ML20096B354

Amend 75 to License DPR-16,changing App a Tech Specs to Allow Operation in Cycle 10 Following Refueling

ML20096B354
Person / Time
Site:	Oyster Creek
Issue date:	08/27/1984
From:	Paulson W Office of Nuclear Reactor Regulation
To:
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ML20096B356	List: ML20096B354 ML20096B361
References
NUDOCS 8409040113
Download: ML20096B354 (65)

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UNITED STATES EI NUCLEAR REGULATORY COMMISSION g

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WASHINGTON. D. C. 20555

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GPU NUCLEAR CORPORATION AND ';

JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 75 License No. DPR-16 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by GPU Nuclear Corporation and.

Jersey Central Power and Light Company- (the licensees) dated _

April 21, 1980 as supplemented March 9, 1981, August 31, 1982, July 22 and October 28, 1983, and May 1, 1984, complies with the standards and re as amended (the Act)quirements of the Atomic Energy Act of 1954, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, i

the provisions of the Act, and the rules and regulations of the.

Comission; C.. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

.l E.

The issuanc'e of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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A'ccordingly, the license is amended by changes to the Technical Specifications as indicated in the. attachment to this license amendment and Paragraph 2.C(2) of Provisional Operating License No. DPR-16 is hereby amended to read as follows:

(2)' Technical Specifications The. Technical Specifications contained in Appendices A and B, as revised through Amendment No. 75, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in a.ccordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION n

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't Walter A. Paulson, Acting Chief Operating Re' actors Branch #5 Division of-Licensing

Attachment:

Changes to the Technical.

Specifications Date of-Issuance: August 27, 1984 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 75 PROVISIONAL OPERATING LICENSE NO. OPR-16 OOCKET NO. 50-219 Replace the following pages of the Append.ix A Technica.1 Specifications with the enclosed pages. The revised pa'ges are' identified by the captioned amendment number and contain vertical lines indicating the area of change.

Section

. Description of Changes 1

Add page 1.0-6 2.1 Rep ~ lace entire section 2.2 Replace entire section 2.3 Replace entire section except page 2.3-5 3.1 Replace pages 3.1-11 and 12 only

.3.2 Replace entire section

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3.4 Replace entire section 3.5 Replace pages 3.5-4a through 3.5-7 only 3.10 Replace entire section 4.2 Replace entire section 4.10 Replace entire section e

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~ t-I 1.0-6 1.26 Fraction of Limiting Power Density (FLPD)

The fraction of limiting power density is the ratio of the linear heat generation rate (LHGR) existing at a given location

' to the design LHGR for that bundle type.

1.27 Maximum Fraction of Limiting Power Density (MFLPD) - The maximum fraction of limiting power density is the highest value. existing in the core of the fraction of limiting power density (FLPD).

1.28 Fraction of-Rated Power (FRP) - The fraction of rated power is.the ratio of core thermal power to rated thermal p'ower.

1.29 Top of Active Fuel ~(TAF) - 353.3 inches above vessel Zero.

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l-Amendment No. 7 5

2.1-1

-SECTION 2

. SAFETY LIMITS AND LIMITING SAFETY ' YSTEM SETTINGS S

i 2.1 SAFETY' LIMIT - FUEL CLADDING INTEGRITY Applicability:

' Applies to the interrelated variables associated with fuel thermal behavior.

Objective:

To establish limits on the important thermal hydraulic variables to assure the integrity of the fuel cladding.

Specifications:

A.

When the reactor pressure is greater tha.n or equal to 800 psia and the core flow is greater than or equal to 10% of rated, the existence of a minimum critical power ratio (MCPR) less than,1 07 shall constitute violation of the fuel cladding integrity safety limit.

B.

When the reactor pressure is less than 800 psia or the core flow is less than 10% of rated,.the core thermal power shall'not exceed 25%'of rated thermal power.

C.

In the event that reactor parameters exceed.the limiting safety system settings in specification 2.3 and a reactor scram is not. initiated by the associated protective instrumentation, the reactor shall be 4

brought to, and remain in, the cold shutdown condition until an analysis is performed to determine whether the safety limit established in specification 2.1. A and 2.1.B was exceeded.

D.

During all modes of reactor operat. ion with irradiated fuel in the reactor vessel, the water level shall not be less than 4'8" above the top of active fuel.

E.

During all modes of operatio'n except when the reactor head is off and the reactor is flooded to a level above the main steam nozzles, at least two [2]

recirculation loop su.ction valves and their associated discharge valves will be in th,e full.open position.

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Amendment No. 75 1

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4 2.1-2 Bases:

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 from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.

Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a con-venient limit.

However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore the fuel cladding integrity safety limit is defined as the critical power ratio in the limi. ting fuel assembly for which moretthan 99.9% of the-fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (1), which is a statistical model that combines all of the uncer-tainties in operating parameters and the procedures used to calculate critical power.

The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X)

Boiling Length (L), GEXL, correlation.

The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated.

Therefore, the fuel cladding integrity safety limit is protected by limiting the core thermal power.

At pressures below 800 psia, the' core elevacion pressure drop (0 power, O flow) is greater than 4.56 psi.

At low power and all flows this pressure

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Amendment No.

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2.1-3 differential is' maintained'in the-bypass region'of the core.3 'Since the pressure ~ drop in the bypass _ region is essentially all elevation head,- the. core pressure drop -

at low power and alliflows willialways be-greater.than 4.56 psi.

Analyses ~shoW that'with a-flow of 28 x 103 lbs/hrl 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'4.56 psi driving head will be greater than 28 x 103 lbs/hr irrespec-tive of total core flow and independent of. bundle power forythe. range of bundle powers of' concern.

Full scale' ATLAS test-data taken at-pressures'from 14.7 psia to 800 psia indicate that the fuel assembly crit-ical power at this. flow is'approximately 3.35 MWt. ~

With the design peaking factors this corresponds to a core. thermal power of more than 50%. _Thus, a core thermal power limit of 25% for reactor pressures below 800 psi _or core flow less than 10% is conservative.

Plant safety analyses have shown that the scram's caused by exceeding any safety setting will assure that the Safety Limit' of-Specification 2.-l.A or 2.1.B i

will not be exceeded.. Scram times are checked. period-ically_ to assure the insertion-times are adequate.

The thermal power transient r,esulting when a scram is j

accomplished other than.by.the expected scram signal-i (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage.

Specification 2.1.C. requires that appropriate analysis be performed to verify that backup protective instrumentation has prevented exceeding the fuel cladding integrity safety limit prior to resumption of power operation.

The concept of not approaching a Safety Limit provided scram signals are operable is supported by the extensive plant safety analysis.

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If reactor water level should drop below the top of i

the active fuel, the ability to cool the core is reduced.

This reductio'n in core cooling capabi-lity could lead to elevated cladding temperatures and clad perforation.

With a water level above the top of the 1

active fuel, adequate cooling is maintained and the decay heat can~ easily be accommodated.

It should be noted that'during power generation there is no clearly defined water level inside the shroud and what actual-ly exists is a mixture level.

This mixture begins t

Amendment No. 75 f

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within the active. fuel region and extends up through 1

the moisture separators.

For the purpose of this specification water level is defined to-include mix-l ture: level during power operations.

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The lowest point at which the water level can present-ly be monitored is 4'8" below the top of-active fuel.

Although the lowest reactor water level limit which ensures. adequate core cooling is the top of the a'ctive fuel,-the safety limit has been conservatively established at 4'8" above the top of active-fuel.

Specification 2.1.E assures that an adequate flow path exists from the annular space, between the pressure vessel wall and the core shroud, to the core region.

This provides for good communication.between these areas, thus assuring that reactor water level instru-ment readings are indicative of the water level in the-core region.

i REFERENCES f

(1) NEDO-24195, General Electric Reloa'd Fuel' Application for

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Oyster Creek.

i Amendment No. 75 I

.m 2.2-1

-2.2 SAFETY ~ LIMIT - REACTOR COOLANT SYSTEM PRESSURE

- Aoplicability:-

Applies to the limit on reactor coolant system

' pressure.

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' Obj ec tive : -

Preserve the integrity of the reactor coolant system.

Specification:

The. reactor coolant system pressure'shall not.

exceed 1375 psig whenever irradiated fuel is in the reactor vessel.

5:

Bases:

The reactor coolant system (l) represents an important i-barrier in the prevention of the uncontrolled release of fission products.. It is essential that the in-tegrity of this system be protected by establishing a pressure limit to be observed whenever'there is irradiated fuel in the reactor. vessel.

e The pressure safety limit of-1375 psig was derived from -

the-design presaures of the-reactor pressure vessel, 2

coolant piping, and isolation condenser.

The i

respective design pressures are 1250 psig at 575*F, 1200 psig at 570*F and 1250 psig at 575'F.

The pressure safety limit was chosen as the lower of 'the pressure transients permitted by the applicable design codes: -ASME Boiler and Pressure Vessel Code Gection I

1 for the pressure vessel, ASME Boiler and Pressure Vessel Code Section III for the isolation condenser and

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the ASA Piping Code Section B31.1 for the reactor coolant system piping.

The ASME Code permits pressure transients up to 10% over the design pressure (110% x 1250 = 1375 psig) and the ASA Code permits pressure transients up to 15% over the design pressure (115% x 1

1200 = 1380 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against' i

failure at the safety pressure limit of 1375 psig.

The vessel has been designed for a general membrane stress no greater than 20,000 psi at an ' internal pressure of

i 1250 psig and temperature of 575'F; this is more than a i

factor of 2 below the yield strength of 42,300 psi at this temperature.

At the pressure limit of 1375 psig, the general membrane stress increases to 22,000 psi, still almost a factor of 2 below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit.

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2.2-2 The normal operating pressure of the reactor coolant system is 1020 psig.

An overpressurization analysis-(2) is performed each cycle to assure the pressure safety limit is not exceeded...The reactor fuel cladding can withstand pressures up to the safety limit, 1375 psig, with~out collapsing.(3)

F inally, -

reactor system pressure is continuously monitored.in the control room during reactor operation.

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. l REFERENCES (1)

FDSAR, Volume I,Section IV.

(2)

NEDO-24195, General Electric Reload Fuel ARplication for Oyster Creek.

(3)

FDSAR, Volume I, Section III-2.3.3 f

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2.3-1 2.3:

LIMITING SAFETY SYSTEM SETTINGS Applicability:

' Applies to trip-settings on automatic.

protective devices related to variables on which safety: limits have.been.placed'.

Objective:

To provide automatic' corrective action to prevent the safety limits from being exceeded.

Specification:

Limiting safety system ~ settings shall be as.

follows:

FUNCTION LIMITING SAFETY SYSTEM SETTINGS A.

Neutron Flux, Scram-A.1 APRM When the. reactor mode switch is in the Run position, the APRM flux scram setting shall be S 6 ( (1. 34 x 10 -6 ) W + 34.0)( FRP ]

MFLPD with a maximum setpoint of 115.7% for.

core flow equal 'to 61 x 106 lb/hr and

greater, where:

S = setting in percent of rated power-W = recirculation flow (1b/hr)

FRP =

fraction of rated thermal power is the ratio of core thermal power to

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rated thermal power l

MFLPD = maximum fraction of limiting power density where the limiting power density for each bundle is the design linear heat

• generation rate for that bundle.

