ML17157A337

Proposed Tech Specs in Support of Facility Cycle 5 Reload

ML17157A337
Person / Time
Site:	Susquehanna
Issue date:	09/24/1990
From:	PENNSYLVANIA POWER & LIGHT CO.
To:
Shared Package
ML17157A336	List: ML17157A335 ML17157A337 ML17157A338 ML18026A395
References
NUDOCS 9010010143
Download: ML17157A337 (57)

Text

~ SUSQUEHANNA SES UNIT 2 CYCLE 5 TECHNICAL SPECIFICATION CHANGES SEPTEMBER 1990 PENNSYLVANIA POWER 5. LIGHT COMPANY

~ ~ ~

4 9010010143 900924 PDR ADOCK 05000388

INDEX DEFINITIONS SECTION OEFINITIONS (Conti nued) PAGE 1.26 OPERABLE - OPERABILITY..................................... 1-4 1.27 OPERATIONAL CONDITION - CONDITION.......................... 1-4 1.28 PHYSICS TESTS..... 1-5 1.29 PRESSURE BOUNDARY LEAKAGE.................................. 1-5 1.30 PRIMARY CONTAINMENT INTEGRITY.............................. 1-5 1.31 PROCESS CONTROL PROGRAM.................................... 1-5 lo32 PURGE-PURGINGo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-5 1.33 RATED THERMAL POWER... 1-6 1.34 REACTOR PROTECTION SYSTEM'RESPONSE TIME.................... 1-6

~ .. ~........... ~...

1.35 REPORTABLE EVENT... ~ ~ ~ . ~ 1-6 1.36 ROD DENSITY..................;....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 1-6 scsaw sussex /<~crt a

l. 37 SECONDARY CONTAIlNENT INTEGRITY...... 1-6 38 SHUTDOWN MARGIN...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 1.39 SITE BOUNOARY...........;.................................. 1-7 1.40 SOLIDIFICATION................ ~... -. ~........ ~ ~ ~ . ~ ~ ~ . .

~ ~ ~ ~ ~ 1-7 1 41 SOURCE CHECK o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ o ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 1.42 STAGGERED TEST BASIS....................................... 1-7 1 F 43 THERMAL OERo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ 1-7 1.44 TURBINE BYPASS SYSTEM RESPONSE TIME........................ 1-7 1.45 UNIDENTIFIED LEAKAGE.................. 1-7 1.46 UNRESTRICTED AREA............................-. 1-8 1.47 VENTILATION DHAUST TREATMENT. SYSTEM....................... 1-8 1 ~ 48 VENTINGo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ 1-8 SUSQUEHANNA - UNIT 2

INOEX-LIST OF TABLES TABLE PAGE SURVEILLANCE FREQUENCY NOTATION ................... 1-9 1.2 OPERATIONAL CONOITIONS ........................ 1-10 2.2. 1-1 REACTOR PROTECTION SYSTEH INSTRUMENTATION SETPOINTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2-4 pa Q 5Cgha SPE'CD FRACWWCP VeRS~S PVERRae g<RP~ ~Z>eS g/0 Z-Ch 3.3. 1"1 REACTOR PROTECTION SYSTEM INSTRUMENTATION ......... 3/4 3-2

3. 3. 1-2 REACTOR PROTECTION SYSTEM RESPONSE TIHES .......... '/4 3-6 4.3.1. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............'............. 3/4 3-7 3.3. 2-1 ISOLATION ACTUATION INSTRUHENTATION ............... 3/4 3-11 3.3. 2-2 ISOLATION ACTUATION INSTRUMENTATION SETPOINTS ..... 3/4 3-17 3.3. 2-.3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME ..... 3/4 3-21 4.3.2. 1-1 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE RE(UIREMENTS ........................ ............. 3/4 3-23

3. 3. 3-1 EHERGENCY CORE COOLING SYSTEM'CTUATION INSTRUMENTATION ................................... 3/4 3-28 3.343 2 INSTRUMENTATION SETPOINTS ................

