ML17156B521

Amend 60 to License NPF-22,revising Tech Specs to Reflect Resolutions Arrived at in Response to NRC Bulletin 88-007 & Suppl 1, Power Oscillations in Bwrs

ML17156B521
Person / Time
Site:	Susquehanna
Issue date:	11/22/1989
From:	Butler W Office of Nuclear Reactor Regulation
To:
Shared Package
ML17156B522	List: ML17156B521 ML17156B523
References
IEB-88-007, IEB-88-7, NUDOCS 8912070400
Download: ML17156B521 (22)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 PENNSYLVANIA PO'1lER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

OOCKET NO. 50-388 SUS UEHANHA STEAM ELECTRIC STATION UNIT 2 AMEHDMEHT TO FACILITY OPERATING LICENSE Amendment No.

60 License No.

NPF-22 1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

S A.

The application for the amendment filed by the Pennsylvania Power 5

Light Company, dated June 23, 1989 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended

{the Act),

and the Commission's regulations set forth in 10 CFR Chapter I; B.

The facility wi 11 operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; 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 wi 11 be conducted in compliance with the Commission's regu'Iations set forth in 10 CFR Chapter I; 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 E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License Ho.

NPF-22 is hereby amended to read as follows:

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment Ho.

60 and the Environmental Protection Plan con-tained in Appendix B, are hereby incorporated in the license.

PPSL shall operate the facility in accordance with the Technical Specifica-tions and the Environmental Protection Plan.

89'2070400 85'1122 PDR.

AGOCK 0500038S IP PDC

C~

I

3.

This 1icense amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technica1 Specifications Date of Issuance:

November 22, 1989

/s/

lla1ter R. But1er, Director Project Directorate I-2 Division of Reactor Projects I/II PDI-2/PM MThadani:tr (O /(5/89 0

PD I-2/D llBut1er

]a /lg/89

( (/Z.$89 3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

November 22, 1989 Walter R. Butler, Direct or Project Directorate I-2 Division of Reactor Projects I/II

ATTACHMENT TO LICENSE AMENOMENT NO, 60 FACILITY OPERATING LICENSE NO.

NPF-22 OOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of'hange.

The overleaf'ages are provided to maintain document completeness.*

REMOVE xxi xxii 3/4 4-1 3/4 4-la 3/4 4-1h 3/4 4-1c 3/4 4-1d 3/4 4-le 3/4 4-lf 3/4 4-3 3/4 4 4 8 3/4 4-1 8 3/4 4-2 8 3/~ 4-3 8 3/4 4 4 INSERT xxi*

xxii 3/4 4-1 3/4 4-Ia 3/4 4-1h 3/4 4-1c 3/4 4-1d 3/4 4-le 3/4 4-1f 3/4 4-3*

3/4 <-4 B 3/4 <-1 8 3/4 4-2 8 3/4 4-3 8 3/4 4-4*

INOEX AQMINISTRATIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM.

6-23

6. 14 OFFSITE DOSE CALCULATIONMANUAL.......................

6-24 6.15 MAJOR CHANGES TO RAQIOACTIVE WASTE TREATMENT SYSTEMS.....

6-24 SUSQUEHANNA - UNIT 2 XX1

LIST OF FIGURES INDEX FIGURE

3. 1. 5" 1

3. 1. 5-2

3. 2. 1-1

3. 2. 2-1

3. 2. 3" 1

3. 2. 3-2

3. 2. 4" 1

3. 4. l. 1. 1-1 3, 2. 6. 1-1

4. 7. 4-1 B 3/4 3-1 8 3/4. 4. 6" 1

5. 1. 1-1

5. l. 2" 1 5.1. 3"ja

5. 1. 3-lb SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS *..

SODIUM PENTABORATE SOLUTION CONCENTRATION.........

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS.

AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, ANF FUEL.....

~...,

FLOW DEPENDENT MCPR OPERATING LIMIT REDUCED POWER MCPR OPERATING LIMIT.

LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL............

THERMAL POWER RESTRICTIONS MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE SAMPLE PLAN 2)

FOR SNUBBER FUNCTIONAL TEST......

~..

REACTOR VESSEL WATER LEVEL FAST NEUTRON FLUENCE (E>jMeY) AT 1/4 T AS A

FUNCTION OF SERVICE LIFE EXCLUSION AREA..

