ML17277B121

Marked-up Tech Specs Re Recirculation Sys Single Loop Operation

ML17277B121
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Site:	Columbia
Issue date:	11/23/1983
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
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NUDOCS 8312020309
Download: ML17277B121 (22)

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2.0 SAFE, LIMITS and LIMITING SAFETY SYSTBf SETTINGS BASES INTRODUCTION I l I

The fuel cladding, reactor pressure vessel and primary system piping are

~the principal barriers to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limi".

is not violated. Because fuel damage is not directly observable, a step-back appr oach is used to establish a Safety Limit such that the MCPR is not less than 1.06 MCPR reater than 1.06 represents a conservative margin relative to e con stsons required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perfor'ations or cracking. Although some corrosion or use related cracking, may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design condi-tions and the Limiting Safety Sys em Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding integrity Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.'.

1 SAFETY LIMITS 2.1.1 THERMAL POWER Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows 'less than 10~ of rated flow. Therefore, the fuel claading in egrity Safety Limit is established by a/her means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always. be greater than 4.5 psi. Analyses. show that with a bundle flow of 28 x 10'bs/h, bundle pressure drop is nearly independent of bundle power and l as a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10s lbs/h. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel'ssembly critical power at this flow is aoproximately 3.35 llWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50Ã of RATED THERMAL POWER.

Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

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Bases Table 82.1.2"1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFE7Y LIMIT" Standard Deviation

~URllt, 1 of Point Feedwater Flow 1. 76 Feedwater Temperature 0. 76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow,. v~o R~<<c. <<'P 2.5 opuaUI" I'~~'cacc-Loop 4.o Channel Flow Area '3.0 Friction Factor Multiplier 10. 0 Channel Fric ion Factor Multiplier \ 5.0 P

TIP Readings, 7gyo ZcGlcc ~ Lo P 0/cc God 6.3 Factor Oh'e ie~ L~p Op<ca4 R 1.5

'Critical Power 3.6

" The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

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MASHINGTON NUCLEAR - UNIT 2 B 2-3

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3/4.2 . POWER DISTRIBUTION LIMITS 3/4.2.i AVERAGE PLANAR LiNEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION I

3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (A'PLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3. The limits of Figures 3.2.1-1,

3. 2. 1-2, and 3. 2. 1-3 shall be reduced to a value of 0. 84 times the two recircu-lation loop operation limit when in single recirculation loop operation.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.

ACTION:

With an A?LHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, or 3.2.1-3, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25Ã of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 1

4.2.1 All APLHGRs shall be verified to be equal to or less than the limits 1 deearnined from iigores 3.2.1-1, 3.2.1-2, and 3.2.1-3$ ofisofne1d>> od~oshdpea sfeo>boo~ii<

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a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after, completion of a THERMAL POWER increase of at least 15,. of RATED THERMAL POWER, and

c. Initially and at leas once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

WASHINGTON NUCLEAR - UNIT 2 3/'0 /.

OCT > 7 lo83

3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM a

RECIRCULATION LOOPS LIMITING CONOITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS l~ and 2".

ACTION:

a. Mith one reactor coolant system recirculation loop not in operation:

1. Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

LOC./AL a) Place the recirculation flow control system in the H~~

Manualmode, and

( POZi4yQ QbPR fCOL) b) Reduce THERMAL POWER to < 50M of'ATED THERMAL POWER, and, c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.07 per Specification 2.1.2, and, g Iphregge gate M)?PR L'tipg Co d?tp6n fg/ Op?/raP'ion/by/O.{f{

p'er Sgecyficion . 2. 3g an, g) + Reduce the Rate (MAPLHGR)

Maximum Average Planar Linear Heat limit to a value of 0.84 times Generation the two recircula~ion loop operation limit per Specification 3.2.1, and, d) ~f Reduce the Average Power Range Mon?tor {APRM) Scram and Rod Block and Rod Block Monitor rip Setpoints and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6.

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a) Ve i fy at t f e AP P lug noise verag d ove 30 . 'nutes doe not xcee 5X P ak tb peak; therw se, re uce he eci ulat'on lo f! w unt'1 the RM f x noi . is less t an t e 5~ eak o pe lime and, b) Ver fy t qt th core lat M noi g does ot gceed ps?

peak o pekoe; ot erwise, re use the e~circ a.id~loop flow u til tge LPWoise s less than the 1 ps lim<<4<<

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INSERT 1 f.) Reduce the volumetric flow rate of the operating recirculation loop to < 45,000** gpm g.) Perform surveillance requirement 4.4.1. 1.3 if thermal power is

~ 30%*** of rated thermal power or the recirculation loop flow in the operating loop is < 50/*** of rated loop flow.

h.) Reduce recirculation loop flow in the operating loop until the APRN flux noise does not deviate from the established flux noise patterns at 50% power by more than 50/ when averaged over a 30 minute period.

i.) Reduce recirculation loop flow in the operating loop until the core plate 4 P noise does not deviate from the established core plate 3P noise patterns at 50/ power by more than 50%;

INSERT 2 This value represents the design volumetric recirculation loop flow which produces 100/ core flow at 100% thermal power. The actual value to be applied will be determined during the Startup Test Program.

