the steady state 14te. it is required that the resulting MC71 dess not 7.":

d: crease below the safsty Limit MC?1 at any stas curing w transient assuming instrument trip setting given in Specification

sditions 2.1.

De required operating limit MC71 at.

udy stat in specification 3.11.C was chosen conservatively at a value higher than MC71's of past analysis with the objective of establishing an operating limit MC71 which is fuel type and cycle independant.

ne difference between the speci.fied Operating Limit MC71 in Specification 3.11C and the safety Limit MC71 La Specification 1.1A definea the largest reduction in critical power ratio (C71) permit:ed during any antleipated abnormal operating transient.

To ensure that this reduction is cat exceeded, the most limiting transients are analised for each reload and fusi type (8x8 and 78x81) to determine that transient which ytalds the largest vs.lue This value, when added to the safsty Limit MC71 aust cf A C71.

he less than the =h'=u= operating limit MC71's of Specification 3.11.C.

De result of this evaluation is doessanted in the C3upplenantal Raload Licensing Submittal" for the current reload.

Medels used in the transient analyses are discussed in Raf arance 1 (page 20$D).

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4 3

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i.

WC?R LIMITS TOR CORE TT.DW CTEER '3.A.M RAnu_

ne purpose of the %g factor is to defins operating limies at other than rated flow conditions. At less than 100: flow the required McF1 is the product of the operating 11nic M.'.".'1 Specifically, the K, factor provides cud the Kg iactor.

the required thermal margin to ptotect Tagainst a flow in-De most limiting transient initiated from

. crease transient.

less than rated flow conditions is the recirculation pump speed up caused by a motor-g+.sarator speed control f ailure.

Tor operation in the automatic flow control mode, the K, factors assure that the operating limit MC71 gi en in Specificacion 3.11C vill not be violated should the most !* dt =g transient occur at d

In tbs manual flow control mode, the Kg j

1ess than rated flow.

f actors assure that the Safety Limic MC21 vill not be viciated for the same postitiated transient event.

I ne Is f actor curves shewn in figure 3.11-8(9 vere developed generically which are applicable to all EW2, Ea/3, and ER/4 f actors were derived using the flow control reactors, ne If J

line corresponding to rated thermal pcwer at rated core flow as described in Ref erence I (page 206DI, 205C-5 s

l

a The K, factors shown in 71gure 3.11-8(2) are conservative for the Pilgr_:n Unit 1 operation because the operating limit MCPR 2

giver. in Specification 3.11C is greater than the original 1.20 cperating limit MCPR used Dr the generic derivation of Q.

1.C MINIMllM CRIUCAL POWER RAno (MC?RI - SUR7EII. LANCE REQUIR.ND*T At core thern.a1 power levels less than or equal to 25%, the reactor vill be operating at mininum recirculation pu=p speed cod the moderator void content v'11 be very small. For all designated control rod patterns voich may be e=plo:"sd at this point, operating planc experience indiested that the resulting MCPR value is in excess of requirements by a considerable nar gi=.

With this low void concent, any inadvertent core ficu increase vould only place operation in a acre conservative mode relative to MC71.

During initial start-up testing of the plant, a MC2R evaluation vill be made at 25% thermal power level with ni d um recicculation m speed. The MC?R nargin vill thus be demonstrated such that fu:.ure MCPR evaluation below this pever level vill be shown to be unnecessary. De daily re-quire =ent for calculating MCPR abcve 25% rated thermal power is sufficient since power distribution shifts are very slow vtan there have not hem.significant power or control rod changes.

De requirement for calculating MCPR vhen a li=.iting control rod pattern is approached ensures char MCP3 vill be known followi=g a cha=ge in power or power shape (regardless of magnitude) that could place operation at a thernal ifsit.

,s 1

1 205c-6 e

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- - ^..

r

. s.

REFERENCES _

J General Electrie 3VR Ceneric taload 7uel Application, NDE-24011-7.

1.

Letter from J. E. Howard, Boston Edison Company to D L, Ziemann 2.

USNRC, dated October 31, 1975.