LOW POPULATION ZONE MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS PAGE 3/4 1-21 3/4 1-22 3/4 2-2 3/4 2-5 3/4 2-7 3/4 2-8 3/4 2-10 3/4 4-lb 3/4 4-18 3/4 7-15 8 3/4 3"8 B 3/4 4-7 5-2 5"4 5-5 SUSQUEHANNA - UNIT 2 Amendment No. 6O

3/4.4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS -

TWO LOOP OPERATION LIMITING CONOITION FOR OPERATION 3.4. 1. l. 1 Two reactor coolant system recirculation loops shall be in operation with the reactor at a

THERMAL POWER/core flow condition outside of Regions I and II of Figure 3.4. 1. l. 1-1.

I APPLICABILITY:

OPERATIONAL CONOITIONS 1" and 2~+, except during single loop operation.¹ ACTION:

a.

In OPERATIONAL CONOITION 1:

With:

a)

No reactor coolant system recirculation loops in operation, or b)

Region I of Figure 3.4. l. 1. 1-1 entered, or c)

Region II of Figure 3.4. l.l. 1-1 entered and core thermal hydraulic instability occurring as ev'idenced by:

1)

Two or more APRM readings oscillating with at least one oscillating greater than or equal to 10% of RATED THERMAL POWER peak-to-peak, or 2)

Two or more LPRM upscale alarms activating and deactivating with a 1 to 5 second period, or 3)

Observation of a sustained LPRM oscillation of greater than 10 w/cm2 peak-to-peak with a 1 to 5

second period, or d)

Region II of Figure 3.4. 1. 1. 1-1 entered and less than 50%

of the required LPRM upscale alarms

OPERABLE, immediately place the reactor mode switch in the shutdown position.

• See Special Test Exception

3. 10.4.

¹See Specification

3. 4. 1. 1. 2 for single loop"operation requirements.

+The LPRM upscale alarms are not requi red to be OPERABLE to meet this specification in OPERATIONAL CONOITION 2.

SUSQUEHANNA - UNIT 2 3/4 4-1 Amendment No.

60

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS ACTION: (Continuedi 2.

If Region II of Figure 3.4. l. 1. 1-1 is entered and greater than or equal to 50Yo of the required LPRM upscale alarms

OPERABLE, immediately exit the region by:

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.

b.

In OPERATIONAL CONDITION 2 with no, reactor coolant system recirculation loops in opera.

on, return at least one reactor coolant system recirculation loop to operation, or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With any pump discharge valve not OPERABLE remove the associated 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 each valve through at least one complete cycle of full travel during each startup"" prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER.

4.4. l. l. 1.2 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5 and 105/, respectively, of rated core flow, at least once per 18 months.

4.4. 1. 1. 1.3 At least 50K of the required LPRM upscale alarms shall be determined OPERABl E 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.

"*Ifnot performed within the previous 31 days.

SUS(UEHANNA - UNIT 2 3/4 4-1'a Amendment No.

60

Figure 3.4.1.1.1-1 THER MAL POWER R ESTR ICTIONS 100 Qo i ~ ~ I

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66 7o SUSQUEHANNA - UNIT 2 3/4 4-1b Amendment No.60

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONDITION FOR OPFRATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed

< 80K of the rated pump speed and the reactor at a

THERMAL POWER/core flow condition outside of Regions I and II of Figure 3.4.1.1.1-1, and a.

the following revised specification limits shall be followed:

1.

Specification

2. 1.2:

the MCPR Safety Limit shall be increased to 1.07.

2.

Table 2.2. 1-1:

the APRM Flow-Biased Scram Trip Setpoints shall be as follows:

Tri Set oint

< 0.

8W + 5

~

Allowable Value 8W +

Specification 3.2.2:

the APRM Setpoints shall be as follows:

Tri Set oint Allowable Value 5g)T

~X)7 SRB (0.58W

+ 45K)T SRB

< (0.58W + 48K)T Specification 3.2.3:

The MINIMUM CRITICAL POWER RATIO (MCPR) shall be greater than or equal to the largest of the following values:

a.

the MCPR determined from Figure 3

~ 2.3-1 plus 0.01, and b.

the MCPR determined from Figure 3.2.3-2 plus 0.01.

5.

Table 3.3.6-2:

the RBM/APRM Control Rod Block Setpoints shall be as follows:

a.

RBM - Upscale Tri Set oint

< 0 6

W +

3 Allowable Value b.

APRM-Flow Biased Tri Set oint

+

Allowable Value APPLICABILITY:

OPERATIONAL CONDITIONS 1~ and 2"+, except during two loop operation.k ACTION:

a.

In OPERATIONAL CONDITION 1:

1.