~*~ Initial values. Final values to be determined during Startup Testing based upon the threshold thermal power and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

REACTOR COOLANT SYSTEM r'

LIMITING CONDITION FOR OPERATION Continued ACTION: (Continued)

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3. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. With no reactor coolant system recirculation loops in operation, immediately initiate measures to place the unit in at least HGF QiuXMHN

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.4.4.1.1.KEach reactor coolant'ystem recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifyi'ng that the control valve fails "as is" on loss of hydraulic pressure (at the hydraulic control unit), and

b. Verifying that the average rate of control valve movement is:

1. Less than or equal to lI of stroke per second opening, and

2. Less than or equal to 12K of stroke per second closing.

l ~'zen,4 3 "See Special Test Exception 3.10.4.

I y4CCt.T WASHINGTON NUCLEAR - UNIT 2 3/4 4"2 OCT1 7 l983

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I 4.4.1.1.3 With one reactor coolant system recirculation loop not in operation , within no more than 15 minutes prior to either THERNAL POWER increase or recirculation loop flow increase verify that the fo11owing differential temperature requirements are mettIi Py wibc l S 0 0 ~ gt' D ERNR~ P~~~ olt,

a. < 145'F between reactor vessel steam space coolant and bottom Read drain line coolant,

b. < 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. < 50'F between the reactor coolant within the loop not in operation and the operating loop.

The differential temperature requirements, 4.4.1.1.3.b and c, do not apply when the loop not in operation is isolated from the reactor pressure vessel.

REACTOR COOLANT SYSTEM r~ JET PUMPS l LIMITIHG CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

Wi h one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REOUIREMEHTS

4. 4. 1.2. 1 Each of the above requi~ed jet pumps shall be demonstrated OPERABLE prior .o THERMAL POWER exceeding 25X of RATED THERMAL POWER and at least once per 2o hours by determining recirculation loop flow, .otal core flow and diffuser to-lower plenum differential pressure for each jet pump and verifying that na two of the following canditions occur when tive recirculation loops are r s<<w<<(

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a. The indicated reci'rculation loop flow differs by more than 10Ã from the established flow control valve position-loop flaw character isticsgoe g g4C'Cud¹kiffvt PC ~P tPPOCtt6~.

b. The indicated total core flow differs by more than 10K from the established total core flow. value derived from recirculation loop fl OW meaSurementSmt'<k Z fu<<dd'tM¹4~ p.~p Opgfg¹6~.

c. The indicated diffuser-to-lower plenum differential pressure of any individuaI jet pump differs from es ablished patterns by more than 1'. 2. RCuCQ¹6ffw PtisviP

4. 4.1.2.2 , luring single recirculation loop operation, each of the above required jet pumps shall be demonstrated OP'ERABLE 'at'least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that ~ms-

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'ffASHIHGTON NUCLEAR - UNIT 2 3/S +-8v.

OCT1 7 Lo83

INSERT 4 a.) The indicated recirculation loop flow in the active loop differs by more than 10Ã from the established. single pump flow control valve position - loop flow characteristics.

b.) The indicated total core flow differs by more than 10/ from the established total core flow value from single recirculation loop flow measurements.

c.) The indicated difference-to-lower plenum differential pressure of any individual jet pump differs from established single pump patterns by more than 10K.

REACTOR COOLANT SYSTEH RECIRCULATION LOOP FLOW LIHITING CONOITION FOR OPERATION 3.4.1.3 Recirculation loop flow mismatch shall be maintained within:

a. 5Ã of r'ated recirculation flow with core flow greater than or equal to 70K of rated core flow.

b. 10Ã of rated recirculation flow with core flow less than 7QX of rated core flow.

APPLICABILITY. OPERATIONAL CONOITIONS 1~ and 2m~4, 4u o ACTION:

With the recirculation loop flows different by move than the specified limits, either:

a. Restore the recirculation loop flows to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or onC of

b. DeclareVthe recirculation loops~~ l~op~h<

operon. and take the ACTION required by Specification 3.4.1.1.

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SURYEILLANCE RE UIREHENTS 4.4.1.3 Recirculation loop flow mismatch shall be verified to be within .he limits at least once pev 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

"See Special Test Exception 3.10.4.

WASHINGTON NUCLEAR - UNIT 2 3/4 4-4 OC~1 I 1983

REACTIVITY CONTROL SYSTEMS BASES 3/4. 1.3 CONTROL RODS

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The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the safety 'analyses, and (3) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operaiion. A limitation on. inoperable rods is set such that the resultant effect on 'total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to 'a time period which is reasonable'o determine the cause of the inoperability and at the same tine prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully inserted position are consis.ent with the SHUTDOWN MARGIN requirements.

The number of control rodspermitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable

.rods could be indicative of a generic problem and the reactor must be shutdown for investiqation and resolution of the problem.

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The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than during the limiting power transient analyzed in Section 15.2 of the FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than limit/. The occurrence of scram times longer then

~/fuel those cladding safety specified should be viewed as an indication of a problem with the rod drives and there" fore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time witil a potentially serious problem.