Amendment No. 7 5 1

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2.3-2 FUNCTION LIMITING SAFETY SYSTEM' SETTINGS The ratio of FRP/MFLPD shall be set equal I

to 1.0 unless the actual operating value is less than 1.0 in which case'the actual.

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operating value will be used.

This a'djustment may be accomplished by increasing the APRM gain and thus reducing.the flow reference APRM High Flux Scram Curve by the reciprocal of the APRM gain change.

A. 2 ~.

IRM f.38.4 percent of rated neutron flux B)

Neutron Flux, Control Rod Block The Rod Block s,etting shall be S6((1.34 x 10-6) W + 24.3] (MFL DI with a maximum setpoint of 106% for core flow equal to 61 x 106 lb/hr and

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greater.

The definition's of S, W, FRP and MFLPD used above for the APRM scram trip apply.

The ratio of FRP to MPLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be used.

l This adjustment may be accomplished by j

increasing the APRM gain and thus 4

reducing the flow referenced APRM rod

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block curve by the reciprocal of the APRM gain change.

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.'r 2.3-3 FUNCTION LIMITING SAFETY SYSTEM SETTINGS 4

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Raactor!High,

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61060 psig Pressure, Scram f

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Reactor.High Press'ure, 2 3 6 1070 psig ReliefValvey: Initiation 3 0 4 1090 psig E.

Reactor High Pressure,

• 1060 psig with ti.me delay Isolation Condenser

' '* 3 s e'c o n d s Initiation F.

Reactor High Pressure, 4 0 1212 psig Safety Valve. Initiation 4 0 1221 psig

+ 12 psi 4 0 1230 psig.

4 9 1239 psig

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G.

Low Pressure Main Steam

2. 825 pstg (initiated in IRM range 10)

Line', MSIV Closure H.

Main Steam Line Isolation 610% Valve Closure from Valye Closure Scram full.coen I.

Reactor Low-Water Level, 2 1135" above the top cf the

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Scram acti~ve' fuel as indicated under normal operating conditions J.

Reactor Low-Low Water 2 7'2" above the top of the Level, Main Steam Line active fuel as indicated

solation Valve Closure under normal operating conditions K.

Reactor Low-Low Water 1 7'2" above the top of the Level, Core Spray

. active fuel' Initiation L.

Reactor L'ou-Low Water E 7'2" above the tap of the Level, Isolation.. Con-active fuel with time denser Initiation delay 6 3 seconds i j;

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Turbine Trip, 10 percent turbine stop ij Scram

' valve (s) closure from full open N.

Generator Load Rejection, Initiate upon loss of oil Scram pressure from turbine acceleration relay O.

Recirculation Flow, Scram 3 71.4 Mlb/hr (117s of rated flow) 4

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Safety limits have;been established in. Specifications 2.1 and 2.2 to protect the integrity:of the fuel. cladding and reactor coolant. system barr.i.ers, respectively., Automatic protective devices have'been provided in the plant design for corrective actions to prevent the safety limits from being exceeded in normal. operation or operational transients caused by' reasonably expected single operator error-or equipment malfunction.

This Specification establishes the trip settings for these automatic.

j protection devices.

The Average Power Range Monitor, APRM(1),. trip setting has-been established to assure never reaching the fuel cladding integrity-safety limit.

The APRM system responds to changes in neutron flux.

However, near the rated thermal power, the APRM-is calibrated using a plant heat balance, so that the neutron flux that is sensed is read

.out as percent of the rated thermal power.

For slow maneuvers, such as those where core thermal power,-surface heat flux, and the power. transferred to the water follow -

the neutron flux, the APRM will read reactor thermal power.

For fast transients, the neutron flux will lead the power transferred from the cladding to the water due to the effect of the fuel time constant.

Therefore, when-1

'the neutron flux increases to the scram setting,-the percent increase in heat flux and power transferred to the water will be less than the percent increase in neutron flux.

t' The APRM' trip setting will be varied automatically with recirculation flow, with the trip setting at the rated l

flow of 61.0 x 106 lb/hr or greater being 115.7% of j:

rated neutron flux.

Based on a complete evaluation of the reactor dynamic performance during normal operation as well as expected maneuvers and the various mechanical 4

failures, it was concluded that sufficient protection is provided by the simple fixed scram setting (2,3).

How-ever, in response to expressed beliefs (4) that variation i

l of APRM flux scram with recirculation flow.is a prudent i j measure to ensure safe plant operati'on, the scram setting

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will be varied with recirculation flow.

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

The APRM scram trip setting was determined by an analysis of margins required to provide a 4

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reasonable range-for maneuvering'during-operation.

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Reducing this operating; margin would increase the-frequen-cy' of spurious scrams, which haye an adverse effect on-g!. j.

reactor safety because of the thsulting thermal stresses.

Thus, the APRM scram trip setting was-selected because it

'provides' adequate margin for the fuel cladding integrity safety limit and yet allows operating margin that reduces the possibility of unnecessary scrams.

The' scram trip setting must be adjusted to ensure that the

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LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power.'

The scram setting is adjusted in accordance with the formula in Specification 2.3.A, when'the MFLPD'is greater than the fraction of the rated poCer (FRP).

The adjustment'may be accomplished by

-increasing the APRM gain and 'thus reducing the flow referenced APRM High Flux Scram Curve by the reciprocal of the-APRM gain change.

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2.3-6

. Reactor power 11evel may be varied by moving control rods or.by varying the recirculation' flow rate.

The APRM Jsystem provides,a control rod. block to prevent gross rod withdrawal: at: constant recirculation flow. rate to ~ protect against-grossly exceeding the MCPR Fuel Cladding Integrity Safety Limit.

This ~ rod block trip setting, which is automatically varied with recirculation loop flow rate,

-prevents an increase-in the reactor power level to excessive values. due to - control rod withdrawal.

The flow variable trip setting provides substantial margin from fuel damage,. assuming a' steady-state operation at the trip setting, over the entire recirculation flow range.

The margin to the safety limit 1 increases as the flow decreases

'for the specified trip setting versus flow relationship.

Therefore, the worst-case MCPR,'which could occur during steady-state' operation, is at-106% of'the. rated thermal power because of the APRM rod block trip setting.

The actual power, distribution;in the core is established by

.specified control rod seq'uences and is-monitored continuously by the incore LPRM system.

As with APRM scram trip setting, the APRM rod block trip setting is adjusted' downward if the maximum fraction of limiting power-density exceeds the fraction of the rated power, thus preserving the APRM rod block safety margin. - As with the scram setting, this may be accomplishedLby adjusting the APRM gains.

The settings on the reactor high pressure scram, anticipatory scerss, reactor coolant system relief valves and isolation condenser have been established to assure never reaching'the reactor coolant system pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety 1,imit.

In addition, the APRM neutron flux scram and the turbine bypass system-also provide protection for these safety limits, e.g.,

turbine trip and loss of electrical t

load transients (5)..

In addition to preventing power i

operation above 1060 psig, the pressure scram backs up the 1-other scrams for these transients and other steam line isolation. type transients.

Actuation of-the isolation condenser during these transients removes the reactor decay heat without furtherLloss of r'eactor coolant thus

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protecting the reactor water level safety limit.

'I The reactor coolant system saf,ety valves offer yet another protective feature for the reactor coolant system pressure safety limit since these valves are sized assuming no x

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2.3-7 credit for other pressure relieving devices.

In compli-ance with Section I.of the ASME Boiler and Pressure Vessel.

Code,-'the safety valve must be set to open at a pressure-no higher than 103% of. design pressure,.and they must limit the reactor pressure to no more than 110% of design pressure. LThe safety valves *are sized according to the Code for_a condition of. turbine stop valve closure while operating at 1930 MWt, followed by [1] a delay..-of all escrams, [2] failure:of the turbine bypass valves to open, and -[3] failure of the isolation-condensers and relief valves to operate.

Under these conditions, a total of 16 safety..valveseare required.to turn the pressure transient.

The ASME-B&PV Code allows a +1% of working pressure (1250 psig) variation.in the lift point of the valves.

This variation is recognized in Specification 4.3.

The low pressure isolation of the main steam lines at 825 psig was provided to give protection-against fast reactor depressurization and the resulting rapid cooledown of the vessel. Advantage was taken of the scram feature which occurs when the main steam line isolation valves are I

closed, to provide for reactor shutdown so that high power -

l operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety. limit. Operation of the reactor at pressures lower than 825 psig requires that the reactor mode switch be in the STARTUP position and the IRMs be in the range 9, or lower, where protection of the fuel. cladding integrity safety limit is provided by the IRM high neutron flux scram.

l Thus, the combination of main steam line low pressure isolation and isolation valves closure scram assures the l

availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity L

safety limit.

In addition the isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure.

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The low water level trip setting of 11'5" above.the top of 1

the active fuel has been established to assure that the reactor is not operated at a water level below that for which the fuel cladding integrity safety limit is'appli-cable.

With the scram set at this p61nt, the generation of steam, and thus the loss of inventory, is stopped.

For-example, for a loss of feedwater flow a reac_ tor scram at the value indicated and isolation valve closure at the low-low water level set point results in more than 4 feet of water remaining above the core af ter isolation (6).

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Amendment No.,Y, 7 5

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During periods when the reactor is shut down, decay heat is present and adequate water level must be maintained to provide core cooling.

Thus,.the low-low level trip point of 7'2" above the. core is provided to actuate the core,

i spray system to provide cooling water should the level 1

drop to this point.

In addition, the normal reactor feedwater. system and control rod drive hydraylic system

' provide protection for the water level safety limit both 4

when-the reactor Lis operating at power and in the shutdown condition.

The turbine stop valve (s) scram is provided to anticipate

.the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valve (s) and failure of the turbine bypass system.

l The generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves to a load rejection and failure of the turbine bypass system.

This scram is initiated by the loss of turbine accelerr. tion relay oil pressure.

The timing for this scram is almost identical to the turbine trip.

l The. total recirculation flow scram is provided to terminate g

a flow increase transient. Flow transients are normally

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protected against by employing the kf factor and using mechanical stops on the recirculation pumps. Oyster Creek does not have mechanical stops on its recirculation pumps and maximum flow is beyond the limit for which the kf factor provides protection. The recirculation flow scram is set to the maximum flow level corresponding to the kf curve to be

.used (Section 3.10).

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References (1)

FDSAR, Volume I, Section VII-4.2.4.2

-(2)

FDSAR, Amendment 28, Item III.A-12 (3)

FDSAR, Amendment 32, Question 13

-(4)

Letters, Peter A. Morris, Director, Division of Reactor Licensing, USAEC to John E.

Logan, Vice President, Jersey' Central Power and Light Company, dated November 22, 1967 and. January 9, 1968

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FDSAR, Amendment 65, Section B.XI.

(6)

FDSAR, Amendment 65, Section B.IX.

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Amendment No. 75 1

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4 3.1-11 TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD)'

Min. No, Min. No'. of-2 of Oper-Operable able or Instrument Reactor Modes 'in Which Operating Channels Function must be Operable (Tripped)

Per Operable' Action f.

Function Trip Setting Shutdown Refuel Startup Run Trip Sys.

Trip Systems Required

• K.

Rod Block No con-trol rod

1. SRM Upscale f.5 x 105 cps

.X X ( 1) 1 2

withdrawals

2. SRM Downscale 6 100 cps (f) permitted X

X (1) 1 2

3.