EMERGENCY CORE COOLING SYSTEM ACTUATION 3/4 3-31 3.3. 3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ...... 3/4 3-33

4. 3. 3. 1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREHENTS;........ 3/4 3-34

3. 3. 4. 1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION ................................... 3/4 3"37

3. 3.4. 1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS 3/4 3-38 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-39

3. 3.4. 2-1 ENO-OF-CYCLE R'ECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION ................................... 3/4 3-42 3.3.4.2-2 ENO-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS .... 3/4 3-43 3.3.4. 2-3 ENO-OF-CYCLE RECIRCULATION PUHP TRIP SYSTEM

~ ~

RESPONSE TIME 3/4 3"44 SUSQUEHANNA - UNIT 2 xxfv Aaendaent No. 31

INOEX LIST OF FIGURES FIGVRE PAGE

3. 1. 5" 1 SODIUM PENTABORATE SOLUTION TEMPERATURE/

CONCENTRATION REQUIREMENTS . 3/4 1-21

3. 1. 5-2 SODIUM PENTABORATE SOLUTION CONCENTRATION ..... 3/4 1-22

3. 2. 1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL ....... ........

~ . . ... . . ~ 3/4 2-2

3. 2. 2" 1 LINEAR HEAT GENERATION RATE FOR APRM SETPOINTS VERSUS AVERAGE PLANAR EXPOSURE, ANF FUEL .... ....

~ ~ 3/4 2-5

3. 2. 3-1 FLOW DEPENDENT MCPR OPERATING LIMIT.... 3/4 2-7

3. 2. 3-2 REDUCED POWER MCPR OPERATING LIMIT(~...~P .."~.... 3/4 2-8

~uk eiu< '8i'PA~S' pggps~r)

3. 2. 4-1 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE, ANF 9 X 9 FUEL............ 3/4 2-10 3.4.1.1.1-1 THERMAL POWER RESTRICTIONS...............,......... 3/4 4-jb

3. 2. 6. 1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS.

REACTOR VESSEL PRESSURE ~ . 3/4 4-i3

4. 7. 4-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST......... 3/4 7-15 8 3/4 3-1 REACTOR VESSEL WATER LEVEL . 8 3/4 3-8 8 3/4. 4. 6-1 FAST NEUTRON FLUENCE (E)1MeV) AT 1/4 T AS A FUNCTION Of SERVICE LIFE . 3/4 4" 7 5.1. 1-1 EXCLUSION AREA. 5-2

5. 1. 2-1 LOW POPULATION EONE. 5-3

5. 1. 3-la MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS ANO LIQUID EFFLUENTS 5-4

5. '. 3-1b MAP DEFINING UNRESTRICTED AREAS FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.. 5-5 g.Q.3 g ggoucED T vRSx, u<

PoeER McPR OPERII TrP6-SY pAs5 ~uo PERASc.e) aI~z7(nk~u S/0 g-fe.

~~<g(e.oc-RP> '3/0 ~-F4 geouceo eo et e<<R wgoPeRAB~e)

SUSQUEHANNA - UNIT 2 XX11 Amendment No. 6O

1 l

DEFINITIONS RATED THERMAL ONER 1.33 RATED THERMAL PANNER shall be a total reactor core heat transfer rate to the reactor coolant of 3293 HIT.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time fntervai fram when the monitored parameter exceeds fts trip setpofnt at the channel sensor untf1 deenergfzatfon of the scram pilot valve solenofds. The response tfae aay be measured by any series of sequential, overlapping or total steps such that the entire response tfme fs measured.

REPORTABLE EVEHT 1.35 A REPORTABLE EVENT shal'i be any of those conditions specified in Sectfon 50.73 to 10 CFR Part 50.

ROO OENSITY 1.36 ROO OENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted fs.equivalent to 100 ROO OEHSITY.

