Text

Proposed Tech Specs Re Power Uprate W/Increased Flow
	ML20058F828
Person / Time
Site:	Susquehanna
Issue date:	11/24/1993
From:	; PENNSYLVANIA POWER & LIGHT CO.
To:	;
Shared Package
ML17158A030	List: ML20058F821; ML20058F828; PLA-4055, Forwards Proposed Amend 117 to License NPF-22 Re Power Uprate W/Increased Flow;
References
	NUDOCS 9312090007
	Download: ML20058F828 (41)
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ATTACHMENT 2 TO PLA-4055 '

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r-SSES UNIT 2 TECHNICAL SPECIFICATIONS CHANGES' i

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PDR ADDCK 05000388 4

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.a ATTACHMENT 2 TO PLA-4055 2

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2.

Accordingly; for the Facility Operating License No. NPF-22, paragraph 2.,C.(1) is hereby amended to read as follows:

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(1) Maximum Power Level e

d Pennsylvania Power & Light Company (PP&) is authorized to at 0

the facility at reactor core power levels not in excess of ega-watts thermal (100!! power) in accordance with the conditions specified

i herein and in Attachment I to this license.

The preoperational tests, h

startup tests and other items identified in Attachment 1 to this license 9l shall be completed as specified. Attachment 1 is hereby incorporated into this license.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

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Darrell G. Eisenhut,' Director 6

-:: cp, Division of Licensing Office of Nuclear Reactor Regulation Date of Issuance: dE FRoM S5e5 u to a-I OPseAnN(.n Uceu se NO. NPF-rz.

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ATT ACHMENT 2 TO PLA-4055 r

i DEFINITIONS e

RATED THERM L POWER L

i L33 RATED THEREL POWER s a total reactor core heat transfer rate to the reactor coolant o REACTOR PROTECTION SYSTEN RESPONSE TIME L34 REACTOR PROTECTION SYSTEM RESPONSE TILE shall be the time intervai from 3

when the monitored parameter exceeds its trip setpoint at the channel l

sensor until doenergization of the scram pilot valve solenoids. The

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response time may be esasured by any series of sequential, overlapping i

or total steps such that the entire response time is asasured.

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REPORTABLE EVENT L35 A REPORTABLE EVENT shali be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

R00 DENSITY L36 ROD DENSITY shall be the number of control rod notches inserted as a fraction of tha total number of contrc,1 rod notches. All rods fully inserted is. equivalent to 100% R00 DENSITY.'

SECONDARY C0KfAIMMEN1 INTEGRITY L37 SECONDARY CONTADMENT INTEGRITY shall exist when:,

All secondary containment penetrations required to be closed during a.

accident conditions are either:

L Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or

. deactivated automatic damper secured in its closed posit kn, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b.

All secondary coritainment. hatches and blowout panels are cidted and sealed.

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c.

The standby gas treatment systas is OPERABLE pursuant to Specification 3.6.5.3.

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At least one door in each access to the secondary containment is closed.

The saaling mechanism associated with each secondary containment e.

penetration, e.g., welds, bellows, resilient matarial saals, or 0-rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or equal m

to the value required by Specification 4.6.5.la.

SUSQUEHANNA - UNIT 2 1-6 I

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ATTACHMENT 2 TO PLA-4055 g

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2. 0 SAFETY LIMITS AND LIMITIE SAFETY SYSTEM SETTINGS a

L 2.1 SAFETY LIMITS THERMAL POWER. Lew Pressure or Low Flow 2.1.1 THERPEL POWER shall not exceed 25% of RATED THERMAL POWER with the gtor vessel steem done pressure less than 785 psig or core flow less than G" ef :td "!ewp APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTIDW:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel staes dome pressure less than 785 psig or core flow less thanC" :f r;te :q be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

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i THERMAL POWER. Hich Pressure and Hich Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06*

I with the reactor vessel steem done pressure greater n 785 s and core flow greater than@ J rea:y go m; l;o, gm/s.,

APPLICABILITY: OPERATIONAL CONDITIONS-1 and 2.

ACTION:

With MCPR less than 1.06" and the reactor vessel steam done pressure greater l

than 785 psig and core flow greater thanGC% cf reted fQ be in at least HOT i

SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specific on 6.7.1.

10 milliow bh REACTOR COOLANT SYSTEM PRESSURE 2.1. 3 The reactor coolant system pressure, as measured in the reactor vessel steam done, shall not axceed 1325 psig.

APPLICABILITY: OPERATIONAL COMalTION$ 1, 2, 3 and 4.

ACTION:

l With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant systes pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

See Specification 3.4.1.1.2.a for single loop operation requirement.

l SUSQUEHANNA - UNIT 2 2-1 Amendment No. 26 s

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ATTACHMENT 2 TO PLA-4055

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2-4 Amendment No. 35 i

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2.1 SAFETY LIMfTS ATTACHMENT 2 TO PLA-4055

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2.0 INTRODUCTION

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 estabhshed to protect the integrity of these barriers during normal plant operations and anticipated 1

transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violstad. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than the limit specified in Specification 2.1.2 for SNP fuel MCPR greater than the l

2 specified limit represents a conservative margin relative to the conditions required to y

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 bemer is related to its relative freedom from perforations or craciong. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementatty cumulative and continuously measurable. Fust cladding perforations, however, can result from thermal stresses which occut from reactor operation significantly above design conditions and the Umiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermafiy caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore. the fuel cladding Safety Limit is defined with a rnargin to the conditions which I

would produce onset of transstion boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity Safety Limit assures that during nctmal operation and during '

j anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling (ref. XN-NF 524(A) Revision 11.

2.1.1 THERM A1. POWER. Low Pressure or low Flow The use of the XN 3 correlation is valid for critical power calculations at pressure greater 8

than 580 psig and bundle mass fluxes greater than 0.25 x 10' lbs./hr ft. Fw operation at low pressures or low flows, the fuel cladding integrity Safety Limit is established by a limiting condition on core THERMAL POWER with the following basis:

Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to assure a minimum bundle flow for all fuel i

assembt;es which havs a rs!stivcty high power and potentis!!y can approach a critics! hast flux condction. For the SNP 9 x 9 fuel design, the mirwnum bundle flow is greater than 30,000 lbs/hr. For the SNP 9 x 9 design, the coolant minimum flow and maximum flow area 8

is such that the mass flux is always greater than 0.25 x 10' Ibs/hr ft. Full scale crmcal power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical 8

power at 0.25 x 10' lbs/hr ft is 3.35 Mwt or greater. At 25% thermal power a bundle power of 3.35 Mwt corresponds to a bundle radial peaking f actorgffreetoshe3 0 which is significantly higher than the expected peaking f actor. Thus, alt 4ERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.