IRM Downscale i S/125 fullscale(g)

X X

2 3

4. APRM Upscale X (s)

X X

2 3 (c)

5. APRM Downscale f. 2/150 fullscale X

2 3 (c)

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6'. IRM Upscale d'. 108/125 fu11 scale X

X 2

3 7.

Scram Discharge 18g$t11ons X(z)

X(z)

X(z) 1-1 a.

Water Level High Il L.

Condenser Vacuum Pump Isolation Insert con-l trol rods 1.

High radia-

< 10 x Normal During Startup and 2

2 tion in Main

Background

run when vacuum pump Steam Tunnel is operating N.

Diesel Generator Time delay after l

j Load Sequence energiz. of relay

1. Containment 40 sec +.15%

X X

X X

2(m) 1(n)

Consider Spray Pump containment sorav 1000 inoperable and comply with Spec.

3.4.C (See Note q).

l AESndment No.,f/, 7 5 j

i

.-... u _c.:_. ~. _

~

d.

.b p

q.. -

3.1.12-TABLE 3.1.1 (CONTD)

Action required when minimum conditions for operation are not satisfied..Also' permissible to trip inoperable trip system.

When necessary to conduct tests and. calibrations,.one channel-may1 J

~

be made inoperable for up to one hour per month without. tripping its. trip system.

[

See Specification 2.3 for Limiting Safety System Settings.

~i Notes:

4

-c.- Permissible to bypass, with control rod block, for reactor protection ~ system reset in. refuel i

mode.

~

b.

Permissible to bypass below 800 psia in refuel and startup modes.

c.

One (1) APRM in each operable trip system may be bypassed,or inoperable provided the require-ments of specification 3.1.C and 3.10.C are satisfied.- Two APRM's in the same quadrant shall not be concurrently bypassed except as noted below or permitted by~ note.

Any one APRM may be removed from service for up to one hour for test or calibration without inserting trips in its trip system only if the remaining operable APRM's. meet the requirements of specification 3.1.B.1 and no control rods are moved nutward during the calibration or test, j

During this short period, the requirements of specifications 3.1.B.2, 3.1.C and 3.10.C need not be met.

d.

The'IRM shall be inserted and operable until the APRM's are operable and readinh at 'least -2/150 i

full scale.

e.

Air ejector isolation valve closure time delay shall not exceed 15 minutes.

f.

Unless SRM chambers are fully inserted.

g.

Not applicable when IRM on lowest range.

i i

-h.

One instrument channel in each trip system may be inoperable provided the circuit which it operates in the trip system is placed in a simulated tripped condition.

If repairs cannot be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the reactor shall be placed in the cold shutdown condition.

If more than one instrument channel in any trip system becomes inoperable, the reactor shall be' placed.

in the cold shutdown condition.

Relief valve controllers shall not be bypassed for more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30-day period and only one c'elief valve controller i

may be bypassed at a time.

1 Am:ndment No. 7 5 j

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.._a__.._..._.

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P 3.1.12a t

This page not required, information on page 3.1-13.

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Amendment No. 7 5

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3.2-1 3.2 REACTIVITY CONTROL Applicability:

' Applies to core reactivity and the operating-status of'the reactivity control systems for the reactor.

Obiective:

To assure reactivity control capability of the reactor.

Soecification:

A.

Core Reactivity The core reactivity shall be limited such that the core.could be made subcritical at any time during the operating cycle, with the. strongest operable control rod fully withdrawn and all other operable rods fully inserted.

'B.

Control Rod System

- 1.

The control rod drive housing support shall be in' place during power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all control rods are fully inserted and Specification 3.2.A is met.

2.

The Rod Worth Minimizer (RWM) shall be operable during-each reactor startup until reactor power reaches 10% of rated power except as follows:

(a)

Should the RWM become inoperable after the first twelve rods have been withdrawn, the startup may continue provided that a second j

licensed operator verifies that the licensed operator at the reactor console is

{

following the rod program.

t (b)

Should the RWM be inoperable before a I

startup is commenced or before the first twelve rods are withdrawn, one startup during each calendar year may be performed without the RWM provided that the second licensed operator verifies that the licensed operator at the reactor con-sole is following the rod program and provided that a reactor engineer from the Core Engineering Group also verifies that Amendment No.

75

3.2-2 the rod program is being followed.

A startup without the RWM as described in this subsection shall be reported in a special report to the. Nuclear Regulatory Commission (NRC) within 30 days of the startup stating the reason for the failure of the RWM, the action taken_to repair it and the schedule for completion of the repairs.

Control rod wit $drawa'l sequences shall be,'

{

established with a banked position withdrawal secuence so that the red dree

^

accident design limit of 280 cal /g= is not exceeded.

For control rod withdrawal sequences not in strict compliance to BPWS, the maximum in sequence rod worth shall be

$1.0% 4K.

3.

The average of the ' scram insertion times of all operable control rods shall be no greater than:

Rod Length Insertion Time Inserted (Percent)

(Seconds) 5 0.375 20

~ ~

0.900 50 2.00 90 5.00 The average of the scram i'nsertion times for the three.fa'stest control rods of all groups of four contr'ol rods in a two by two ' array shall be no I

greater than:

Rod Length Insertion Time 2nserted (Percent)

(Seconds) 5 0.398 20 0.954 50 2.120 90 5.300 1

Any four rod group may contain a control rod i

which is valved out of service provided the above requirements and Specification 3.2.A are met.

.i Time zero shall be taken as the de-energization l}

of the pilot scram valve solenoids.

c t

4.

Control rods which cannot be moved with control rod drive pressure shall be considered inoper-able.

If a partially or fully withdrawn centrol red drive cannot be moved with drive or scram pressure the reactor shall be brought to a h

J f-Amendment No. 75

3.2-3 shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless inves--

tigation demonstrates that the cause of the failure is not dde to a failed control rod drive-mechanism collet housing.

Inoperable control.

rods shall.be valved out.of service, in such positions ~ that. Spec'ification 3.2. A is met.

In no case shall the number of rods valved out of

-service be greater than six during the power operation.

If this specification is not met, the reactor shall be placed in the shutdown condition.

5.

Control Rods shall not be withdrawn for approach to' criticality unless at least two source range channels have an observed count rate equal to or greater than 3 counts per second.

C.

Standby Licuid Control System 1.

The standby liquid control system shall be operable at all times when the reactor is not shutdown by the control rods such that Specification 3.2. A-is met and except as provided in Specification 3.2.C.3.

2.

The standby liquid control solution shall be maintained within the volume-concentration requirement area in Figure 3.2-1 and at a temper-ature not less than the temperature presented in

. Figure 3.2-2 at all times when the standby liquid control system is required to be operable.

~

3.

If one standby liquid control system pumping circuit becomes inoperable during the RUN mode and Specification 3.2.A is met, the reactor may remain in operation for a period not to exceed 7 days, provided the pump in the other circuit is demonstrated daily to be operable.

,y D.

Reactivity Anomalies f'

The difference between an observed and predicted con-

)'

trol rod inventory shall not exceed the equivalent of i

one percent in reactivity.

If this limit is exceeded j

and the discrepancy cannot be explained, the reactor i

shall be brought to the cold shutdown condition by normal orderly shutdown procedure.

Operation shall not be permitted until the cause has been evaluated and appropriate corrective action has been completed.

The NRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of this situation in accordance with Specification 6.6.

iit Amendment N$e 73

= - -

L

- " - - - ' =

-. m u y ;

.j 3.2-41 Bases:'

)

Limiting conditions of~ operation.on. core reactivity-and the reactivityfcontrol systems-are: required to assure that

~

the excess reactivity of theLreact.or core is controlled at o

all. times.

The conditions specified.herein~ assure-the r'

. capability to provide reactor shutdown from steady state

^

and transient conditions and assure the capability of limiting reactivity insertion rates under accident 1 conditions to values-which do 'not-~ jeopardize the reactor P

-coolant system integrity or operability-of required safety

~

features.

The core reactivity limitation is required to assure the~

reactor.can be. shut down at any time when fuel is in the core.~

It is a restriction that must be incorporated into-I the design of the. core fuel; it must be applied to the conditions resulting from core alterations; and it must be applied in determining the required operability of.the

' core reactivity control devices.

The basic criterion is

.that the core at any point in its operation be capable of t

being made suberitical in the cold, xenon-free condition with the operable control rod of highest worth fully j.

withdrawn and all other operable rods fully inserted.

At

'~

most. times'in core life more than one control rod drive could fail mechanically and this. criterion would still.be.

met.

In order to assure that the basic criterion will be satisfied an additional design margin was adopted; that the kegg~be less than 0.99 in the cold xenon-free condition with the rod of highest worth fully withdrawn and all others fully inserted.

Thus the design re-i quirement is 'kegg 40.99, whereas the minimum condition for operation is kegg 41.0 with the operable rod of highest worth fully withdrawn (1).

This limit allows control rod testing at any time in core life and assures

]

that the plant can be shut down by control rods alone.

.i,

Fuel bundles containing gad,olinia as a burnable neutron absorber results in a core reactivity characteristic which increases with exposure, goes through a max'imum and then i r, decreases.

Thus it is possible that a core could be more ij:

reactive later in the cycle than at the beginning.

i Satisfaction of the above criterion can be demonstrated

.I conveniently only at the time of refueling since it requires the core to be cold and xenon-free.

The demonstration is designed to be done at these times and is such that if it is successful, the criterion is satisfied i

l '

e Amendment No.

75 gy.

8eg[,+

d.

e d

3.2-5

,p

< for tiheientire subsequent fuel cycle.

The criterion will

.be satisfied by demonstrating Specification 4.2. A at the beginning of each. fuel cycle with the core in the cold,.

xenon-free condition.

This demonstrativ. will include consideration.for the calculated reactivity characteristic during the.following operating cycle and the uncertainty in this calculation.

The control rod drive housing support restricts the outward movement of a control. rod to.less than 3 inches in' the extremely remote event of a housing failure (2).

The j

amount 'of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage to the reactor coolant system.

The support is not required when no fuel is in the_ core since no nuclear consequences could occur in the absence of fuela The support is not' required if the reactor ecolant system is at atmospheric pressure since there would then be no driving force to rapidly eject a-drive housing.

The support is not required.if all control rods'are fully inserted since the reactor would remain suberitical even in tbg event of complete ejection of the,ctrongest control rodL33 l

The Rod Worth Minimi=er(4b provides automatic supervision of conformance to the'specified control rod.

patterns. ~ It serves as a back-up to procedural control of control rod worth.

In the event that the RWM is out of service when required, a licensed o.perator can manually

{

fulfill the control rod pattern conformance functions of the RWM in which case the normal procedural controls are

' backed'up by independent procedural controls to assure j

conformance during control rod withdrawal.

This allowance j

to perform a startup withcut the RWM is limited to once each calendar year to assure a high operability of the RWM which is preferred over procedural controls.

1.

Control rod drop accident (RDA) results for plants using banked position withdrawal sequences (BPWS) show that in i

all cases the peak fuel enthalpy in an RDA would be much I

less than the 280 cal /gm design limit even with the maximum incremental rod worth. The BPWS is developed l

prior to initial operation of the unit following any refueling outage and the requirement that thg operator follow the BPWS is supervised by the RWM or a second ji licensed operator.

If it is necessary to deviate

')

slightly from the BPWS sequence (i.e., duc co an inoperable control rod) no further analysis is needed if

+

the maximum incremental rod worth in the modified sequence is 41.0% 4K.