SECONDARY CONTAINMEHT INTEGRITY 1.37 SECONOARY COHTAINEHT INTEGRITY shall exist when:

a. All secondary cantafnment penetratfons required to be closed during accf dent condf tfons are ef ther:

1. Capable of being closed by an OPERABLE secondary containment autoaatfc'fsolatfon system, or

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

.deactivated automatic damper secured fn fts closed position, except as provided fn Table 3.6.5.2-1 of Specification 3.6.5.2.

b. All secondary containment hatches and blowout panels are cloEad and sealed.

c. The standby gas treatment systea fs OPERABLE pursuant ta Speci ficatfon 3.6.5.3.

d. At least one door fn each access ta the secondary containment is closed.

e The sealing mechanfsa associated with each secondary containment penetration, e.g., welds, bellows, resilient material seals, or O-rings, fs OPERABLE.

The pressure within the secondary containment fs less than or equal to the value required by Specification 4.6.5.1a.

SUSquEHANNA - UNIT 2 1-6

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 1 SAFETY LIMITS THERMAL POWER Low Pressure or Low'low

2. 1. 1 THERMAL POWER shall not exceed 2RL of RATED THERMAL POWER with the 1'f reactor vessel stela rated flow.

dome pressure less than 785 psig or core flow less than APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER steam dome pressure exceeding 2'f RATED THERMAL POWER and the reactor vessel less than 785 psig or core flow less than 10K of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER Hi h Pressure and Hi h Flow 2.1.2 I W 1th the reactor vessel steam dome pressure greater than 785 psig and core flow greater than lGZ of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

4 M3 than 785 psig and core flow greater than 1'f Wft (hQ@1 essthan +8@and the reactor vesseI steam dome pressure greater rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above L325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

SUSQUEHANNA - UNIT 2 2-1 Amendment No. 26

2. 1 SAFETY LIMITS BASES

2. 0 INTRODUCTION The fuel cladding, reactor pr essure vessel and primary system piping are gg) principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel dama e is calculated to occur if the limit is not violated. ause dama is t dir tly o rv le, a tep- a appr ch is se to es lish afet Limi such at e R is ot le than e ii t s cifie ln Spe 'fioat' 2. for F fu . M grea r tha the s cif d lim repre nts a onser ative ar lati to th condi ons r uir to m ntain el cl ddin 'nte r he ue cladding is one of the physical barriers whic separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product migration from clad-ding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses ma cause ross rather than incremental cladding deteri-oratio er re he el add g S et e e i a m in ili io wh' d odu ons o tr sit' PR 1.

Ties represen a significant de arture from the condition intended y esign for planned operation e M f el add g i egri e y imi ur s ing rma ope at' a d d'ng anti ipat d o rat'al ccu enc t eas 99.9 of e f 1 r s 'n t co e d not xpe en tr si 'on b lin ef. -.NF 24( Re sio 1)

2. 1. 1 THERMAL POWER Low Pressure or Low Flow The use of the XN"3 correlation is valid for critical power calculations at pressures greater than 580 psig and bundle mass fluxes greater than 0.25 x 106 lbs/hr-ft~. For 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 mini-mum bundle flow for all fuel assemblies which have a relatively high power and potentiallj can approach a critical heat flux condition. For the ANF 9 x 9 fuel design, the minimum bundle flow is greater than 30,000 lbs/hr. For this design, the coolant minimum flow and maximum flow area is such that the mass flux is always greater than 0.25 x 106 lbs/hr-ft<. Full scale critical power tests taken at pressures down to 14.7 psia indicate that the fuel assembly critical power at 0.25 x 10 lbs/hr-ft is 3.35 Mwt or greater. At 25K thermal power a bundle power of 3.35 Mwt corresponds to a bundle radial peaking factor of greater than 3.0 which is significantly higher than the expected peaking thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressures below 785 psig is conservative.

SUSQUEHANNA - UNIT 2 8 2-1 Amendment No. 58

SAFETY LIMITS BASES

2. 1.2 THERMAL POWER Hi h Pressure and Hi h Flow Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure. However, the existence of critical power, or boi ling transition, is not a directly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow,, feedwater temperature, and core power distribution.