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F1GURE 3.W1

~

MAXIMUM AVERAGE PLANAR LINE AR HEAT GENERATI

~

VERSOS PLANAR AVERAGE EXPOSURE File L TYPE BD B 219 L

~

12 C C R. E 1:'L O W m

b 907c.VTED

11. 4 It.3
13. 3 11.1 11 g e,,

id C.5 g,.g to 1 10 96 3

,S/7 5.s CCRG FLOW 4. 907. Rg:.ga

)

~

I 9

8.T

'h o

5,0cc 1Qcco 15,000 209C0 25,0C0 30,000 B

l PLANAR AVERAGE EXPOSURE (MWft)

I f

  • For two recirculation loop Reduction factors for one recirculation loop were derived in I

" Pilgrim Single-Loop Operation", NEDO 24268, June 1980.

l l

205E-1

'a F IG UR E 3.11-2.

MAXIMUM AVERAGE PLANAR LINE AR HEAT gel 45! RATION RATE VERSOS PLAN AR AVERAGE EXPOSURE FUEL TYPE SDB219H

?

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?.

coesvtow j/

>, sc% RATtn g

^

it.4 si.s ei.3 3]U

~

n.0 M.

,o.s 1o.5 io.s

g

.3

'2 10 ie.c D

s5

3. r,,

s.s cou Flow 4 907, RAM 5

s.s w

r 9

u av i

l I

B O

5,000 iQOCO 15,000 20,000 25J2OO 30,000 s

\\

PLANAR AVERAGE EXPOSURE (MWgt)

  • For two recirculation loop Reduction factors for one recirculation loop werederived in "Pilgrict Single-Loop Operation", NEDO 24268, June 1980.

205E-2

c.

o F IGURE 3e11-3 hiAXIMUM AVERAGE PLANAR LINE AR HE AT GENERATION RATE VERSOS PLANAR AVERAGE-EXPOSURE

_F_u E LlYP E._8D B 2G?.

i

- coR.E Flow 3 90% # ATED 12, m

H iI L

11. 5 Il+
11. 4

,g 9y 11 7.

10.9 Q

lr h

W 10.4 30 5 tc.4

,,,4 y

10.1.

H 10 (0.8

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/oR.E Flov[4 907. R.AT23 C

9."!

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4 W

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9 7-

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o spoo 1Qooo 15000 20,C00 25,000 30,000

\\

PLANAR AVERAGE EXPOSURE (MWft) l

  • For two recirculation loop Reductica fac ors for one recirculation loop were derived in

" Pilgrim Single-1. cop Operatien", NEDO 24268, June 1980.

i

~

~~

20SE-3

F IG U R E 3.11 - -

MAXIMUM AVERAGE PLANAR LINE AR HEAT GENERATION RATE VERSOS PLANAR AVERAGE EXPOSURE

_ F.U E L TYPE PSIR32G5L 12.

CORE Flow ><9c7. RATEn

^

O 11 el o ln.0 st.a to.s C

,e,5 ta.t.

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'c 3 To.s To.2 ie.1 3o

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  • s **

CORE. FLCW t SC'/. R ATE.3

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S-O 5,0C0 1QOCO 15,000 2C,000 25,000 3C,0C0

\\

PLANAR AVERAGE EXPOSURE (MW[t) 4 i

  • For two recirculation loop Reduction factors for one recirculation loop were derived in-i

" Pilgrim Single-Loop Operation", NEDO 24268, June 1980.

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a F IG U R E 3.11-f MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RAT VERS 05

~~-

  • '. f.

PLANAR AVERAGE EXPOSURE

~

FUEL TYPE FSDR3282 Q

a ccRE Flew % S07c KATED m

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18.1 44 l1,9

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10.9 Id.S w

10. 4

[

14.3 io.3 IC.7.

IC.2 jQ tc.o 8 C.0 w

S4 3

c3gg pte w f,,co7,gg7 a

9. 4 35 6

9 s

i 1

l 1

o spoo 1Q000 15,000 20000 25000 30000 B

PLANAR AVERAGE EXPOSURE (mwd [t) g

~

+

1

  • For two recirculation loop Reduction f actors for one recirculation loop were derived in "Pilgri:n Single-1. cop Operation, NEDO 24268, June 1980 i

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