An incremental control rod worth of f1.0% AK will not result in a peak fuel enthalpy above the design-limit of 280 cal /gm as documented in reference 10.

I e

s Amendment No. 7 5 j

.w 3.2o6 The BPWS limits the reactivity worths of control rods and I

together with the integral rod velocity limiters and the

, action of the control rod drive system limits potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal /gm.

Method and basis for the rod drop accident analyses.are documented in Reference 5.

The control rod system is designed to. bring the reactor subcritical from a scram signal at a rate fast enough to prevent fuel damage.

Scram reactivity curve for the transient analyses is calculated and evaluated with each reload core.

In the analytical treatment of the tran-sients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods.

This is adequate and conservative when compared to the typical time delay of about 210 millisec-onds estimated from scram test results.

Approximately the first 90 milliseconds of each of these time intervals result from the sensor and circuit delays when the pilot scram solenoid de-energizes.

Approximately 120 millisec-onds later, the control rod motion is estimated to actually begin.

Howe ve r*, 200 milliseconds is conserva-tively assumed for this time interval in the transient analyses and this is also included in the allowable scram insertion times of Specification 3.2.B.3.

The specified limits provide suf ficient scram capability to accommodate failure to scram of any one operable rod.

This failure is in addition to any inoperable rods that exist in the core, provided that those inoperable rods met the core reacti-vity Specification 3.2.A.

Control rods (6) which cannot be moved with control rod drive pressure are citarly indicative of an abnormal operating condition on the affected rods and are, there-fore, considered to be inoperable.

Inoperabe rods are valved out of service to fix their position in the core and assure predictable behavior.

If the rod is fully inserted and then valved out of service, it is in a safe position of maximum contribution to shutdown reactivity.

i If it is valved out of service in a non-fully inserted positione that position is required to be consistent with the shutdown reactivity limitation stated in Specification 3.2. A, which assures the core can be shut down at all times l

t

}

with control rods.

Before rod is valved out of service in a non-fully inserted position an analysis is performed to insure specification 3.2.A is met.

1 Amendment No. 75 1

d_

3.2-7 The number of rods permitted to be valved out of service coulo.be many more than the six allowed by the specifica-tion, particularly late in the operating cycle: however, the occurrence of more than six coula be indicative of a

~

generic problem and the rea'ctor will b~e shutdown.

Also if damage within the control rod drive mechanism and in particular, cracks in drive internal housings, cannot be

~

ruled out, then a generic problem affecting a number of drives cannot be ruled out.

Circumferential cracks resulting from stress assisted intergranular corrosion have occurred in the collet housing of drives at several BWRs.

.This type of cracking could occur in a number of drives and if the cracks propagated until severance of the collet housing occurred, scram could be prevented in the affected rods.

Limiting the period of operation with a potentially severed collet housing and requiring increased surveillance after detecting one stuck rod will assure that the reactor will not be operated with a large number of rods with failed collet housings.

Placing the reactor in the shutdown condition inserts the control rods'and accomplishes the objective of the specifications on con-trol rod operability.

This operation is normally expected to be accomplished within eight hours.

The source range monitor (GRM) system (7) performs no automatic safety function.

It does provide the operator with a visual indication of neutron level which is needed for knowledgeable and efficient reactor startup at low neutron levels.

The results~of the reactivity accidents are functions of the initial neutron flux.

The require-ment of at least 3 cps assures that any transient begins

.at or above the initial value of 10-8 of rated power used in the analyses of transients from cold conditions.

One operable SRM channel would be adequate to monitor the approach to critical using homogeneous patterns of scattered control rods.

The standby liquid control system is designed to bring the reactor to a cold shutdown condition from the full power steady state operating condition at any time in core life independent of the control rod system capadilities (8),

If the reactor is shutdown by the control rod system and would be subcritical in its most reactive condition as required in Specification 3.2.A, there is no requirement for operability of this system-To bring the reactor from full power to cold shutdown, sufficient liquid control Amendment No.

75 i

5 ' -_ p p

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a.,J

~+ ;,-

a-w.

.. - ~. -.

=

i.E

[1

_ t 4

s L

'3.2-8 I

'e, l mustibe inserted' to givesa negative reactivity worth equal to1 the -combined ~ef f ects.of ' rated: coolant voids, fuel

-Doppler,1. xenon,. samarium,.and: temperature change plus shutdown' margin.

.This requires a boron concentration of sf V

600! ppm'in.the, reactor.

An additional-25% boron, which;r

.results in an average boron concentration in the reactor LofL750 ppm, is inserted toiprovide margin for. mixing:

-uncertainties inEthe reactor.

The system is required to.

' insert the solution in a. time-interval between 60-120

~

L minutes.torprovice r,or. good mixing in the reactor and to

. override the rate of reactivity insertion due.to cooldown l_

- of the reactor following the. xenon peak.

. The liquid control: tank volume-concentration requirements of Figure 3.2-1 assure 'that the above requirements for,

• 1iquid controlJinsertion are met with one 30 gpm liquid control pump.- The-point (1937 gal, 19.4% solution)

-( 9 )'

results in the required amount of solution being inserted

.into the reactor in not less than 60 minutes, and there-fore, z defines the maximum concentration-minimum volume requirement.

The point.(3737 gal,1110.3% solution) (9) results in the required amount of solution being injected l

into the reactor is -not more than 120. minutes,.and there-fore, defines the minimum concentration requirement.

The bounda,ry joining these points results in the ' required amount of solution being inserted into the reactor in the~

interval 60-120 minutes. -The maximum volume of 4213 gal is established by the tank capacity.

The tank volume' requirements include consideration for 137 gal of solution which is contained below the point where the pump takes suction from the tank and, therefore, cannot be inserted into the reactor.

The range of solution volume during normal: operation is expected to be 2387-2937 gal.

b The solution saturation temperature varies with the con-centration of' sodium pentaborate.

The solution will be maintained at least 5'F above the saturation temperature to guard against precipitation.

The 5'F margin is included in Figure 3.2-1.

Temperature and liquid level alarms, fo,e the system are annunciated in the control room.

U The acceptable time out of service for a standby liquid Lp contro1' system pumping circuit as well as other safety f'

features is determined to be 10 days.

However, the allowed time out of service for a standby liquid control system pumping circuit is conservatively set at 7 days in the specification.

Systems are designed with redundancy to increase their availability and'to provide' backup if one of the components is temporarily out of service.

i' D

]

Amendment No.

75 1

+

. sM 4 3.2-9 During eachl fuel cycle excess operating reactivity varies as. fuel depletes and as any burnable poison in supplementary control'is burned.

The magnitude of this excess-reactivity is indicated by the integrated worth of

. control. rods inser.ted into the core, referred to as the control rod inventory in the core.. As' fuel burnup

. progresses, ' anomalous behavior in the excess. reactivity may be detected by comparison of actual rod inventory with expected inventory based on appropriately corrected past data.

Experience at Oyster Cr,eek and other operating BWR's indicates that'the control rod inventory should be predictable to the equivalent of one percent in reactivity.. Deviations beyond this magnitude would not be expected and would require thorough evaluation.

One

. percent reactivity limit is considered safe since an insertion of this reactivity into the core would not lead to transients exceeding design conditions of the reactor system.

References:

(1)

FDSAR, Volume I,Section III - 5.3.1 (2)

FDSAR, Volume I, Section VI-3 (3)

FDSAR, Volume I,Section III - 5.2.1

-(4)

FDSAR, Volume I,

Section VII-9

~(5)

NIDO-24195,. General Electric Reload Fuel Application for Oyster Creek.

(6)

FDSAR, Volume I, Section III-5 and Volume II, Appendix B

.l (7)

FDSAR, Volume I, Sections VII - 4.R.2 and VII - 4.3.1 (8)

FDSAR, Volume I, Section VI-4 2

(9)

FDSAR, Amendment No. 55, Section 2 (10)

C. J. Paone, Banked Position Withdrawal Sequence, January 1977 (NEDO-21231) i i

I i

Amendment No. 75-

~

. -.,. _. _ - - _, _ _. - _ _ - - - -. ~. _,. -. - _, _,. _- -._, _,---,-,_._,, _

y,,,-

z..

a.

Pt 3.4-1 3.4 EMERGENCY COOLING Applicability:

Appli'es to the operating status of the

.. emergency cooling systems.

Obiectiver To' assure operability of*the emergency cooling systems.

Specifications:

A.

Core Spray System 1.

The core spray system shall be operable at all f

times with irradiated fuel in_the reactor vessel, except as otherwise specified in this section.

l 2.

The absorption chamber water volume shall be at least 82,000 ft.3 in order for the core spray system to be considered operable.

Ifonecorespraysystem[looporitscorespray 3.

header aP instrumentation becomes inoperable l

during the run mode, the reactor may remain in operation for a period not to exceed 7 days provided the remaining loop-has no inoperable components and is demonstrafed daily to be operable.

1 4.

If one of the redundant' active loop components in l

the core spray system becomes inoperable during the run mode, the reactor may remain in operation for a period not to exceed 15 days provided the other similar component in the loop is demon-strated daily to be operable.

If two of the redundant active loop components become inoper-able, the limits of Specification 3.4. A shall apply.

5.

During the period when one diesel is inoperable, the core spray equipment connected to the opera-ble diesel shall be operable.

l 6.

If Specifications 3.4.A.3, 3.4.A.4, and 3.4.A.5 are not met, the reactor shall be placed in the I

cold. shutdown condition.

If the core spray Amendment No. 75 g

i e

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1

< - > = -

~ -

" - ^ ^ ^

' ^ ^ ^ ^ ^ ^ ^ ~

3.4-2 system becomes inoperable, the reactor shall be placed in the cold shutdown condition and no work shall-be performed on the reactor or its con-nected systems which could result in lowering-the reactor water level to less than 4'8" above the top of the active fuel.

7.

If necessary to ' accomplish' maintenance or modifi-

~

cations to the core spray systems, their power supplies or water supplies, reduced system avail-ability is permitted when the reactor is:

(a) maintained in the cold shutdown condition or (b) in the refuel mode with the reactor coolant system maintained at less than 212'F and vented, and (c) no work is performed on the reactor vessel and connected sys.tems that could result in lowering the reactor water level to less than 4'8" above the top of the active fuel.

Reduced Core Spray System Availability is minimally defined as follows:

a.

At least one core spray pump, and system components nec.essary to deliver rated core spray to the r.eactor vessel, must remain operable to the extent that the pump and any necessary valves-can be started or operated from the contro-1-room or from local control stations.

b.

The fire protection system is operable, and c.

These systems are demonstrated to be operable on a weekly basis.

8.

If necessary to accomplish maintenance or modifi-cations to the core spray systems, their power supplies or water supplies, reduced system availability is permitted when the reactor is in the refuel mode with the reactor coolant system

~

maintained at less than 212*F or in the startup mode for the purposes of low power physics testing.

Reduced core spray system availability is defined as follows:

L a.

l At least one core spray pump in each loop, and system components necessary to deliver I

Amendment flo. 75 4

d

~

V

- _a -

n 3.4-3 rated core spray to the reactor vessel, must remain o,perable to the extent that the pump and any necessary valves in each loop can be started or operated from the control room or

-from local cont.rol s-tations.

b.

The fire protection system is operable and, c.