The margin for each fuel assembly is characterized by the critical power ratio (CPR), which is the ratio of the bundle power which would produce onset of tran-sition boiling divided by the. actual bundle power. The minimum value of this ratio for any bundle in-the core is the minimum critical po~er ratio (MCPR).

h Saf y 'mit PR assures sufficient conservatism in the operatin R imi ha in t e event o an anticipa e opera ional occurrence from the limiting con i ion for operation, at least 99.9X o the fuel rods in the core would be expected to avoid boilin transitio T e m gin en ~lc ate+

b, ai i an io tatis ica procedure

. 0 a th af im' e

i aged a considers t e uncertainties in moni oring

~

re operating state. One pecific uncertainty included in the safety limit 0 Ape is the uncertainty inherent in the XN-3 critical power correlation. XN-NF-524

~

(A Revision 1 describes the methodology used in determining the The XN-3 critical

~

'i.- dF- fd- oog

~

power correlation is based on a significant

~,~M body of prac-7.

I tical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small percentage of the actual criti-

'al power being estimated. As long as the core pressure and'flow are within the range of validity of the XN-3 correlation (refer to Section 8 2. 1. 1), the assumed reactor conditions used in e sa ety limit introduce conser-vatism into th 'ecause bounding high radial power factors and bounding flat local peaking distributions are'sed to estimate the number of rods in boiling transition. Still further conservatism is induced by the tendency of the XN"3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation rovide a reasonable de ree of assurance that in s ai er io a af it r ou ~ t s io o'n n e' . f boiling transi-ion were occur, s reason o e ieve that the integrity of the fuel would not necessarily be compromised. Significant test data accumulated by the U.S. Nuclear Regulatory Commission and private organizations indicate that the use of a boiling transition limitation to protect against cladding fai lure is a very conservative approach. Much of the data indicates that LWR fuel can sur-vive for an extended period of time in an environment of boiling transition.

SUSQUEHANNA - UNIT 2 B 2-2 Amendment No. 5B

POWER DISTRIBUTION >HITS 3 2. 3 MINIMUM CRITICAL POWER RATIO LIMITI CONOITION FOR OPERATION 3.2.3 The NIMUM CRITICAL POWER RATIO (MCPR) shall be reater than or equal to the greate of the two values determined from Figur 3.2.3-1 and Figure

3. 2. 3" 2.

APPLICABII ITY: 'PER TIONAL CONDITION 1, when T ERMAL qWq IIA KD IIERIIAL PO Klt.

POWER is greater than or ACTION:

With MCPR less than the appl'cable MCP limit determined above, initiate correc-tive action within 15 minutes nd r tore MCPR to within the required limit;with-in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL PO ER o less than 25K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.3.1 MCPR shall be determined to be eater than or equal to the applicable MCPR limit determine from Figure 3.2.3-1 d Figure 3.2.3-2:

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

b. Wit ~n 12 hours after completion of a ERMAL POWER increase of at 1 st 15K of RATED THERMAL POWER, and

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

d. 'The provisions of Specification 4.0.4 are not ap licable.

R +p>~~+ uJx'T H THE / dc c put ~<6 THR e W p++~~

SUS(UEHANNA " UNIT 2 3l4 2-6 Amendment No. 58

3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL

~ ~ POWER RATIO (MCPR) shall be greater than or equal to the greater of:

a) The Flow-Dependent MCPR value determined from Figure 3.2.3-1, and b) The Power-Dependent HCPR value determined from the following equation:

HCPR = HCPR, + (MCPR - HCPR,) x SCRAM SPEED FRACTION where:

HCPR and HCPR, are determined from curve A and curve B of one of the kollowing figures, as appropriate:

il Figure 3.2.3-2: EOC-RPT, and Hain Turbine Bypass Operable Figure 3.2.3-3: Hain Turbine Bypass Inoperable Figure 3.2.3-4: EOC-RPT Inoperable SCRAM SPEED FRACTION is a number between 0.0 and 1.0 (inclusive) based on measured core average scram speed. This fraction is used to interpolate between Curve B HCPR values corresponding to an average scram speed of 4.2 feet/second and Curve A HCPR values corresponding to the maximum allowed core average scram insertion times given in Specification 3. 1.3.3., The SCRAM SPEED FRACTION is obtained from Table 3.2.3-1.