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a SUSOUEHANNA - UNIT 2 B 2-1 Amendment No. 91,

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6 LfMITING SAFETY SYSTEM SETTING revisions -dfugggefe),2

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2.2. l.1 O.

BASES

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REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 9.

Turbine Ston Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux Inge,_-t L sEQa increases that would result from closure of the stop valves. With a trip setting of 5.5%

b of valve closure from full open, the resultant increase in heat flux is such that adequate thermal margins are maintained during the worst case transient assuming the turbine bypass valves operate.

10. Turbine Control Vatve Fast Closure. Trio Oil Pressure-tow The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failure of the turbine bypass valves. The Reactor P~)tection System initiates a trip when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 mi!!iseconds after the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in rapidly reducing hydraulic trip oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure switches whose contacts M-form the one-out-of-two-twice logic input to the Reactor Protection System. This trip e,, setting, a faster closure time. and a different valve characteristic from that of the p M M b 3"'fh 6 turbine stop valve, combine to produce transients which are very similar to that for the stop valve. Relevant transient analyses are discussed in Section 15.2 of the Foal

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Safety Analysis Report.

11. Resetor Mode Switch Shutdown Position The reactor mode switch Shutdown position is a redundant channel to the automatic protective instrumentation channels and provides additional manual reactor trip capability.

12. Manual Scram The Manual Scram is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

1 SUSOUEHANNA - UNIT 2 B27 Amendment No.E w

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ATTACHMENT 2 TO PLA-4055 lf 3-INSERT TO BASIS 2.2.1.9 l)?

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This function is not required when THERMAL POWER is below 30% of i/

RATED THERMAL POWER.

The Turbine Bypass System is sufficient at

'3 this low power to accommodate a turbine stop valve closure A,

witnout the necessity of shutting down the reactor.

This

p function is automatically bypassed at turbine first stage ~

pressures less than the analytical limit of 147.7 psig, equivalent to THERMAL POWER of about 30% RATED THERMAL POWER.

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Turbine first stage pressure of 147.7 psig is equivalent to 22%

1, of rated turbine load.

INSERT TO BASIS 2.2.1.10 (PARAGRAPH "B")

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This function is not required when THERMAL POWER is below 30% of

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RATED THERMAL POWER.

The Turbine Bypass System is sufficient at this low power to accommodate a turbine control valve closure i

without the necessity of shutting down the reactor.

This function is automatically bypassed at turbine first stage pressures less than the analytical limit of 147.7 psig,.

l equivalent to THERMAL POWER of about 30% RATED THERMAL POWER.

l Turbine first stage pressure of 147.7 psig is equivalent to 22%

of rated turbine load.

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ATTACHMENT 2 TO PLA-4055 m.

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y REACTIVITY CONTROL SYSTDtS

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3 SURVEILLANCE REQUIREMENTS (Continued 1 Q

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b.

At least once per 31 dqys by; 3

1.

Verifying the continuity of the explosive charge.

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Determining that the available weight of sodium pentaborate.is

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greater then or equal to 5500 lbs and the concentration of boron 3l g

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in solution is within the limits of Figure 3.1.5-2 by chemical.

h analysis."

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Verifying that each valve, manuel, power operated or automatic,

!l in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

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Demonstrating that, when tested pursuant to Specification 4.0.5, the

.~1 minimum f1 14 irement of 41.2 gym at a pressure of greater than or equal _ to psig is set.

d.

At least once per 28 months during shutdown by; 6,

1.

Initiating one of the standry liquid control system loops, 1[

including an explosive valve, and verifying that a flow hath-from the pumps to the reactor pressure vessel is available by 1'

pumping domineralized water into the reactor vessel. The replace-1 ment charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired.

Both injection loops shall be tested in 36 months.

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- Demonstrating that all heat traced piping-is unblocked by pumping from the storage tank to the test tank and then draining and flushing the discharge piping and test tank with c) damineralized water.

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Demonstrating that the storage tank heaters are OPERA 8LE by verifying the expected temperature rise for the sodium 1mtaborate solution in the storage tank after the heaters are energized.

1 2

"This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the lief t of Figure 3.1.5-1.

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- This test shall also be performed whenever both heat tracing circuits have.

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been found to be inoperable and may be performed by any series of sequential,.

overlapping or total flow path steps such that the entire flow path is 4

included.

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Am m 2 m m 055 POWER DISTRIBtm0N LIMITS 314.2.2 APRM SETPOINTS LIMmNG CONDmON FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (S g) shall be established according g

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o the following relationships:

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NO CHAUGe (d0CMAMG6 TRIP SETPOINT ALLOWABLE VALUE #

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S s (0.58W + 59%) T S s (0.58W + 62%) T f Spg s (0.58W + 50%) T.

SRB s (0.58W + 53%)

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where: S and SRB are in percent of RATED THERMAL POWER, Loop recirculation flow as a percentage of the loop recirculation flow which W

produces @re flow of 100 mi!! ion Ibs/hr, Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (FRTP)

T

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divided by the MAXIMUM FRACTION OF LIMmNG POWER DENSITY. The FRACTION OF LIMmNG POWER DENSITY (FLPD) for SNP fuelis the actual LHGR divided by the LINEAR HEAT GENERATION RATE for APRM Setpoints limit specified in the CORE OPERATING LIMITS REPORT.

i T is always less than or equal to 1.0.

APPLICABILITY: OPERATIONAL CONDmON 1, when THERMAL POWER is greater than nr equal to 25% of RATED THERMAL POWER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB, as determined above, initiate corrective action within 15 minutes and adjust S and/ or Sag to be consistent witn the Trip Setpoint value within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as recuired:

With MFLPD greater than the FRTP ouring power ascension up to 90% of RATED THERMAL POWER, rather than adjustinD the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10% of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

See Specification 3.4.1.1.2.a for single loop operation requirements.