Each core spray pump and all components in 3.4.A.8a are demonstrated to be operable every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

9.

If Specifications 3.4.A.7 and 3.4.A.8 cannot be met, the requirements of Specification 3.4.A.6 will be met and work will be initiated to meet minimum operability requirements of 3.4.A.7 and 1

3.4.A.8.

l 10.

The core spray system is not required to be

\\

operable when the following conditions are met:

a.

The reactor mode: switch is locked in the

" refuel" or " shutdown" position.

b.

(1)

There is an operable flow path' capable of taking suction from the condensate -

storage tank' and transferring water to the reactor vessel, and

?

(2)

The fire protection system is operable.

c.

The reactor coolant system is maintained at less than 212*F and vented.

d.

At least one core spray pump, and system components necessary to deliver rated core spray flow to the reactor vessel, must i

remain operable to the extent that the pump and any necessary valves can be started or operated from-the control room or from local i

control stations, and the torus is mechanically intact.

i Amendment No.

75 l

4

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a

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nu

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~ "

3.4-4 e.

(1)

No work shall be performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel and the conden-sate storage tank level is greater than thirty.(30) feet (360,000 gallons).

At least two redundant systems including core spray pumps'and system components must remain operable as defined in d.

above.

OR (2)

The reactor vessel head, fuel pool gate, and separator-dryer pool gates are removed and the water level is above elevation 117 feet.

NOTE:

When filling the reactor cavity from the condensate storage tank and draining the reactor cavity to the con-densate stonage tank, the 30 foot limit does not app >1y provided there is suf-ficient amount of water to complete the flooding operation.

B.

Automatic DepressurizationJSystem 1.

Five electromatic relief valves of the automatic depressurization system shall be operable when the reactor water temperature is greater than 212*F and pressurized above 110 psig', except as specified in 3.4.B.2.

The automatic pressure relief function of these valves (but not the automatic depressurization function) may be inoperable or bypassed during the system hydro-static pressure test required by ASME Code Section XI,15-500 at er near the end of each ten j

year inspection interval.

2.

If at any time there are only four operable elec-tromatic relief valves, the reac' tor may remain in operation for a period not to exceed 3 days pro-vided the motor operated isolation and condensate Amendment No.

75 L_

+e

'9 7

.u

~

-s 3.4-5

makeup valves in both isolation condensers are demonstrated daily to be. operable.

i 3.-

.If Specifications 3.4.B.1 a'nd 3.4.B.2 are not

-met; reactor pressure shall be reducedsto.110 psig:oruless, within-24 hours.

4.-

The -time' delay set point for initiat' ion: af ter coincidence of low-low-low reactor water level and high drywellipressure.shall tue set.tcrexceed

'two minutes.

C.

Containment Spray System'and Emergency Service Water System 1.

The containment spray system and the emergency service water system shall.be operable at all-

~

times with irradiated fuel in the reactor vessel, except as specified'in Specifications 3.4.C.3, 3.4.C.4, 3.4.C.6 and 3.4.C.8..

2.

The absorption chamger water -volume shall not be less l than. 82,000 f t.in order for the

• ~

containment spray and emergency service water system to be considered operable.

7 3.

If one emergency service water system loop

~

becomes inoperable, its associated containment spray system loop shall'be considered inoper-able.

If one containment spray system loop and/or its associated emergency service water system loop becomes inoperable during the run mode, the reactor may remain'in operation for a period not to exceed 7 days provided.the remaining containment spray system loop and its

-r.

associated emergency service water system loop each have no inoperable components and are demonstrated daily to be operable.

i 4.

If a pump in the containment spray system or umergency service water system becomes inoper-able, the reactor may remain in operation for a

.{.

period not to exceed 15 days provided the other

~

similar pump is demonstrated daily to be oper-able.

A maximum of two pumps may be inoperable b

I t

Amendment-No.7 5

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3.4-6 provided the-two. pumps are not.in the same loop.

If more-than two pumps become inoserable, the limits of Specification 3.4.C.3 stall apply.

5.

During the period when one diese.1 is' inoperable, the. containment spray loop and emergency service water system loop connected to the operable diesel shall hay (no inoperable components.

6.

If primary containment integrity is not required (see Specification 3.5.A), the containment spray system may be made inoperable.

7.

If Specifications 3.4.C.3, 3.4.C.4, 3.4.C.5 or j

3.4.C.6.are not met, the reactor shall be placed in the cold shutdown condition.

If the contain-ment spray -system or the emergency service water system becomes inoperable, the reactor shall be placed in the cold shutdown condition and no work shall be performed on the reactor or its connec-ted systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel..

I The' containment'spriysystenmaybemadeinoper-8.

able during-the integrated primary containment leakage rate test required by Specification 4.5, provided that the reactor is maintained in the cold shutdown condit.io'n and that no work is performed on the rea~ctor or its connected systems which could result in lowering the reactor leve1 to less than 4'8" above the top of the active fuel.

D.

Control Rod Drive Hydraulic System 1.

The control rod drive (CRD) hydraulic system shall be operable when the reactor water temperature is above 21,2*F except as specified in j-3.4.D.2 below,

's 2..

If one CRD hydraulic pump becomes inoperable when the reactor water temperature is above 212*F, the reactor may remain in operation"for a period not i

to exceed 7 days provided the second CRD hydrau-j lic pump is operating and is checked at least t

once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

If this condition cannot be i

met, the reactor water temperature shall be reduced to<212'F.

Amendment Nei.

75 6

+.g n.~,_.m--nn--nwm,.-e.c

t:..

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t 3.4-7 o

E.

. Core Spray and Containment Spray Pump Compartments Doors The'coreLspray and containment spray pump compartments doors shall be closed at all times except-during-passage in order to consider the core spray system and the containment spray,5ystem. operable.

F.

Fire Protection-System L

1.

The fire protection system shall be' operable at all times with fuel in the reactor vessel except as specified in Specification 3.4.F.2.

2.

If the fire protection system becomes inoperable during the run mode, the reactor may remain in operation provided both' core spray system loops are operable with no inoperable components.

Bases:

This-specification assures that adequate emengency core cooling capability is available when the core spray system is requirede Based on the loss-of-l coolant analysis for the worst line break, a core l

~~

spray of at least 3400 gpmtis required withl;g 35 seconds to assure effective core cooling.*l J

Thus, if one loop becomes inoperable, the operable loop is -

capable of providing coo 1~ing to the core and the reactor may remain in operation for a period of 7 da'ys provided repairs can be completed within that time.

The 7 days is based upon the consideration discussed in the bases of Specification 3.2 and the pump operability tests of Specification 4.4.

If repairs cannot be made, the reactor is depressurized and vented to prevent pressure buildup and no work is allowed to be performed on the reactor which could result in lowering the water level below 4'8" above 4

the top of active fuel.

Each. core spray loop contains redundant active compo-n.e n ts.

Therefore, with the loss of one of these components the system is still capab1'e of supplying iq.

• Core Spray System 2' is required to deliver 3640 gpm.

Amendment No. 75 a

-(.

i......

x

~

3.4-8 rated flow and the system as a whole (both loops) can tolerate an additional single-failure of one of its active components and still perform the intended func-tion and prevent clad melt.

Therefore, i f a redundant active component fails, a longer reprir period is jus-tified based on the consideration given in the bases of Specification 3.2.

The consideration indicates that for a one out of.4 requirement the time out of service would be I

30 days 17.5 days TT7T =

1.11

=

Specification 3.4.A.5 ansures that if one diesel is out of service for repair, the core spray system loop on the other diesel must be operable with no compo-nents out of service.

This ensures that the loop can perform its intended function, even assuming one of its active components fails.

If this condition is not met, the reactor is placed ip a condition where core spray is no longer required.

When the reactor is in the shutdown or refueling mode and the reactor coolant system is less than 212'F and vented and no work is being. performed that could result in lowering the water level to less than 4'8" above the core, the likelihood of a leak or rupture leading to uncovering of -the core is very low.

The only source of energy that.must be removed is decay heat and one day after sh~utdown this heat generation rate is conservatively calculated to be not more than O.6% of rated power.

Sufficient core spray flow to cool the core can be supplied by one core spray pump or one of.the two fire protection system pumps under these conditions.

When it is 'necessary to perform repairs on the core spray system components, power supplies or water sources, Specification 3.4.A.7 per-i mits reduced cooling system capability to that which could provide sufficient core spray flow from two 1;

independent sources.

Manual initiation of'these systems is adequate since it can be easily accomplish-

~

r.

ed within 15 minutes during which time the temperature

,1, rise in the reactor will not reach 2200*F.

11 i

'l i

Amendment No.

75 1

F-

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-,...,,.,.,m..,.,,__c.--

_.,,.-m__.,,___.__,.-4 m_.

_p

,,.-..,, -. - -,-v.

1 3.4-9 In order to allow for certain primary system main-tenance, which_will' include control rod drive repair, LPRM removal / installation,; reactor. leak test,vetc.,

y.

-(alltperformed'according to approved procedure),

Specification 3.4.A.8 requires the availability of-an additional core spray pump in an, independent' loop,-

while this maintenance is being performed the likeli-hood of the core being1 uncovered is-still considered to be very low, however, the requirement of a second core. spray. pump capable of full rated flow and the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> operability demonstration of both core spray-pumps-is specified.

Specification 3.4.A.10 allows the core spray system to be inoperable in the cold shutdown or refuel modes if the reactor cavity is flooded and the spent fuel pool gates are removed and a source of water supply to the reactor vessel is available.

Water would:then be available to keep the cor.e flooded.

.: e The relief valves of the automatic depr~essurization

~

system enable the core spray system to provide pro-tection against the small break'in the event-the feedwater system is not activ,e.;

The containment spray system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident.

The flow from one pump in either loop is more than ample to provide the re-quired heat removal capability (2).

The emergency service water system provides cooling to the contain-ment spray' heat exchangers and, therefore, is required to provide the ultimate heat sink for the energy release in the event of a loss-of-coolant accident.

p The emergency service water pumping requirements

-)

are those which correspond to containment cooling heat exchanger-performance implicit in the containment i-cooling description.

Since the loss-of-coolant accident while in the cold shutdown condition would not require containment spray, the system may be deactivated to permit integrated leak rate testing of the primary containment while the reactor is in the cold shutdown condition.

(

i Amendment No. )MI 7 5

cr o.a.

u---

+~.

. - ~ - - - - - - -

3.4-10 The control rod drive hydraulic system can provide high pressure coolant injegtion capability.

For break sizes up to 0.002 ft, a single control rod drive pump with flow of.110 gpm is adequate for_ main-taining the water level nearly five feet above the core, thus alleviating the necessity for auto-relief actuation (3).

The core spray main pump compartments and containment spray pump compartments were provided with water-tight doors (4).

Specification 3.4.E ensures that the doors are in place to-perform their intended function.

Similarly, since a loss-of-coolant accident when pri -

mary containment integrity is not being maintained

-would not result in pressure build-up in the drywell or torus, the system may be made inoperable under these conditions.

This prevents possible personnel injury associated with contact with chromated torus water.

6

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Li I

References f

(1)

Licensing Application, Amendment 34, Question 1 (2)

Licensing Application, Amendment 32, Question 3 (3)

Licensing Application, Amendment 18, Question 1 l

(4)

Licensing Application, Amendment 18, Question 4 i

i Amendment No. jM( 7 5 j

1

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,o

w

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my

_7

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4

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l 3.5-4a

s and fission pror.ucts which could be released from any fuel failures resul~cing from the accident.