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

ACTION:

With MCPR less than the applicable HCPR limit determined above, initiate corrective action within 15 minutes and restore MCPR to within the required limit within,2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

0 SURVEILLANCE RE UIREME 4.2.3

~ ~ HCPR, with a) SCRAM SPEED FRACTION = 1.0 prior to the performance of the initial scram time measurement for the cycle in accordance with Specification 4. 1.3.2(a), or b) SCRAM SPEED FRACTION as determined from Table 3.2.3-1 used to determine the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required, by Specification 4. 1.3.2 shall be determined to be greater than or equal to the applicable HCPR limit determined from Figure 3.2.3-1 and the applicable figure selected from Figures 3.2.3-2 through 3.2.3-4:

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

2) Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

3) Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for HCPR.

4) The provisions of Specification 4.0.4 are not applicable.

0

TABLE 3.2.3-1 SCRAM SPEED FRACTION

. VERSUS AVERAGE SCRAM TIMES Rod Positions MAXIMUM TIMES TO ROD'OSITIONS Seconds 45 .38 .39 .40 .41 .42 .43 39 .74 .76 .79 .82 .85 .86 25 1.57 1.63 1.70 1.78 '1. 87 1.93 2.76 2.88 3.01 3.16 3.32 3.49 SCRAM SPEED FRACTION 0.0 0.2 0.4 0.6 0.8 1.0 NOTE: Determine SCRAM SPEED FRACTION from farthest left-hand column whose listed values are all greater than measured average scram times using most recent measurement for each rod.

I' AD fll 2.0

{30,$ .9 CURVE A: EOC-BPT Inoperable; i8 Main Turbine Bypass erable CURVE B: Main Turbine By ss Inoperable; EOC-BPT Oper e la ~

CURVE C: EOC-BPT d Main Turbine 8 ass erable O) u

~

C 4J 40 La Ho.).SO)

Q.

p 1.6 CL O I60.n. w tl {6o.s.<2i 1.41 1.40 1.64.l.col B H

--- l68.93,i.33 1.33 l.3 C 30 40 50 ao Io So 80 100 Total Core Flow (% OF RATED)

FLQW DEPENDENT MCPR OPERATING LIMIT

. FIGURE 3.2.3-1

1.9 (30,1.83) 1.8 1.7 (37,1.69)

(45,1.57)

(60,1.43)

(100,1.32) 1.3 (74.9,1.32) 1.2 30 40 50 60 70 80 90 100 Total Core Flow (% OF RATED)

FLOW DEPENDENT MCPR OPERATING LIMIT FIGURE 3.2.3-1

1.7 CURVE A: EOC-APT Inoperable:

Main Turbine Bypass Operable CURVE 8: Main Turbine Bypass Inoperable<

EOC-APT. Operable CURVE C: EOC-RPT and Main Turbine

{26.1.64) Bypass Operable 0.1 62)

E

~ {ds, I.4Q)

~

Ul 1.5 - {26.1.62)

{80.1.47)

~

C

~ {40.1 50)

{56.1.47)

Cal

-n

~'a 0 CL

{25,1.44) {so.1.46) 1.41 bJ p

{40.1 42)

I Q) t 1.4 1.40 O

CL {Qo Q 141) a C3 {a6.1.

{80.1.37) C 1.33

{QO.Q.1.33).

1.2 20 30 40 60 ao 70 eo 80 100 O Core Power (% OF RATED)

REDUCED POWER MCPR OPERhTINQ LIMIT FIGURE 3.2.9-2

l 2.0 (25,1.90) 1.9 Legend CURVE A: SCRAM SPEED CORRESPONDING TO T.S. 3.1.3.3 1.8 CURVE B: SCRAM SPEED 4.2 ft/sec E (40,1.73) 1 7 U) (25,1.65)

C 1.8 O.