SUSOUEHANNA - UNIT 2 3/4 2-2 knend:nerit No. 9I, 95 O

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TABLE 3.3.1-1 (Continued) v 9

REACTOR PROTECTION SYSTEM INSTRLMDf7ATION i

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ACTION STATEMENTS

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d ACTION 1 Se in at least NOT SHUTDOW within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

_9 ACTION 2 i

Verify all insertable control rods to be inserted,in the core and lock the reactor mode switch in the Shutdown position

- f within I hour.

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Suspend all operations involving CORE ALTERATIONS and insert all insertable control rods within I hour.

ACTION 4 Se in at least STARTUP within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

i' ACTION 5 Se in STARTUP with the main steam line isolation valves closed 4

within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months or in at least NOT SHUTDOW within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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i ACTION 7 * -

Verify all insertable control rods to be inserted within 1 hou'r.,

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ACTION 8 Lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 9 Suspend all operations involving CDRE ALTERATI0MS, and insert all insertable control rods and lock the reactor mode switch in the SHUTDOWN position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

P Litiate. a. ve A d on :n-i w;%m.W wnotes, awa vTuseMAs._ vower e duc.e. TRER H A L power +o ieu %nw 30% of R ATG D TH ER M A L powp wWn 2 houe.

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rn9-m TABLE 3.3.1-1 (WATTACHMENT 2 TO PLA-4055 J

HACTOR PROTECTTON SYSTEM INSTRUMENTATION f

TABLE NOTATIONS

.1 (a) A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the 9

same trip synem is monitoring that parameter. Upon determination that a trip setpoint cannot be restored to within its specified value during performance of the CHANNEL CALIBRATION, the j

appropriate ACTION,3.3.la or 3.3.lb, shall be followed.

(b) This function is automatically bypassed when the reactor mode switch is in the Run position.

(c) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any I

control rod is withdrawn

and shutdown margin de.:nonstrations performed per Specification 3.10.3.

(d) The non-coincident NMS reactor trip function logic is such that all channels go to bo6 trip systems.

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nerefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMS and 6 IRMS.

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6 (e) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

(f) His function is not reviired to be OPERABLE when the reactor pressure vessel head is unbolted or removed per SpeciScation 3.10.1.

(g) His function is automatically bypased when the reactor mode switch is not in the Run position.

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(h) His functmo is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.

(i) With any control rod withdrawn.*

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equivalente-THERMAL-POWER of et.ee 24Wof-RATED-THERMAL POWER.-

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Not required for coetrol rods removed per Specificanon 3.9.10.1 or 3.9.10.2.

I SUSQUEHANNA - UNIT 2 3/43-S Amendment No.84 l

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ISOLATION ACTUATION INSTilUMENTATION SETPOINTS Trip FUNCTX)N TIup SETPOWT ALLOWABLE VALUE 9>

MAIN Eit AM LINE EQLAil0N fcon*me Z

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Condoneer Veeuum. Law a 9.0 inches Hg vacuum a 8.8 inchee He vacuum f.

Reector Buildine Make Steam the Tunnel Temperstwo. ##gh 5 177'F s 184*F g.

Reacter thAhens Mehi Stoem Une Twinal & Temperstwo. High s 99'F s 10g'F h.

Manuel Weine6en HA NA L TwWne Beeng Meh Steens uns Tunned Tennpereewe legh 5 197'F s 200'F

4. MACTOR WATER plEANUP EYETEM EDLATION e.

RWCU & Flow ifgh 5 90 gem s 80 girn t.

NWCU Aree Tempersano. High s 147' F er 131*F 5154'F er 137'F I

e.

RWCUfAree Veneeselors A Teenpereews. itsh s es'F er 40.5'F s 72'F er 43.5'F E

d.

SLCS Weteeken NA NA e.

React = veeees ween Levet. Lew Lew, tevel 2

=.se Irwe=e%,

a.45 inche [

(1. MWCU Flow tech hb2 )

)h'72 i f 2. L.

. _selve Heat Excheneer Deecherse Teenporeews -

s 144'F s 150'P High 3 Manusi Helsetars NA NA d

m

[ bel

/g

5. HEACTOR CQfE ISQLATION C00LMG SYSTEM ISOLATIDN e.

ROC Steam Une a Preseus. Sieh M H,0 s

H,0 g

{

[

h.

Roc steem se Preeeure. Low a 80 pele

. = 53 pese o

3e c.

ROC TwWne Emhouet th Pressure. High s 10 0 pois s 20.0 pele

,3 3

9 l

z

These arty funeelene rood not be OPERA 8LE fran October 19,1989 to January 19,1990 g

?

w-e 1

l l

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3

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_l Allil 3. 3. 2-2 (Continued) b 150lAll0N ACTUAll0N IN51RUHlNIAll0N SLIPOINIS Ni x.

2 ALLOWA8LE E

IRIP iUNCIION IRIP $[IPOINI VALUE RfACIOR CORE 150LAll0N COOLING SY511H ISOLAll0N (Continued)

EZ d.

RCIC Equipment Room lemperature - liigh

< 161'r

$ 174'I

~

e.

RCIC Equipment Room i

o lemperature - liigh

$ 89'T

$ 98'I

l f.

RCIC Pipe Routing Area Temperature - liigh

$ 167'i##

$ 1/4'I##

g.

RCIC Pipe Routing Area a lemperature - liigh

< 89*f##

$ 98*i##*

l h.

RCIC leergency Area R

Cooler Temperature - liigh

< 147"f

< 154*f i.

Manual Initiation NA NA j.

Drywell Pressure - liigh

$ 1.72 psig

$ 1.88 psig 6.

IllGil PRES 5URE C00LANI INJECll0N SYSIIM 150 tall 0N gg 3gg a.

ilPCI Steam Line i low - liigh inclus N 0 nches !! 0

~

2 b.

IIPCI Steam Steply Pressure - Low

> 04 psig IIPCI Turbine Exhaust Diaphragm

- 90 psig c.

Pressure - liigh

< 10 psig

~< 20 psig d.