If the reactorfcoolant is not above 2125F, there would be no pressure. rise un the containment.

In addition, the coolant cannot be expelled, at a rate which could cause fuel failure to occar before the core spray system s

restores cooling-to'the core.

Primary containment is not needed while performing low power physics tests since procecures and the Rod Worth Minimjzer would limit rod worth such that a rod drop would.n0t result in any fuel damage.

In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer,a sufficient barrier to keep off-site doses well below 10 CFR 100 limits.

The absorption chamber water volume provides the heat sink for the reactor coolant system energy released following the loss-of-coolant accident.

The core spray pumps and containment spray pumps are locat~ed in.the corner rooms and due to their proximity to the torus, the ambient temperature in those rooms could rise during-the-design basis accident.. Calculations (7) made, assuming an initial torus water temperature of 100*F and a minimum water volume of 82,400 f t.3, indicate that the corner room ambient temper,ature would not exceed the core spray and containment spray pump motor operating temperature limits, and, therefore, would not adversely -

affect the long term core cooling capability.

The maximum water volume limit allows for an operating range without significantly affecting accident analyses with respect to freu air volume in the absorption chamber.

For example, the containment capability (8) with a maxtnum water volume of 92,000 <f t3 is. reduced by not more than 5?5% metal-water reaction below the capability 1

with 82,000 ft3 Experimental data indicate that excessive steam condensing loads can be avoided if the peak temoerature e

of the suppression pool is maintained below 1600F during any period of relief valve operation with sonic.

L conditions at the discharge exit. Specifications have been placed on the envelope of reactor operating l

conditions so that the reactor can be depressurized in a

'l timely manner to avoid the regime.cf potentially high suppression chamber loadings.

amendmene no. 7 q 4 p,

va w -

3.5-4b The technical specifications allow for torus repair work or inspections that might require draining of the suppression pool'when all irradiated fuel is removed or when the potential for draining the reactor vessel has been minimized..This specification also provides assurance that the irradiated fuel has an adequate cooling water supply for normal and emergency conditions with the reactor mode switch in shutdown or. refuel whenever the suppression pool is drained for inspection or repair.

The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression chamber and suppression chamber and reactor building so that the containment external design pressure limits are not exceeded.

The vacuum relief system from the reactor building to the pressure suppression chamber consists of two 100%

vacuum relief becaker subsystems (2 parallel sets,of 2 valves in series).

Operation of either subsystem will maintain the containment external pressure less than the external design pressure of.the drywell by 2 psi; the external design pressure of the suppression chamber is 1 psi (FDSAR Amendment 15, Section 11).

The capacity of the fourteen Suppression chambers to drywell vacuum relief valves is sized to limit the external pressure of the drywell during post-accident

~

drywell cooling operations to the design limit of 2 psi.

They are sized on the basis of the Bodega Bay pressure suppression tests. (9) (10)

In Amendment 15 of the Oyster Creek FDSAR,Section II, the area of 2920 sq.

in. is stated as the minimum area for flow of non-condensible gases from the suppression chamber to the drywell.

To achieve this requirement, at least 12 of the 14 vacuum breaker valves (18" diameter) must be i

operable.

t Each suppression chamber drywell vacuum breaker is fitted with a redundant pair of limit switches to provide fail safe signals to panel mounted indicators in j

the Reactor Building and alarms in the Control Room when the disks are open more than 0.1" at any point along the seal surface of the disk.-

These switches are capable of transmitting the disk closed-to-open signal with 0.01" movement of the switch plunger.

Continued reactor operation with failed components is justified because of the redundancy of components and circuits and, most i

Amendment No. 75 4

/

.m-a M

i i n.

I

L 9

3.5-5 importantly, the accessibility of the valve lever arm and position reference external to the valve.

The fail-safe feature of the alarm circuits assures operator attention if a line fault occurs.

Conservative estimates o5 the hydrogen produced, consis-tent with the core cooling system provided, show that the hydrogen air mixture resulting from a loss-of-coolant accident is considerably below the flammability limit and hence it cannot burn, and inerting would not be needed.

However, inerting of the primary containment was included in the proposed design and operation.

The 5% oxygen limit is the oxygen concentration limit stated by the American Gas Association for hydrogen-oxy mixtures below which combustion will not occur.(gen j

H To preclude the possibility of starting up the reactor and operating a long period of time with a significant leak in the primary system, leak checks must be made when the system is at or near. rated temperature and pressure.

It has been shown(9)(10) that an acceptable margin with respect to flam4 ability exists without containment inerting.

Inerting the primary containment provides additional margin to that already considered acceptable, Therefore, permitting access to the drywell for the purpose of leak checki,ng would not reduce the margin of safety below that considered adequate and is.

judged prudent in terms of the added plant safety offered by the opportunity for leak inspection.

The i

24-hour time to provide inerting is judged to be a reasonable time to perform the operation and establish the required 02 limit.

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown.

The

)

consequence of an inoperable snu'ober is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads.

It is, therefore, required that all snubbers required to protect the primary co,olant system or any

.t other safety system or component be operable during reactor operation.

i All safety related hydraulic snubbers are visually inspected for overall integrity and operability.

The I

Amendment No.

7g 4

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- - ~ ~ ~ ~ - -

+- !

-=

+

3.5-6 inspection ~will include verification of proper orienta-tion, adequate hydraulic fluid level and proper attach-ment of snubber to piping and structures.

Examination of defective snubbe'rs at reactor faciliti.es

.c and material tests performed:at several laboratories

-(Reference.11) has shown that millable gum polyurethane deteriorates rapidly.under the temperature and moisture conditions present in.many snubber. locations.

Although molded polyurethane exhibits greater resistance to these conditions, it also may be unsuitable for application.in the higher temperature environments.

Data are not cur-rently=available to define precisely an upper tempera-ture li,mit for the molded polyurethane.

Lab tests and in-plant experience-indicate that seal materials cre available, primarily ethylene propylene compounds, which should give satisfactory performance under the most severe conditions.expec.ted in reactor. installations.

Because snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacements.

In*chse~a' shutdown is required, the allowance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a cold shutdown con-dition will permit an orderly shutdown consistent with standard operating proceduresT' Since plant startup.

I should not commence with knowidgly defective safety related equipment, Specification 3.5.A.7.d prohibits startup with inoperable snubbers.

Secondary containment (5) is designed to minimize any ground level release of radioactive materials which might result from a serious accident.

The reactor

]

building provides secondary containment during reactor operation when the drywell is sealed and in service and j

provides primary containment when the reactor is shut-down and the drywell is open, as during refueling.

I Because the secondary containment is an-integral part.of the overall containment system, it is required at all-t times that primary containment is required.

Moreover, 1;

secondary containment is required during fuel handling

['

operati'ons and whenever work is being performed on the reactor or its connected systems in the reactor building since their operation could result in inadvertent q

release of radioactive material.

The standby gas treatment system (6) filters and ex-hausts the reactor building atmosphere to the stack 1

a.

?-

t.

.i'

[li

~ AmandmeEt No. 7 5

)

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a 3.5-7 during secondary containment isolation conditions, with a minimum release of radioactive materials from the reactor building to the environs.

Two separate filter trains, are provided each having 100%

capacity.(6)

If one filter train becomes inoperable, L

there is no~immediate threat to secondary containment and reactor operation may continue while repairs are l

being made.

Since the test interval.for this system is one month (Specification 4.5), the time out-of-service allowance of 7 days is based on-considerations presented in the Bases in Specification 3.2.for a one-out-of-two system. '

References:

(1)

FDSAR, Volume I, Section V-1 (2)

FDSAR, Volume I, Section V-1.~4.1 (3). FDSAR, Volume I, Section V-1.7 (4)

Licensing Application, Amendment 11,

-Question III-25 l

(5)

FDSAR, Volume I,.Section V-2 t

(G)

FDSAR, Volume Ii_ Section V-2.4 (7)

Licensing Application, Amendment 42 l

(8)

Licensing Application, Amendment 32, Question 3 (9)

Robbins, C.

H., " Tests of a Full Scale 1/48 Segment of the Humbolt Bay Pressure Suppres-sion Containment ~," UEAP-3596, November 17, 1960.

(10)

Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

(11)

Report H. R. Erickson, Bergen-Paterson to K. R. Goller, NRC, October 7,

1974,

Subject:

Hydraulic Shock Sway Arrestors.

In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed on August 2, 1976, which demonstrated ti factor of safety of at i

least two for the weakest element in the suppression chamber support system.

The maintenance of a drywell-suppression chamber differential p,ressure within the i

range shown on Figure 3.5-1 with a suppression chamber water level corrccponding tc a downcomer submergence range of 3.0 to 5.3 feet will assure the integrity of the suppression chamber when. subjected to post-LOCA suppression pool hydrodynamic forces.

e Amendment No.

{

75

?!

1

~

3.10-1 3.10, CORE. LIMITS Applicability:

Applies to core conditions required,to meet the Final Acceptance Criteria for Emergency Core Cooling _ Performance.

Objective:

To-assure conformance to the peak clad temperature limitations during a postulated loss-of-coolant accident as specified in 10 CFR 50.46 (January 4, 1974) and to assure I

conformance to the 14.5 KW/ft (for V and VB fuel)~and 13.4 KW/ft (for P8x8R fuel) operating limits for local linear heat' generation rate.

Specification:

A.

Average Planar LHGR During power operation, the average linear heat generation rate (LHGR) of all the rods in.any fuel assembly, as a function.-of average planar exposure, at any axial location shall not exceed:

~

A.1 Fuel Types V and VB I

The product of the maximum average planar LHGR l

(MAPLHGR) limit snown in Figures 3.10-1 (for 5-loop l

f operation).and 3.10-2 (for 4-loop operation) and l

the axial MAPLHGR multiplier in Figure 3.10-3.

A.2 Fuel Type P8x8R l

l Tne maximum average planar LHGR (MAPLHGR) limit shown in. Figure 3.10-4 (for 5-loop operation) and 3.10-5 (for 4-loop operation).

A.3 If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be l

initiated to restore operation to within the I

• pr'escribed limits.

If the APLHGR is not returned i

I to within the prescribed limits within two (2]

T hours, action shall be initiated to bring the i

reactor 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 />.

During this period surveillance and corresponding action shall continue until reactor operation is within the prescribed limits at which time power operation may be continued.

t

1. ~

Amendment No.

75 W-

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. i 3.10-2 B.

Local LHGR During power operation, the linear-heat generation rate (LHGR)- of any rod in any fuel assembly, at any.pxial. location shall not exceed the maximum allowable LHGR:

,s.

B.1 Fuel' Types 7 and VB o

As calculated by'the following equation; LHGR LHGRd [1-M max ( L_) ]

P

_LT Where:

LHGRd = Limiting LHGR (=14.5)

[iP

= Maximum Power Spiking Penalty P

(=0.033 and 0.039 for Fuel Types V and VB respectively)

LT

= T,otal' Core Length - Id4 inches L

= Axial position above bottom of core d.2 Fuel Type P8x8R 13.4$W/ft.