With 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 II of Figure 3.4. l. l. 1-1 entered and core thermal hydraulic instability occurring as evidenced by:

SUSQUEHANNA " UNIT 2 P

3/4 4" 1c Amendment No. 60

REACTOR COOLANT SYSTEM LIMITING CONDITION fOR OPERATION Continued) heu iy ACTION:

(Continued) 1)

Two or more APRM readings oscillating with at least one oscillating greater than or equal to 10% of RATED THERMAL POWER peak-to-peak, or 2) 3)

Two or more LPRM upscale alarms activating and deactivating with a 1 to 5 second period, or Observation of a sustained LPRM oscillation of greater than 10 w/cm2 peak-to-peak with a 1 to 5 second period, or c

~

d.

d)

Region II of Figure 3.4. 1. l. 1-1 entered and less than 50%

of the required LPRM upscale alarms

OPERALBE, immediately place the reactor mode switch in the shutdown position.

2.

If Region II of Figure 3.4. l. 1. 1-1 is entered and greater than or equal to 50% of the required LPRM upscale alarms are OPERABLE, immediately exit the region by:

a) inserting a predetermined set of high worth control rods, or b) increasing core flow by increasing the speed of the operating recirculation pump.

3.

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

3. 4.1.1

~ 1-1.

In OPERABLE CONDITION 2 with no reactor coolant system recirculation loops in operation, return at least one reactor coolant system recirculation loop to operation, or be in HOT SHUTDOWN within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> s.

With any of the limits specified in 3/4.1.1.2a not satisfied:

l.

Upon entering single loop operation, comply with the new limits within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or be in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

If the provisions of ACTION c. 1 do not apply, take the ACTION(s) required by the referenced Specification(s).

With 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 />.

With any pump discharge valve not OPERABLE remove the associated loop from operation, close the valve and verify closed at least once per 31 days.

SUSQUEHANNA - UNIT 2 Amendment No.60

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 RE UIREMENTS 4.4. 1. 1. 2. 1

4. 4. 1. 1. 2. 2

4. 4. 1. 1. 2. 3

4. 4. l. 1. 2. 4

4. 4. l. l. 2. 5

4. 4. l. l. 2. 6 Upon entering single loop operation and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, verify that the pump speed in the operating loop is

< 80K of the rated pump speed.

At least 50/o of the required LPRM upscale alarms shall be 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.

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

< 30K*""*of RATED THERMAL POWER or the recirculation loop fTow in the operating recirculation loop is

< 50%M'of rated loop flow:

a.

< 145~F between reactor vessel steam space coolant and bottom. head drain line coolant, b.¹¹

< 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure

vessel, and c.¹¹

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

The pump discharge valve and bypass valve in both loops shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup"" prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER.

The pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5X and 105K, respectively, of rated core flow, at least once per 18 months.

During single recirculation loop operation, all jet pumps, including those in the operable loop, shall be demonstrated OPFRABLE 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 no two of the following conditions occur:¹¹¹ a.

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

SUSQUEHANNA - UNIT 2 3/4 4-le Amendment No. 60

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS (Continued)

C.'he indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from estab-lished single recirculation loop patterns by more than 10/o 4.4.1.1.2.7 The SURVEILLANCE REQUIREMENTS associated with the specifications referenced in 3.4. l. 1.2a shall be followed See Special Test Exception 3. 10.4.

If not performed within the previous 31 days.

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 test completion.

¹ See Specification 3.4. 1. l. 1 for two loop operation requirements.

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

¹¹¹ Ouring startup testing following each refueling outage, data shall be recorded for the parameters 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 subsequent required surveillances.

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

SUSQUEHANNA - UNIT 2 3/4 4-l.f Amendment No. 60

~ l.

P

REACTOR COOLANT SYSTEM RECIRCULATION PUMPS LIMITING CONDITION FOR OPERATION 3.4. 1 '

Recirculation pump speed mismatch shall be maintained within:

a.

5X of each other with core flow greater than or equal to 75K of rated core flow.

b.

10K of each other with core flow less than 75K of rated core flow.

APPLICABILITY:

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, either:

a.

Restore the recirculation pump speeds 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 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.

SURVEILLANCE RE UIREMENTS 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

"See Special Test Exception

3. 10.4.

SUSQUEHANNA - UNIT 2 3/4 4-3 Amendment No.

AUG 3 0 1988

REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.4. 1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 145'F, and:

a.

When both loops have been idle, unless the temperature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 50~F, or b.

When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to 50'F, the operating loop flow rate is less than or equal to 50% of rated loop flow, and the reactor is operating at a

THERMAL POWER/core flow condition below the 80%

Rod Line shown in Figure 3.4. l. l. 1-1.

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop.

SURVEILLANCE RE UIREMENTS 4.4. 1.4 The temperature differentials and flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idle recirculation loop.

SUS(UEHANNA - UNIT 2 3/4 4-4 Amendment No.6O

3/4.4 REACTOR'OOLANT SYSTEM BASES 3/4. 4. 1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in accordance with Specification 3.4.1.1.2.