IIPCI [quipment Room lemperature - liigh

< 161*f

< 174'r e.

IIPCI Equipment Room g

l

[RQ A Iemperature - liigh

< 89'T

< 98'T

5

[g**

f.

IIPCI Emergency Area Cooler gh Temperature - liigh

< 14/*f

< 154*f

[

', g *7 g.

IIPCI Pipe Routing Area

~

o Iemperature - liigh

< 167'If#

< 174'ffs

[

i*R h.

IIPCI Pipe Routing Area A

l Gg a lemperature - liigh e 89'l##

< 98'I##*

l 2 3" i.

Manual' Initiation MA NA J.

Drywell Pressure - liigh

' l./2 psig

< l.811 psig

lhese trip f unctions necil not be Ul'IRAllit f rom Octuher 19 1989 to January.19. 1990.

g.

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TABLE 3.3.2 2 (Continued)

C h

ISOLATION ACTUATION INSTRUMENTATION SETPOINTS M

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 7.

RHR SYSTEM SHUTDOWN COOLING / HEAD SPRAY MODE ISOLATION a.

Reactor Vessel Water Level - Low, level 3 2 13.0 inches

2 11.5 inches b.

Reactor Vessel (RHR Cut-in Permissivel Pressure. High s 98 psig s 108 psig c.

RHR Flow. High s 25,000 ppm 5 20,000 gpm

'f8 d.

Manual ini;iation NA NA MO e.

Drywell Pressure. High 5 1.72 psig 5 1.88 psig j#

See Bases Figure B 3/4 31.

Lower setpoints for 75H-G33 2N000 E. F arxi TDSH G33 2N602 E, F.

1. 15 minute time delay.

y L:+;a.1 valuc, L al ya\\ue.+.be.

4.+e. m twe.A base.d 5

56.\\1 be submdre &guti.e j

hb. Awy re.

4 on b e.

u %+e sb+=7 e

3 to N.

g thage. tv %ts vake.

Comwissiow wlWa 90 days of %+ ampidion.

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ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

O Cm CHANNEL CHANNEL CHANNEL OPERATIONAL CONDtTIONS FOR -

I TRIP FUNCTION CitECK FUNCTIONAL Call 8 RATION WHtCH SURVEILLANCE REQUIRED D

TEST

^Z>

HIGH Pf1 ESSURE COOLANT INJECTION SYSTEM 1500ATION NA.

M O

1,2,3 C

IContmued) d.

HPCI Equipment Room Tempersture - liigh NA M

O 1,2,3 HPCI Equipment Room a Temperature.

M e.

liigh NA M

O 1, 2, 3

f. HPCI Emergency Area Cooler Temperature. High HA M

O 1,2,3, HPCI Pipe Routing Area Tempeteture.

g.

High HA M

Q 1,2,3 h.

HPCI Pipe Routing Area a Tempef atore -

High e

NA R

NA 1, 2, 3 Y

1. MenuelInitiation N

NA M

R 1, 2, 3

l. Drywen Pressure High RHR SYSTEM SHUTDOWN COOLING / HEAD 7.

SPRAY MODE ISOLATION Reector Vessel Water Level Low, level S

M R

1,2,3 3

e.

3 Reector Vessel (RHR Cut in Permissivel NA M

O 1,2,3 2

b.

!C Pressure - High S

M R

1, 2, 3 y

RHR Flow High c.

MA R

NA 1,2,3

[

f d.

Manuel Initiation o

NA M

R 1,2,3-y o

Drywes Pressure - High e.

g g

When hondung Irradieted fuelin the secondary conteinment and during CORE ALTERATIONS and operations with a potentie g

m e

vosoof.

j

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= 4 __

-?,

wp turtene stop voeve is e,en.

When VENT 500 or PURGING the drywell por Specificetion 3.11.2.8.

This trip function rwed not be OPERABLE from October 19,1999 to January 19,1990.

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a g

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M, ATTACHMENT 2 TO PLA-4055

M.

e

'i INSERT TO TABLE 3.3.6-2 r

4 ij' 9, 4 1.

ROD BLOCK MONITOR lQ n,,

a.-Upscale ##

5 0.63 W + 41%

.s 0.63 W + 43%

&c; 3

2.

APRM i

H 5

a. Flow Biased T

Neutron Flux c.

3 d

Upscale ##

5*

w

-[

1) Flow Biased 5 0.58 W + 50%F

$ 0.58 W + S3%)D, l

2) High Flow

$ 108% of RATED 5 111% of RATE e

Clamped THERMAL POWER THERMAL POWER i

\\

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J 1

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ATTACHMENT 2 TO PLA-4055 e.

'A.

REACTOR COOLANT SYSTEM X

SURVEILLANCE REQUIREMENTS

.s 0

ACTION: (Contirued) 2.

If Region II of Figure 3.4.1.1.1-1 is entered and greater than 4

or equal to 50% of the required LPRM upscale alarms OPERABLE,

]

immediately exit the region by:

5 3

a) inserting a predetermined set of high worth control rods.

or b) increasing core flow.

3.

With less than 50% of the required LPRM upscale alarms OPERABLE, follow ACTION a.1.d upon entry into Region II of Figure 3. 4.1.1.1-1.

In OPERATIONAL CONDITION 2 with no reactor coolant system b.

recirculation loops in operadon, return at least one reactor coolant system recirculation icop to operation, or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With any pump discharge valve not OPERABLE remove the associated c.

loop from operation, close the ' valve and comply with the requirements of Specification 3.4.1.1.2.

d.

With any pump discharge bypass valve not OPERABLE close the valve and verify closed at least once per 31 days.

4.4.1.1.1.1 Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling eacn valve through at least one complete cycle of full travel during each startup** prior t TH RMA POW R ce c' of RATED ofIO t S W U " (in THERMAL POWER.

a core -flou

/b 4.4.1.1.1.2 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with oversoeeo setpoints less than or equal to TT2..=

r and respecti vely, (kA-rata" c+-e *!:42 at least once per 18 months.

4.4.1.1.1.3 At least 50% of the required LPRM upscale alarms shall be determined OPERABLE by performance of the following on each LPRM upscale a la rm:

1)

CHANNEL FUNCTIONAL TEST at least once per 92 days, and 2)

CHANNEL CALIBRATION at least once per 134 days, tio.s 4 h on bhr

- If not performed within the previous 31 days.