~

LHGR B.3 If at any time during operation it is deter -

4 mined by normal surveillance that the-limiting value of LHGR is being exceeded, action shall be initiated to restore operation to within the, prescribed li. pits.- If the LHGR is not returned to within the prescribed limits within two [2]

hours, action shall be initiated to bring the reactor to the cold shutdown condition within

'i 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed j

limits at which time power operation may be i

continued.

s,'

f 3

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Amendment No.

75 l

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i 3.10-3 C.

Min'imum Critical Power Ratio (MCPR).

During steady state power operation, MCPR shall he l

greater than or equal to the following APRM Status MCPR Limit 1.

If ar.y two (2) LPRM assemblies which 1.40 l

are input to the APRM system and are separated in distance by less'than i

three (3) times the control rod-pitch contain a combination of (3) out of four (4) detectors 1ocated in either

~

the A and B or C and D levels which are failed or bypassed i.e.,

APRM channel. or LPRM inpu': bypassed or inoperable.

2.

If any LPRM input to 'the APRM system 1.40 l

at the B, C, or D level is failed or bypassed or any APRM channel is in-operable (or bypassed).

~'

3.

All.B, C, and D LPRM inputs-to the 1.40 l

APRM system are operating And no APRM channels are inoperable'or-bypassed.

When APRM status changes due to' instrument fa'ilure (APRM or LPRM input failure), the MCPR requirement for the degraded condition shall be met within a time interval of eight (8) hours, provided that.the control rod block is placed in operation during this interval.

For core flows other than rated, the nominal value for MCPR shall be increased by a f actor of kg, where kg is as shown in Figure 3.10-6.

If at any time during power operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded for reasons other than instre-i.

ment failure, action shall be initiated to restore t'

operation to within the prescribed limits.

If the (2

steady. state'MCPR is not returned to within the l'

prescribed limits within two [2] hou'rs, action shall l

be initiated to bring the reactor to the cold shutdow.i condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

During this period, surveillance and corresponding action shall continue 1

Amendment No.

75 t

i

!l

. I.

a.

=-

~

3.10-4 until reactor operation is w'ithin the prescribed limits at which time power operation ~may be continued.

, Bases:

f The Specification for average planar LEGR' assures that.the

. peak cladding temperature following the postulated design-

= m basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46 (January 4, 1974) 4

considering the postulated effects of fuel pellet

~

densification.

The peak cla'dding temperature following a postulated

~

loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly:at any axial location and~is only dependent secondarily on the rod to. rod power distribution within an assembly. 'Since. expected location variations in power

distribution within a fuel assembly. affect the calculated peak clad tenperature by less than + 20*F relative to the peak temperature for a typicaf fuel design, the limit on

~

the average linear heat generation rate.is sufficient to assure that calculated temperatures are below the limits 4

specified in-10 CFR 50.46 (January 4, 1974).

The maximum average planar LHGR~ imits of fuel types V and TMS are shown in Figure 3.10-1 for five loop operation and 1

in Figure 3.10-2 for four loop operation, and are the result of LOCA analyses performed by Exxon Nuclear Company

<g

~

utilizing an evaluation model developed by Exxon Nuclear Company.in compliance with Appendix K to 10 CFR 50 (1).

Operation is permitted with the four-loop limits of Figure 3.10-2 provided-the'fifth loop has its discharge valve closed and its bypass and suction valves open.

In addition, the maximum average planar LHGR limits shown in Figures 3.10-1 and 3.10-2 for Type V and VB fuel were analyzed with 100% of the spray cooling coefficients

!b specified in Appendix K 'to 10 CFR Part 50 for 7 x 7 fuel.

Th'ese spray heat transfer coefficients were justified in

.i i

the ENC Spray Cooling Heat Transfer Test Program (2).

'The maximum average planar LHGR limits of-fuel type P8x8R

[

are shown in Figure 3.10-4 for five' loop operation and in Figure 3.10-5 for four loop operation, and are based on i.

calcula'tions employing the models described in Reference 3.

l-f 4

Amendment No.

75 1

d

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^

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w _, ga..

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t 3.10-5 y

Power operation with LHGR's at or below those shown in Figures 3.10 and 3.10-5 assures that the peak cladding temperature following a postulated loss-of-coolant

~~

accident will not exceed 'the 2200*F limit.

The effect of axial; power profile peak' location-for 1 fuel.

types V and.VB is evaluated.for the worst break size by_

performing a-series of. fuel heat-up calculations.

A. set of multipliers.is devised'to reduce the allowable bottom skewed axial power. peaks relative to center or above

' center peaked profiles.- The major factors which lead to

-the lower MAPLHGR limits-with bottom skewed axial power.

profiles are the-change in canister quench time at the axial _ peak location _and a. deterioration in heat transfer durings the extended downward flow period during blowdown.

The MAPLHGR multiplier. in Figure 3.10 shall only be

. applied to MAPLHGR determined'by the evaluation model described in reference 1..

The possible effects of fuel pellet densification are:

1) creep collapse of the clad' ding due to axial gap formation;

~ 2) increase in the LHGR because of pellet column shortening,

~

3) power spikes due to axial gap formation; and 4) changes in stored energy-dde to increased radial gap j..

size.

)I Calculations show.that clad collapse is conservatively l

predicted not to occur during the exposure lifetime of the j

fuel.. Therefore, clad collapse is not considered in the

~

analyses.

Since axial thermal expansion of-the fuel pellets is greater than axial shrinkage due to densification,'the analyses of peak clad temperatures do not consider any 4

change in LHGR due to pellet column shortening.

Although j

the formation of axial gaps might produce a local power i

spike at one location on any one rod in a fuel assembly, the increase in local power density would be on the order of only 2% at the axial midplane.

Since small local H

variations in power distribution have a small effect on peak clad temperature, power ' spikes were not considered in the analysis of loss-of-coolant accidents (l).

l Amendment No.

75

[]

i Qw ;_.., y. '

' -- -~

--_w ew

_ w-~

.m,+,.,,,

, _ _,,.. - - - _ ^

r 3.10-6 Changes in gap size affect the peak clad temperatures by their effect on pellet clad thermal conductance and fuel pellet stored energy.

Treatment of this effect combined with the effects of pellet cracking, relocation and subsequent gap closure are discussed in XN-174.

Pellet-clad thermal conductance for Type V and VB fbel was calculated using the GAPEX model (XN-174).

The specification for local LHGR assures that-the linear

~

heat generation rate in any rod is less than the limiting linear heat generation rate even if fuel pellet densifica-tion.is postulated.

The power spike' penalty for Type V and VB fuel is based on analyses presented in Facility Change Request No. 6 and FDSAR Amendment No. 76, respec-tively.- The analysis assumes a linearly increasing variation in axial gaps between core bottom and top, and

~

assures with 95% confidence that no more than one fuel rod exceeds the design linear. heat generation rate due to power spiking.

The power spike penalty for fu'e ' type P8x8R is described in Reference 3.

-: ?

The loss of coolant accident (LOCA) analyses are performed using an initial core flow that-is 70% of the rated value.

The rationale for use of this vafue of flow is based on the possibility of achieving full power (100% rated power) at a reduced flow condition.

The magnitude of the reduced-flow is limited by the flow relationship for overpower i

i scram.

The low flow condition for the LOCA analysis ensures a conservative analysis because this initial con-dition is associated with a higher initial quality in the core relative to higher flow-lower quality conditions at full power.

The high quality-low flow condition for the steady-state core operation results in rapid voiding of the core during the blowdown period of the LOCA.

The rapid degradation of the coolant conditions due to voiding l'

results in a decrease in the time to boiling transition and thus degradation of heat transfer with consequent higher peak cladding temperatures.

Thus, analysis of the LOCA using 70% flow and 102% power provides a conservative basis for evaluation of the peak cladding temperature and the maximum average planar linear heat generation rate

}

(MAPLHGR) for the reactor.

Amendment No.

75

/

4

--7 i

i 3.10-7

'The' APRM response is'used to predict when the rod block occurs in-the analysis'of the rod withdrawal error transient.

The transient rod position'at the rod block and corresponding MCPR can ce determined.

The MCPR has been evaluated for different APRM responses which would result from changes-in the'APRM status as a consequence of bypassed APRM. channel and/or, failed, bypassed LPRM inputs.

The results for'the' reference' cycle (3). indicate that the" steady: state MCPR required to protect the minimum

~

transient ~MCPR of 1.07 is 1.23 or higher for the worst case APRM status condition (APRM STATUS 1).

This steady state limit conservatively applies to APRM status 2 and 3.

The steady state }1CPR values for APRM status condi--

tions 1, 2, and 3 will be evaluated each cycle.

In order to provide for a limit which is considered to be bounding' to future operating cycles, the' limits for each APRM status condition have been conservatively adjusted upward to 1.30.

This is also the assumed valpe for LOCA. analysis.

The time interval of eight (8) hours to adjust the steady state MCPR to account for a degradation in the APnM status is justified on the basis of instituting a control rod block which precludes the possibil,ity of experiencing a rod withdrawal error transient s,ince rod withdrawal is physically prevented.

This tim # i'nterval is adequate to

-allow the operator to either increase the MCPR to the appropriate value or to upgrade.the status of the APRM system while in a condition which prevents the possibility.

of this transient occurring, f"

i The steady-state MCPR limit was selected to provide margin to accommodate transients and'uncertanties in monitoring the core operating state, manufacturing, and in the critical power correlation itself(3).

This limit was derived by addition of the [hCPR for the most limiting abnormal operational transient caused by a single operator error of equipment malfunction to the fuel cladding

. integrity MCPR limit designated in Specification 2.1.

Transients analyzed each fuel cycle will'be evaluated with respect to the steady-state MCPR limit specified in this specification.

The purpose of the Kg factor'is to define opgrating l

limits at other than rated flow condit' ions.

At less than 100% flaw the required MCPR is the product of the i.

operating limit MCPR and the Kg factor.

Specifically, 4

the Kg factor provides the required thermal margin to protect against a flow increase transient.

AmenCaent No. 7 5 u-

__ e

...m

'g We44 r -., mm.m.smire -am M 4.=

E wh,-

J' 4-===re w--

r ses=

e-.-

3.10-8 The Kg 'f actorf curves shown in Figure-3.10-6 were devol-oped. generically using'the flow control line corresponding to. rated thermal-power-at rated core flow and are appli-

~

cable to all -BWR/2, BWR/3 and BWR/4 reactors.- For the

~

manual: flow control mode, the Kg factors were calculated such that at the maximum flok stath (as limited by the pump scoop tube set point).and the corresponding _ core power (along the rated flow control. line),- the limiting

' bundle's relative' power was adjusted until the MCPR was slightly above the Safety Limit.

Using this relative-bundle power, the MCPR's were calculated at different points along the rated flow control line corresponding-to-dif ferent core flows.

The ratio of the MCPR calculated at a given point of core : flow, divided by the operating limit MCPR determines the value of Kg.

REFERENCES (1)

'.XN-7 5-5 5- ( A), XN-75-5 5, Supplement 1- ('A), XN-75-55.,

Sup-plement 2-(A), Revision 2,

" Exxon Nuclear Company-WREM-Based NJP-BWR ECCS E',aluation Model and Application to the, Oyster Creek plant," April-197,'7.

(2)

XN-75-36 ( NP ) - ( A )', XN-75-36(NPl_hupplement1-(A), " Spray Cooling Heat Transfer phase Test Results, ENC - 8 x 8 BWR

~

Fuel 60 and 63 Active Rods, Inte' rim Report," October 1975.