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The scram discharge volume--is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would .result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

WASHINGTON NUCLEAR " UNIT 2 B 3/4 1-.2 JUN

3/4 2 POWER OISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) 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 dependent only second-arily on the rod to rod power distribution within an assembly. The peak clad temperature is'calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for. densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady<tate gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR js shown in Figures 3.2.1-1, 3.2.3.-2 and 3.2.1-3)oc ~u ~ w<a<4~~

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calculational procedure used to establish the APLHGR shown on Figures 5'he 3.2. 1"1, 3. 2.1-2, and 3.2.1-3 is based on a loss-of-coolant accident analysis.

The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR,Part 50.

A complete discussion of each code employed in the analysis is presented in Reference 1. Oifferences in this analysis compared to previous analyses can be broken down as follows.

Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.

2. Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.

3. Corrected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2"1 JUN ~ P- 19&3

These values are to be multiplied by a factor of 0.84 for single loop opera.ion. The multiplier for single loop operation is determined from comparison of the limiting analysis between one loop and two loop operation.

The',

in FSAR Section SA 'he calculational procedures and significant input parameters reduction factor derived for', single-loop operation is justified in 4mendmerrt~,

are'ocumented the FSAR.

5 .~

POWER DISTRIBUTION LIilITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.

2. Incoporate NRC pressure transfer assumption - The assumption used in the SAfE-REFLOOD pressure transfer when the pressure is increasing was changed.

A few of the changes affect the accident calculation irrespective of CCFL. These changes are listed below.

1. . Break Areas - The DBA break area was calculated more accurately.

1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.

A list of the significant pl'ant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2. 1-1.

3/4.2.2 APRM SETPOINTS 1

The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR a. RATED THERMAL PO'~ER. The flow biased simulated thermal power upscale scram settin and control rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than %6& /fuel cladding safety limitg or that > I plastic s rain does not occur in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the'ombination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

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WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-2 JULY 2 2 PPP

POWER DISTRIBUTION LIMITS BASES 3/4. 2. 3 MINIMUM CRITICAL'OWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR ~=66 and an analysis of abnormal operational transients. For any abnormal operating transient analysis evalua" tion with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit, is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

When added.to the Safety Limit,MCPR ~~,

The limiting transient yields the largest delta MCPR.

the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3"1.

The eva'luation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient computer program. The code used to evaluate pressurization events is described in NEDO-24154(l) and the program used in nonpr essurization events is described in NEDO"10802(2). The outputs of '.this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraulic TASC code described in NEDE-.25149(3). The principal result of this evaluation is the reduction in MCPR caused by the transient.

The purpose of the Kf factor of Fiaure 3.2.3-1 is to define operating limits at other than rated core flow conditions. At less than 100K of rated

,low the required MCPR is the product of the MCPR and the Kf factor. The Kf factors assure that the Safety Limit MCPR will not be violated. The Kf factors were derived using THERMAL POWER and core flow corresponding to 105 of rated steam flow.

The Kf factors were calculated such that for the maximum core flow rate and the corresponding THERMAL POWER along the 105K of rated steam 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 MCPRs were calculated at different points along the 105K of rated steam flow control line corresponding to different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kf.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 2-4 SEP 12 583

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 RECIRCULATION SYSTEM' H S~k'P (p gp rat on i h ne retctor r cop1an rpcir ul ti n i/op igdpef'hbl ppohib ed un 1 an v luatioh o t e phrr rephc of th EgC'S upohg ph op'p dpe/at on ha b en fopbedr e a1 ate , nd Ale reu ne tofu aa'cpfltdb e.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design"basis-accident, increase the.blowdown area and reduce the capability of ref looding the core; thus, the requirement for shutdown of the faci'lity with a jet pump inoperable.

Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criteria. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. iHscev In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle loop. The loop temperature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue. stress on the vessel would result if the tempera ure difference was greater than 145~F.

3/4. 4. 2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate'o prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig in accordance with the ASME Code. 'A total of 12 OPERABLE safety/

relief. valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

of the safety/relief valve lift settings will occur only

'emonstration during shutdown and will be performed in accordance with the provisions of

~pecification 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3. 1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

WASHINGTON NUCLEAR - UNIT 2 B 3/4 4-1 SEP 2 2 >983

An assessment has been performed on'the impact of single recirculation loop operation upon plant safety. Results show that single-loop operation is permitted providing the Fuel Cladding Safety Limit is increased, as noted by Specification 2.1.2, APRM scram, rod block and RHM setpoints are adjusted as noted in Specification 3.2.2, Table 2.2.l-l and Table 3.3.6.2; MAPLGHR limits are decreased by the factor given in Specification 3.2.1 and surveillance on the volumetric flow rate of the operating recirculation loop is imposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below (30%) Thermal Power or (50K) rated recirculation loop flow, is to mitigate the undue thermal stress on vessel nozzles, recirculation pump, and vessel bottom head.

IBSEN'T 7 In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

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