SUSQUEHANNA - UNIT 2 3/4 4-la Amendment No. 60 O.

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as so as so os to 30 as ao Core Flow (% RATED)

J Figure 3.4.1.1.1-1 THERMAL POWER RESTRICTIONS i

f i

n 1

3/4 4-lb h h t No. 91 OCT 2 81992 4,

4 I

i

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1 4,

ATTACHMEN12 TO PLA-4055

.f M.

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0, 3D 36 40 46 60 65 60 65 70 i

Core Flow (Million Ibm /hr)

Figure 3.4.1.1.1-1 i

THERMAL POWER STABILITY RESTRICTIONS 1

1' a

k s

l s

' a4 I

i a

5 y.

REACTOR COOLANT SYSTEM ATTACHMENT 2 TO PLA-4055 RECIRCULATION LOOPS - SINGLE LOOP OPERATION 2

O LIMITING CONDITION FOR OPERATION 1

gU

()(

h.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed s; B0%

p of the rated pump speed and the reactor at a THERMAL POWER / core flow condition 2

outside of Regions I and ll of Figure 3.4.1.1.1-1, and U.)

~

j

@N k

D

a. the following revised specification limits shall be followed:

sy p

a VT"j 1.

Specification 2.1.2: the MCPR Safety Limit shall be increased to 1.07.

d. tdh<

U

'W 2.

Table 2.2.1-1: the APRM Flow-Biased Scram Trip Setpoints shall be as folinwn:

1 07J l

CMo c#AUGC h g.J xl Trip Setpoint Alowable Value 9h

[0.58W + 54%

W T--

? *- s

~

T 3.

Specification 3.2.2: the APRM Setpoints shall be as follows: __

A

$Mf (No CHAMGG e

~

pJ O

w S taa Anow

v*.

v p- (0.58W + 57%) T ll!

S s (0.5BW + 54%) T Ss y ]

.I %

y Spa s 10.5BW + 45%) T Spg s (0.58W + 48%) T N W e&

C 4.

Specification 3.2.3 tMUM CRITICAL POWER RATIO (MCPR) shall be

( tv) o [g p

greater than or equal to the applicable Single Loop Operation MCPR limit as Z

specified in the CORE OPERATING LIMITS REPORT.

l o@@

/

g able 3.3.6-2: the RBM/APRM Control Rod Block Setpoints shall be as follows:

pgg-.

u Trip Setpoint ABowable Value t s.

d

.v

a. RBM - Upscale sht" 25 Q s6.SSW > ?99 #

U M~

V "Q g d LU Trip Setpoint Allowable Value I

b.

APRM - Flow s

APPLICABILITY:

OPERATIONAL CONDITIONS 1

and 2' +, exc uring two loop operation.#

ACTION:

,(33@ SS%

,(34 37%

e.

In OPERATIONAL CONDITION 1:

a) no reactor coolant system recirculation loops in operation, or b) Region I of Figure 3.4.1.1.1 1 entered, or c) Region 11 of Figure 3.4.1.1.1-1 entered and core thermal hydraulic instability occurring as evidenced by:

SUSQUEHANNA - UNIT 2 3/4 4-1c Amend'rnent No.9h 95 ud.

.h i

e

ATTACHMENT 2 TO PLA-4055 5*

~

REACTOR COOLANT SYSTEM

^

LIMITING CONDITION FOR OPERATION (Continued) f.

With any pump discharge bypass valve not OPERABLE close the valve

+

and verify closed at least once per 31 days.

SURVEILLANCE REQUIREMENTS 4.4.1.1.2.1 Upon entering single loop operation and at least once per p~

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, verify that the pump speed in the operating loop is 1 80% of the rated pump speed.

4.4.1.1.2.2 At least 50% of the required LPRM upscale alarms shall be

'l determined OPERABLE by performance of the following on each LPRM upscale alarm.

1)

CHANNEL FUNCTIONAL TEST at least once per 92 days, and 2)

CHANNEL CALIBRATION at least once per 184 days.

4.4.1.1.2.3 Within 15 minutes prior to either THERMAL POWER increase resulting from a control rod withdrawal or recirculation loop-flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 30%**** of RATED THERMAL POWER or the recirculation loop fTow in the operating recirculation loop is 150%**** of rated loop flow:

a.

< 145'F between reactor vessel steam space coolant and bottom head drain line coolant, b.N < 50*F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. # 150'F between the reactor coolant within the loop not in operation and operating loop.

.1 The pump discharge valve and bypass valve in both loops shall a core -Eloo e-9 (095 v.;thw b/h* pe demonstrated OPERABLE by cycling each valve through at leas omplete cycle of full travel during each startup** prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER.

4.4.1.1.2.5 I The pump MG set scoop tube electrical and mechanical stops shall j

deegts g 0PERABLE with overspeed setpoints less than or equal to g and spectively, d' n Md cer: '% at least once per 18 mon g;

4.4.1.1.2.6 During single recirculation loop operation, all jet pumps, including those in the inoperable loop, shall be demonstrated l'

OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that no two of the following conditions occur:MiB a.

The indicated recirculation loop flow in the operating loop differs by more than 10% from the established single recirculation pump speed-loop flow characteristics.

i SUSQUEHANNA - UNIT 2 3/4 4-le Amendment No.76 a

R

>*I ftEACTOR COOLANT SYSTEM ATTACHMENT 2 TO PLA-4055

)

~

q SURVEILLANCE REOUIREMENTS (Continued) r J

b.

The indicated total core flow differs by more than 10% from the l'

established total core flow value from single recirculation loop flow measurements.

c.

The indicated diffuser-to-lower plenum differential pressure of any 2

individual jet pump differs from established single recircuistion loop I.

pattems by more than 10%.

a 4.4.1.1.2.7 The SURVEILLANCE REQUIREMENTS associated with the specifications

.i; referenced in 3.4.1.1.2a shall be followed.

See Special Test Exception 3.10.4.

' f not performed within the previous 31 days.

Powe r-ldr ed e-i initial value. Final value to be determined based on startup testing. Any required change to this value shall be submitted to the Commission within 90 days of completion.