(3)

NEDO-24195; General Electric Reload Fuel Application for Oyster Creek.

.+

~

e I

1 l

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s s 4 j, 9 i, l m. . _y_ .7 e 8 1 t 0 5 10 15 20. 25 30 3L 40 j AVERAGE PLANAR EXPOSURE (GWD/HTH) a. g i Amendment No. 75 e I i j J-4- 9i ? 1 6 FIGURE 3.10. AIIAL MAPLHCR MULTIPLIER (FOR FUEL TYPES V AND VB ONLY) 2.2 _ _c _ t e. . c. e . c.. a y . _ y D 0.9 s . g e e d 0.8 0.7 0 2 4 6 8 10 12 u CORE BEIGHT (FT) ~ ~o aw M. Amendment No. 75 e F-w-. .....a w. u : ~. + A: FIGilRii 3.10 I x HAXIttlH AI.l.ONABl.li AVERAGE PLANAR I i', LINEAR lil!AT GliNERATION RATE (FIVE LOOP OPERATION)- 4: l, .g; ....__u.-- j. 33,o l;l ~. ll e l-i l 1 , l l. .p i., i l ~ ' .( \\ so.s \\ ( F m 9 f s I; 10.0 l [ 1 l 'r - y. a 5 raona239 j. l a reonu26 sis ] g*3 .i. I - ti l ii

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, 48 Amendment No. 75 AVERAGE Pl.ANAR.EXPOSilRE (GiD/HTH)

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Amendment No. 75 AVi! RAGE PLANAR liXPOSt'RE (GWD/l'111)

d 3.10-14 i

FIGURE 3.10-6 FLOW FACTOR, Kg i

14 t.3 -

1.2 AUTOMATIC FLOW CONTROL e-

"f 1.1 MANUAL FLdW CONTROL I

scoop tube position limited such that rLOWMAX

• 1023%

• 107.0%

= 112.0%

• 117.0 %

l1 t.0 t

{'

l l

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I 30 40 50 60 70 80 90 100 COR$ FLOW t%

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Amendment No. 75 i

4.2-1 4.2 REACTIVITY CONTROL

' Applicability:

Applies to the surveillance requirements for reactivity control.

objective:

To verify the capability for controlling',

reactivity.

1 Specification:

A.

Suf ficient control rods shall be withdrawn following a refueling outage when core alterations were performed to demonstrate with a margin of 0.25% /hk that the core can be made subcritical at any time in the subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable-rods fully inserted.

B.-

The control rod drive housing support system shall be inspected after reassembly.~

C.

1.

After each major refueling outage and prior to

~

resuming power operation, all operable control rods shall be scram time tested from the' fully withdrawn position with.-reactor pressure above 800 psig.

,. J, 2.

Following each reactor scram from rated pressure, -

4 the mean 90% insertion time shall be determined for eight selected rods.

If the mean 90%

insertion time of the selected control rod drives does not fall within the range of 2.4 to 3.1 seconds or the measured scram time of any one drive for 90% insertion does not fall within the range of-1.9 to 3.6 seconds, an evaluation shall

- j be made to provide reasonable assurance that 3

proper control rod drive performance is j.

maintained.

3.

Following any outage not initiated by a reactor

' cram, eight rods shall be scram tested with' s

reactor pressure above 800 psig provided these have not been measured in six months.

The same

- {~

criteria of 4.2.C(l) shall apply.

4 ll Amendment No.

75

/

m -.

-= ~. - -

c --

ry ;.

Qg..,

c 4.2-la D.

Each partially or fully withdrawn control rod shall be exercised at least once each week.- This-test shall be performed at least.once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation isicontinuing'with,two or more-inoperable

. control rods or.in the event power. operation is-continuing with one. fully or' partially withdrawn rod

. hich cannot be moved and for which control rod drive w

mechanism damage has not been ruled out. The

. surveillance need not'be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of. inoperable rods has been reduced to less than two and if it has been demonstrated that control q

rod drive' mechanism collet housing failure is not the cause of an immovable control rod.

E.. Surveillance of the standby liquid control system shall be as follows:

1.

Pump operability Once/ month

~

2.

Boron concentration

_ 0nce/ month determination 3.

Functional test Each refueling outage-4.

Solution volume and temperature check

-Once/ month

.j:

i

!p l

t

)-

4 Amendment No.-

75 I.

j i

+

- ~

e

~

4.2-2 F.

At specific power operation conditions, the actual control rod configuration will be compared with the

- expected configuration based upon appropriately corrected past data. This comparison shall be made every eouivalent full power month. The initial rod inventory measurement performed when equilibrium conditions are established after a refueling or major core alteration will be used'as base' data' for

~

reactivity monitoring during subsequent power operation throughout the fuel cycle.

I G.

The scram discharge volume drain and vent valves shall be verified open at least once per 31 days, e'xcept in shut-down mode *, and shall be cycled at least one complete cycle of full travel at least quarterly.

H.

All withdrawn contol rods shall be deterpined OPERABLE by demonstrating the scram discharge yolume drain and vent valves OPERABLE. This will be done at least once per refueling cycle by placing the mode switch in shutdown and by verifying that:

The drain and vent valves close with'in 60 seconds a.

after receipt of a signal for control rods to scram, j

ud b.

The scram signal can be reset and the drain and vent valves open when the scram discharge volume trip is bypassed.

Basis:

The core reactivity limitation (Specification 3.2. A) requires that core reactivity be limited such that the core could be t

made suberitical at any time during the operating cycle, with i

the strongest operable contol rod fully withdrawn and all other operable rods fully inserted.

Compliance with this reqsirement can be demonstrated conveniently only at the time of refueling.

Therefore, the demonstration must be such that it will apply to

]

the entire subsequent fuel cycle. The demonstration is performed

'i with the reactor core in the cold, xenon-free condition and will show that the reactor is sub-critical at that time by at least R + 0.25% Ak with the highest worth operable control rod fully i

withdrawn.

These valves may be closed intermittently for testing under administrative control.

j

.?

Change No. 4t; Amendment No.Jr$j, G W

~

~*

-q y

=b

=~

j s'h 4.2-3 I

The value of R is the difference between two calculated-values of reactivity of the cold,

-xenon-free core with the strongest operable control rod fully withdrawn.

The reactivity value at the-

. beginning of life is subtracted from the maximum reactivity -value anytime later in life to determine R, which must be a positive quantity or its value is conservatively taken as zero.

The value of R shall include the potential shutdown margin loss assuming

~

full B C settling in all possi;bly inverted tubes 4

present in the core.

The va*1db 0.2'5%/1k in the expression R + 0.25%.[hk se'rve_s at the beginning of life as a finite, demonstrable shutdown margin.

This margin is demonstrated by.fdII withdrawal of th'e strongest-rod 'and partial withdrawal of a diagonally l

adjacent rod to a position calculated to insert an R +

0.25% [ik reactivity.

Observation of subcriticality in this condition assures subcriticality with not only the strongest rod fully withdrawn but at least an R +

0.25% [ik margin beyond this.

The control rod drive housing support system (2) is not subject to deterioration during operation.

However, reassembly must be assured following a partial or complete removal.

1 The scram insertion times for all control rods ('3) will be determined at the time of each refueling i

outage.

The scram times generated at each refueling outage when compared to scram times previously re-i corded gives a measurement of the functional effects

'l of deterioration for each control rod drive.

The more frequent scram insertion time measurements of eight selectedLrods are performed on a representative sample basis to monitor performance and give an early indication of possible deterioration and required maintenance.

The times given for the eight-rod tests V

Amendment No.-

75

, i.i.,

&~

4 4.2-4 are based on the testing experience of control rod drives which were known-to be in good condition.

'The' weekly control rod exercise test serves as a c

periodic check against deterioration of the control

.c rod system. -Experience with this control rod system has indicated that weekly tests are adequate, and that rods which move by drive pressure will scram when required as the pressure applied is.much higher.

The.

frequency of exercising the control rods has been-increased under~the conditions of two or.more control rods which are valved out of service in order'to provide even further assurance ofethe reliability of the remaining ' control rods.

Pump operability, boron concentration, solution' temperature and volume of the. standby liquid control system (4) are checked on a frequency consistent with

~

instrumentation checks described in Specification 4.1.

Experience with similar systems.has indicated.

that the test frequeacies are' adequate.

The only practical time to functionall?' test the liquid control system is during a refueling outage.

The functional test includes the firing of explosive charges to open the shear plug valves and the* pumping of demine'ralized water.into the reactor to assure. operability of the system downstream of the pumps.

The test also includes recirculation of liquid control solution to 1

and from the solution tanks.

Pump operability is demonstrated on a more frequent basis.

This test consists of recirculation of demin-eralized water to a test tank.

A continuity check of the firing circuit on the shear plug valves is provided by pilot lights in the control room.

Tank level and temperature alarms are provided to alert the operator to off-normal conditions.

The functional test and other surveillance on components, along with the monitoring instrumentation, givis i high reliability for standby lignid control system operability.

.L References (1)

FDSAR, Volume II, Figure III-5-ll (2)

FDSAR, Volume I, Section VI-3 (3)

FDSAR, Volume I, Section III-5 and Volume II, Appendix 3 i

(4)' FDSAR, Volume I, Section VI-4.

l 4

Amendment No.

75 g.

a =~: -

4.10-1 4.10.ECCS RELATED CORE LIMITS Applicability:

Applies to the periodic measurement during power operation of core parameters related to ECCS performancen Obiective:

To assure that the limits of.Section 3.10 are not being violated.

Specification:

A.

Average Planar LHGR.

The APLHGR for each type of fual as a function of average planar exposure shall be checked daily during-reactor operation at greater than or equal to 25%

rated thermal power.

B.

Local LHGR The LHGR as a function of core height shall be checked daily during reactor operation at greater than or equal to 25% rated thermal power.

C.

Minimum Critical Power Ratio- (MCPR).

MCPR shall be checked daily during reactor operation at greater than or equal to 25% rated thermal power.

2 Bases:

The LHGR shall be checked daily to determine whether fuel burnup or control rod movement has caused changes in power distribution.

Since changes due to burnup are slow, and only a few control rods are moved daily, a daily check of power distribution is adequate.

The minimum critical power ratio (MCPR) is unlikely to 4

change significantly during steady state power operation so-that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an acceptable frequency for surveil-I lance.

In the event of a single pump. trip, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance interval remains acceptable because the accompanying power reduction is much larger than the change in MAPLHGR limits for four loop operation at the corresponding lower steady state power level as compared to five loop operation.

The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency is also 1

Amendment Noe 75 y

1u

--q

_m.

-)

l 4.10-2 acceptable for the APRM status check since neutron moni-toring system failures are infrequent and a downscale failure of either an APRM.or LPRM initiates a control rod withdrawal block, thus precluding the p'ossibility of a control rod withdrawal error.'

At core power levels less than or equal to 25% rated ther-mal power the reactor will be operating at or.above the minimum recirculation pump speed.. For all designated control rod patterns which may be employed at this point, operating plant experience and thermal hydraulic analysis indicate that the resulting APLHGR, LHGR and MCPR values all have considerable margin to the limits of Specifica-tion 3.10.

Consequently, monitoring of these quantities below 25% of the rated thermal power is not required.

~

6

?~

4

• ?

9 i

Amendment No.

y3 4

m

,,-m

_