LOCA analyses for two loop operating conditi.ons, which result in Peak Cladding Temperatures (PCTs) below 2200'F, bound single loop operating conditions.

Single loop oper ation LOCA analyses using two-loop MAPLHGR limits result in lower PCTs.

Therefore, the use of two-loop MAPLHGR limits during single loop operation assures that the PCT during a

LOCA event remains below 2200'F.

The MINIMUM CRITICAL POWER RATIO (MCPR) limits for single loop operation assure that the Safety Limit MCPR is not exceeded for any Anticipated Operational Occurrence (AOO).

For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5X decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core and goes up through the inactive loop jet pumps.

Surveillance on the pump speed of the operating recirculation, loop is imposed to exclude the possibility of excessive reactor vessel internals vibration.

Surveillance on differential temperatures below the threshold limits of THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel

nozzles, recirculation pumps and the vessel bottom head during extended opera-tion in the single loop mode.

The threshold limits are those values which will sweep up the cold water from the vessel bottom head.

Specifications have been provided to prevent, detect, and mitigate core thermal hydraulic instability events.

These specifications are prescribed in accordance with NRC Bulletin 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors (BWRs)," dated December 30, 1988.

The boundaries of the regions in Figure 3.4. 1. 1. 1-1 are determined using ANF decay ratio calculations and supported by Susquehanna SES stability testing.

LPRM upscale alarms are required to detect reactor core thermal hydraulic instability events.

The criteria for determining which LPRM upscale alarms are required is based on assignment of these alarms to designated core zones.

These core zones consist of the level A, B and C alarms in 4 or 5 adjacent LPRM strings.

The number and location of LPRM strings in each zone assure that with 50K or more of the associated LPRM upscale alarms OPERABLE sufficient monitoring capability is available to detect core wide and regional osci llations.

Operating plant instability data is used to determine the specific LPRM strings assigned to each zone.

The core zones and required LPRM upscale alarms in each zone are specified in appropriate procedures.

An inoperable jet pump is not, in itself, a sufficient reason to declare a re-circulation loop inoperable, but it does, in case of a design basis accident, increase the blowdown area and reduce the capability of reflooding the core;

thus, the requirement for shutdown of the facility with a jet pump inoperable.

Jet pump fai lure can be detected by monitoring jet pump performance on a

prescribed schedule for significant degradation.

SUSQUEHANNA " UNIT 2 B 3/4 4-X Amendment No.60

REACTOR COOLANT SYSTEM BASES Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two loop operation.

The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.

In the case where the mismatch limits cannot be maintained during the loop operation, continued operation is permitted in the single loop mode.

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 temperature differ-ence was greater than 145'F.

3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code.

A total of 10 OPERABLE safety/relief valves is required to limit reactor pressure to within ASME III a11owable values for the worst case upset transient.

Demonstration of the safety/relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specifica-tion 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.

3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes.

The normally expected background leakage due to equipment design and the detection capability of the, instrumentation for determining system leakage was also considered.

The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.

However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

SUS(UEHANNA - UNIT 2 8 3/4 4-2.

Amendment No.60

REACTOR COOLANT SYSTEM

. BASES 3/4. 4. 4 CHEMISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.

Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.

The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.

During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions.

When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits.

With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chlorides are not exceeding the limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4. 4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods 'with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 microcuries per gram DOSE E(UIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Information obtained on iodine spiking will be used to assess the param-eters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside contain-ment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

SUS(UEHANNA - UNIT 2 8 3/4 4-3 Amendment No. 60

REACTOR COOLANT SYSTEM BASES 3/4.4-6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.

Ouring star tup and shutdown, the rates of temperature and pressure changes are limited so that the maximum speci fied heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

Ouring heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compr essive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions, i.e.,

no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate. of interest must be analyzed on an individual basis.

The reactor vessel materials have been tested to determine their initial RTNpT The results of these tests are shown in Table B 3/4. 4. 6-1.

Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTNpT Therefore, an adjusted reference temperature, based upon the fluence, phosphorus content and copper content of the material in question, can be predicted using Bases Figure B 3/4. 4,6-1 and the r ecommenda-tions of Regulatory Guide 1.99, Revision 1, "Effects of Residual Elements on Predicted Radiation Oamage to Reactor Vessel Materials."

The pressure/

tempera-ture limit curve, Figure 3. 4. 6. 1-1 includes predicted adjustments for this shift in RTNOT for the end of life fluence.

The actual shift in RTNOT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4.6. 1-1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 1.

SUSQUEHANNA - UNIT 2 B 3/4 4-4