See Specification 3.4.1.1.1 for two loop operation requirements.

1. This requirement does not apply when the loop not in operation is isolated from the reactor pressure vessel.

1. 1. At least once per 18 months (555 days), data shall be recorded for the parameters f

listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed shall commence upon the performance of required surveillances.

l

+ The LPRM upscale alarms are not required to be OPERABLE to meet this specification in OPERATIONAL CONDITION 2.

%c.c LApe-okt. skadup Test y,-oyox l

\\

a SUSOUEHANNA - UNIT 2 3/4 4 1f Amendment No.E m

O

.<k

44. -

ATTACHMENT 2 TO PLA-4055 74, REACTOR COOLANT SYSTEM

).g

' i RECIRCULATION PUMPS LIMITING CONDITION FOR OPERATION 9

24

}

m 3.4.1.3 Recirculation pump speed mismatch shall be maintained within:

e a.

if ntec ::-- eQ" 7 75% of each other_ with core flow greater than or equal to M

z

~T5milieg%m[by 3

b.

f each other with core flow less tnan @ :f ntec care AP9LICABILITY:

OPERATIONAL CONDITIONS 1* and 2* when both recirculation loops are in operation.

ACTION:

~

With the recirculation pump speeds different by more than the specified limits, eitner:

i Restore the recirculation pump speeds to within the specified limit a.

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or b.

Declare the recirculation loop of the pump with the slower speed not in. operation and take the ACTION required by Specification 3.4.1.1.1.

~

i SURVEILtANCE REQUIREMENTS l

4.4.1.3 Recirculation pump speed mismatch shall be verified to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~

"See Special Test Exception 3.10.4 i

l SUSQUEHANNA - UNIT 2 3/4 4-3 Amenoment No. 50 AUG 30 588

a a

ATTACHMENT 2 TO PLA-4055 c,

REACTOR COOLANT SYSTEN 4

O

- t 6

3/4.4.2 SAFETY / RELIEF VALVES LIMITING COWITION FOR OPERATION 3.4.2 The safety valve function of at leasth the following reactor coolant i

system safety /rsifer valves shall be OPERABLE with the specified code safety valve function lift settings:" **

l 6 ::*:ty-M!%' W:: 011" A G) 17-l

@ safety-reiter valves 91175 psig f3 l

T:tr :14:* :?::: ?'~mem2 h+G safety-roller valves 5 1195 psig 1 3 safety-relief valves 8 1205 psig + N 8

i APPLICA8ILITY: OPERATIONAL CONDITIONS-1, 2, and 3.

l ACTION:

With the safety valve function of one or more of the above required a.

e' safety / relief valves inoperable, be in at least HDT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b.

With one or more safety / relief valves stuck open, providst that sup-pression pool average water tagwrature is less than 105'F close the stuck open relief valve (s); if unable to close the open valve (s) within 2 minutes or if suppression pool water temperatere is 105'F ^

or greater, place the reactor mode switch in the Shutdown position.

c.

With one or more safety / relief valve acoustic monitors inoperable,.

{

restore the inoperable monitor (s) to.0PERABLE status within 7 days i

or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i SURVEILLANCE REQUIRDENTS 4.4.2 The acoustic monitor for each safety / relief valve shall be demonstrated OPERA 8LE with the setpoint verified to be 0.25 of the full open noise level ('by performance of a:

a.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and a

~

b.

Calibration in accordance with procedures prepared in conjunc+1on with its manufacturer's recommendations at least once per 18 months.N l-

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

- Up to 2 inoperable valves may be replaced with spare OPERA 8LE valves with lower setpoints until the next refueling.

Initial setting shall be in accordance with the manufacturer's recommendation.

Adjustment to the valve full open noise level shall be accomplished during the startup test program.

NThe provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perform the test.

SUSQUEHANNA - tafIT 2 3/4 4-5 o

)

1

,r.g

+.

ATTACHMENT 2 TO PLA-4055 j.

4 I

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION E

3.4.3.2 Reactor coolant systen leakage shall be limited to:

a.

No PRESSURE BOUWARY LEAKAGI.

b.

S gpa UNIDENTIFIED LEAKAGE.

1035' c.

25 gpa total leakage averaged over any 24-hour period.

1 gpa leakage at a reactor coolant systes pressure of@lve spec d.

t 10 psig from any reactor coolant system pressure isolation va in Table 3.4.3.2-1.

~

e.

2 gpa increase in UNIDENTIFIED LEAKAGE within any 4-hour period.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN withth the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

With any reactor coolant systes leakage greater than ttp liatts in b.

and/or c., above, reduce the leakage rate to within the liatts within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c.

With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual or deactivated automatic valve, or be in at least HOT SHUTDOWN within the.next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d.

With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperable monitor (s) to OPERABLE _ status within 7 days or verify the pressure to be less than the alarm pressure at Isast once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; restore the inoperable monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e.

With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than 2 gpa within any 4-hour period, identify the source of leakage increase as not service sensitive Type 304 or 316 austani+.ic stainless steel within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHuidWN withir +he next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 ~

}

SUSQUEHANNA - UNIT 2 3/4 4-7

2 1

^:

M.

ATTACHMENT 2 TO PLA-4055 21 i

i.I REACTOR C0OLANT SYSTEM l

i

~k REACTOR STEAM 00E LIMITING CONDITION FOR OPERATION l

l 3.4.6.2 The pressure in the reactor steam done shall be'1ess than psig.

j APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

1050 J

ACTION:

\\0E0 With the rea steam done pressure exceedi 3949-psig, reduce the pressure to less than 1^^

ysig within 15 minutes or be in at least HDT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1050 t

i i

SURVEILLANCE REQUIREMENTS 4.4.6.

The reactor steam dome pressure shall be verified to be less l

than psig at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

1050 f

i 2

1 i

Not applicable during anticipated transients.

l l'

k a

SUSQUEHANNA - UNIT 2 3/4 4-20 M

^

uU.

a. '

ATTACHMENT 2 TO PLA-4055 EMERGENCY CORE COOLING SYSTEMS 7

SURVEILLANCE REQUIREMENTS t

4.5.1 The emergency core cooling systass shall be demonstrated OPERA 8LE by:

l 3;

~,

a.

At least once per 31 days:

g 1.

3 For the CSS, the LPCI systas, and the HPCI system:

a)

Verifying that the systes piping from the pump discharge

!]

valve to the systes isolation valve is filled with water by:

1.. Venting at the high point vents 2.

y Performing a CHANNEL FUNCTIONAL TEST of the condensate transfer pump discharge low pressure alarm instrumentation.

b)

Verifing that each valve, manual, power-operated, or automa-tic, in the flow path that is not locked, sealed, or other-wise secured in position, is in its correct ** position.

J 2.

For the CSS, performance of a CHANNEL FUNCTIONAL TEST of the core spray header AP instrumentation.

't i

3.

For the LPCI system, verifying that at least one LPCI system subsystem cross-tie valve is closed with power removed from the.

valve operator.

4.

ror the HPCI systas, verifying that the pump flow controller is in the correct position.

b.

Verifying that, when tested pursuant to Specification 4.0.5:

1.

The two CSS pumps in each subsystem together develop a total flow of at least 6350 gpa against a test line pressure of 1 282 psig, corresponding to a reactor vessel steam done pressure of 1105 psig.

2.

Each LPCI pump in each subsystem develops a flow of at least 12,200 gpa against a test line pressure of 1 222 psig, corres-ponding to a reactor vessel to primary containment differentia 1' pressure 120 psid.

gqo 3.

The HPCI pump devel' of at least 5000 gpa against a test a

line pressure of >

sig when staas is being supplied to the turbine at 925, +

, - 20 psig.*

At least once pe'r 18 months:

c.

1.

For the CSS, the LPCI systes, and the HPCI system, per orming a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.

The provisions of Specification 4.0.4 are not applicable provided the surveil-lance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perfor1s the test.

- Except.that an automatic valve capable of automatic return to its ECCS position

~

when an ECCS signal is present say be in position for another mode of operation.

5 e-1 SUSQUEHANNA - UNIT 2 3/4 5-4

[g o

-t-4 M56 4.

INSTRUMENTATION nAsts 3/4.3.4 RECIRCULATION PLMe TRIP ACTUATION INSTRUMENTATION The anticipated transient eithout seres (ATWS) recirculation pump trip systes provides a means of limiting the consequences of the unlikely cccur-rence of a failure to seres during an anticipated transient. The response of the plant to tnis postulated event falls within the envelope of study events in General ElectHc Coassany Topical Report NEDD-10349, dated March 1971 and j

NEDO-24222, dated December 1979.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal sargin which occurs at the end of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor systes at a faster rate than the control rods add ne tion pumps,gative scras reactivity. Each EOC-RPT system tHps both recircula-reducing coolant flow in. order to reduce the void collapse in the core cuMng two of the most limiting petssuritation events.

The twatevents for which @ EOC-RPT protective feature will function are closure of the turbine ste valves and fast closure of the tuttine control valves.

A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT systes; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT systes.

a position switch for each of two turbine stop valves provides input to oneSimilarly, EOC-RPT system; a position switch from each of the other two stop valves

{gg provides input to the other EOC-RPT systes. For each EOC-RPT system, the sensor

[**"S"A tg 'A" relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of i

turbine control valves and a 2 out of-2 logic for the turbine stop valves.

The operation of either logic will actuate the EOC-RPT systes and trip both recirculation pumps.

Each EOC-RPT systes may be manually bypassed by use of a keyswitch which is administrative 1y controlled.

Bypass at less than 30% of RATED THERMAL POWER are annunciate room.

The EOC-RPT rerponse time is the time assueed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e.,

175 as.

Operation with a trip set less conservative than its THp Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Yalue is equal to or less than the drift allowance assumed for each trip in the safety analyses.

4 SUSQUEHANMA - UNIT 2 8 3/4 3-3 v.

I

i ATTACHMENT 2 TO PLA-4055 i,

INSERT TO BASIS 3.4.3.4 (PARAGRAPH "A")

a 3:

This function is not required when THERMAL POWER is below 30% of 6

RATED THERMAL POWER.

The Turbine Bypass System is sufficient at this low power to accommodate a turbine stop valve or control valve closure without the necessity of tripping the reactor recirculation pumps.

This function is automatically bypassed at

.j turbine first stage pressures less than the analytica? limit of 147.7 psig, equivalent to THERMAL POWER of about 30% RATED l

THERMAL POWER. Turbine first stage pressure of 147.7 psig is

3 equivalent to 22% of rated turbine load.

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$,,'T v.c 2ul-3 T REACTOR Vmri TOUGIINRE b

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er Simrting Shot A Mas.

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NI(5)

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N/A l

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Weld N/A 624263/

0.06 0.99

-20 30 N/A

+ 30 l

B204A27A i

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Dese volume ese given enly for the bene 5 of "

she 32 BFFY RTp C

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MON BEL'ILINE # '

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! MATERIAL TYFR OR suma a ras.as IDGHEST STARTW9G COMPONENT WELD SEAM I.D.

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HEAT / Lor

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to Shell Ring #5 5A-533 OR B d.1 AR

+ 10 Bono Head Doom 00462

+20

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Bassaan Head Torue 00472

+ 10 i

Top Head side Flusse 00473-1

+10 Top Head Flange SA 508,0.2 125H446

+ 10 Vessel Please 21.2393

+ 10 t

Q2062W

-10 Feedu m er N ossis l

semen omelet Nosate Q2Q64W

+ 30 h

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Pleases to Shell Top All

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8 Oiher Nam-Belthee All o

9 Closuse Studs SA 540 GR 324, All Meet requim of 45 g

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esponsion et + 10*F O

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SUSQUEHANNA - UNIT 2 3 3/4 4-7 A- =4==t No. 85 -

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ATTACHMENT 2 TO PLA-4055 9

M.

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3/4.5 EMERGENCY CORE COOLING SYSTEM EASES g

4 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN The core spray systes (CSS) is provided to assure that the core is adequately cooled following a loss of-coolant accident and, together,with the LPCI mode of the RHR system, provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the automatic depressurization system (ADS).

The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

t The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated oy recirculation through a test loop during reactor operation, a complete functional test requires reactor shotdown. The pump discharge piping is maintained full to prevent water hammer damage to.

piping and to start cooling at the earliest moment.

The low pressure coolant injection (LPCI) mode of the RHR system 14 provided to assure that the core is adequately cooled following a loss of-coolant accident. Two subsystems, each with two pumps, provide adequate core flooding for all break sizes up to and including the double ended reactor recirculation line break, and for small breaks following depressurization by the ADS.

The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel elad temperature in the event of a small break in the reactor coolant system and loss of coolant i

whii:h does not result in rapid depressurization of the reactor vessel. The HPCI systes permits the reactor to be shut down while maintaining sufficient i

reactor vessel water icvel inventory until the vessel is depressurized. The HCPI system continues to operate until reactor vessel pressure is below the pressure at which CS system operation or LPCI mode of the OfR system operation naintains core cooling.

j i

The capacity of the system is selected to provide the required core i

cooling. The HPCI pump is des ed to deliver greater than or equal to 5000 gpm

'l at reactor pressures between d 150 psig.

Initially, water from the condensate storage tank is uJe nstead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.

//87 SUSQUEHANNA - UNIT 2 B 3/4 5-1 l

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ATTACHMENT 2 TO PLA-4055 j

CNTa!*EW SYS'Em5 SA5t5

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3/a 5.2 DEPRES$URIZaTION SYSTEM 5 The specifications of tais section ensure that the primary containment t

pressure will not escoed the aesten pressure of 13 psig euring primary ststem g

niewee n free full emerating pressure.

gj The suppression chamber water provides the heat sint for the reacierrtN1a systes energy release following a postulated rupture of the systas. Thelsuppee {

n sten enameer water volume must asere the associated escer sad structucatt j

sensitie heat released euring reacter coolant systes sismoown fr-*- OQ ;

1 Since all of the gases in the crywell are purged ints the suopreden eneseer. !

.'i air space euring a less of coolant accleont, the pressure of the lievis must not exceed 53 psig, the s e pressten chammer easteue pressure. The essign volund of the suppression chamber, water and air, uns etsined by consteering that the total volume of reacter ceclant and to to consteered is discharges to the suopression enameer and that the drywell values is purged to the suppression

'y enameer.

Using the minimum er maaimum water volumes given in this specificatten.

containment pressure during the assign basis accident is appremiastaly 45.0 psi !

t whien is below the eesign pressure of 53 psig. Itanteue unter volume of-t

~ :

133.5a0 fts results in a sewacener submergence of 12 feet and the sinlaus volum !

of 122,410 fta results in a summergence approafeately ad inches less. The sajority of the Bodega tests were run with a submerged length ef* 4 feet and wtth complete consensation. Thus, with respect to the eewnceser summergence..

tnis specification is aseeuate. The manieue temperature at the end of the tiew-!

down tested during the MwDeldt Bay and Sedega Say tests mes '170*F and tais is s

conservatively taten to be the 11eit for complete consensatten of the reacter =

coolant, althougn condensatten would occur for temperatures move 170*F.

Should it be necessary to make the suppression chanter inoperele, this shall only to eene as specified in Specification 3.5.3.

Under full pouer operati conditions, bloudoun free an initial suppressier !

cnameer water temperature of results in a meter temperature of appres-instely 128'F immediately following 31oussen which is below the 170*F uses for l

conolete consensatten via T-euencher envices. At this te m erature sad staes-I pneric pressure, the avallele NP5N exceeds that required by Deth the AMR and core spray peps, thus there is no deponeency en containment everseessure eurint ;

the accleont injection phase. If beta RNR leaps are used M OF.iti-mont coellt,

there is ne dependency on containment everpressure for post-LOCA operations.

Emperfeental esta indicate that excessive steam condensing leads can to aveleed if the peak local temperature of the sworessten peel is asintaines below 200*F during any period of relief valve eseratten. Specifications nave, Deen placed en the envelope of reacter operating condttiene se that the reacter can to espressurized in a timely menner ta.aveld the regime of -

potentially high suppression chasser loadings.

m j

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M ATTACHMENT 2 TO PLA-4055

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DESIGN FEATURES 5.3 REACTOR CORE

)

FUEL ASSEM.BLIES N

5.3.1 The reactor core shall contain 764 fuel assemblies. Each assembly consists of 1

6 a matrix of Zircalov clad fuel rods with an initial composition of non-enriched q

or slightly enriched uranium dioxide as fuel material and water rods. Limited substitutions of Zirconium alloy filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel-assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by test or analyses to comply with all fuel safety design bases. A limited number of lead

?

use assemblies that have not completed representative testing may be placed in non-limiting core regions. Each fuel rod shall have a nominal active fuel length of 150 inch'es. Reload fuel shall hrve a maximum average enrichment of 4.0 weight percent U-235.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide powder (B C), and/or Hafnium metal.

The control ro.:i shall have a nominal axial absorber length of 143 inches.

1 Control rod assemblies shall be limited to those control rod designs approved by the NRC for use in BWRs.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2.

1500 psig from the recirculation pump discharge to the jet pumps.

c.

For a temperature of 575'F.

VOLUME y

5.4.2 The total water and steam volume of the reactor vesse! and recirculation system is approximately 22,400 cubic feet at a nominal T ci@* F.

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TABLE 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS SSI E

e CYCLIC OR DESIGH CYCLE h

COMPONENT TRANSIENT LIMIT OR TRANSIENT H

Reactor 120 heatup and cooldown cycles 70*F to h to 70'F 80 step change cycles Loss of feedwater heatars 180. reactor trip cycles 100% to 0% of RATED TilERMAL POWER 130 hydrostatic pressure and Pressurized to > 930 psig i

leak tests and i 1250 psig,

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ADMINISTRATIVE CONTR01.S f.

h CORE OPERATING t.fMITS REPORT (Continued)

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14. XN-NF 512-P-A, Revision 1 and Supplement 1. Revision 1. *XN-3 Critical Power Correistion, October,1982.

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r 6.9.3.3 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermakhyd.aulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

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6.10 RECORD RETENTION in addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

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Records and logs of unit operation covering time interval at each power level.

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Text

2.0 INTRODUCTION

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