0 IZ (40,1.55) (65,1.50) 1.5 (100,1.47)

(6e.4,1.47) 1.4 (65 ,1.41)

(100,1.32) 1.3 84,1.32) 20 30 40 eO BO 7O 80 90 100 Core Power (% RATED)

REDUCED POWER MCPR OPERATING LIMIT EOC-RPT AND MAIN TURBINE BYPASS OPERABLE FIGURE 3.2.3-2

2.0 (25,2.0)

(40,1.91) 1.9 A.

1.8 (65,1 .77)

(25,1.70)

(84,1.64)

(40,1.63)

(65,1. 54) (100,1.56) 1.4 Legend (84,1 .43)

(100,1.38)

CURVE A: SCRAM SPEED CORRESPONDING TO T.S. 3.1.3.3 1.3 CURVE 8: SCRAM SPEED ~ 4.2 ft/sec 1.2 20 30 40 eo eo 7o 80 90 100 Core Power (% RATED)

REDUCED POWER MCPR OPERATING LIMIT MAIN TURBINE BYPASS INOPERABLE FIGURE 3.2.3-3

2.0 (25,1.95) 1.9 Legend CURVE A: SCRAM SPEED CORRESPONDING TO T.S. 3.1.3.3 1.8 CURVE B: SCRAM SPEED = 4.2 ft/sec (40,1.76)

E (25,1.70) 'A 1.7 U)

C 1.6 O.

0 (40,1.57) -A- (100,1.54)

CC 1.5 (64.1,1.54) 1.4 (65,1 42)

(100,1.35)

(77.1,1.3 5) 1.3 1.2 20 30 50 60 70 80 SO 100 Core Power (% RATED)

REDUCED POWER MCPR OPERATING LIMIT EOC-RPT INOPERABLE FIGURE 3.2.3-4

I=lgure 8.4.1.1.t-h THERMAL POWER REST CTtONS 100 40 lan 70 g ~ I ~

40 et~

r ~

So'Il 0

g /:,"II sa 0

io-"'

30-0 20 ~

3d 45 40 dd 40 dd r0 Core Flow (% RATED) 7 8ESa.e.ammu& tI1p~~

SUS(UEHANNA - UNIT 2 3/4 4-lb Amendment N0.60

Figure 3.4.1.1.1-1 THERMAL POWER RESTRICTIONS 100 II ~ '

Qo 80 I

."""~ pop 70 . ~ ~ 1i ~ ~ ~ ~

50 ~ Llg i DP g>~

~ P ID o 50 ",: ', ii B

~ Q I ~

l5 E

I L 30 P

~ ~

20 O ~ ~ 1 10 I r

IP ~ ~ ~ r ~ ~ ~

30 36 40 '6 60 56 50 55 70 Core Flow (% RATED)

REACTOR COOLANT SYSTEM RECIRCULATION LOOPS - SINGLE LOOP OPERATION LIMITING CONOITION FOR OPERATION 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 foilowing revised specification limits shall be followed:

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

Tri Set oint Allowable Value

+

Specification 3.2.2:, the APRM SRB Setpoints shall be Tri Set oint

'4M)T (0.58M + 45K)T

~K) as SRB Specification 3.2.3: The MINIMUH CRITICAL POWER RATIO (MCPR) shall follows:

Allowable Value

< (0.58W + 48%)T be greater than or equal to the largest of the following values:

a. the MCPR determined fr 1 and xdu II'3 0 ~3 o R r I paxArej

b. the MCPR determined .. xsuRc

- us Table 3:3.6-2: the RBM/APRM Control Rod Block Setpoints shall be as follows:

a. RBM - Upscale Tri Set oint Allowable Value

+

b. APRM-Flow Biased Tri Set oint Allowable Value

~ + + io APPLICABILiTY: OPERATIONAL CONDITIONS 1* and 2"+, except during two loop operation.P ACTION:

In OPERATIONAL CONOITION 1:

1. Wi th 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. 1. l. 1-1 entered and core thermal hydraulic instability occurring as evidenced by:

SUSQUEHANNA - UNIT 2 3/4 4-1c Amendment No. 60

I REACTOR COOLANT SYSTEM LIMITING CONOITION 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 Upon entering single loop operation and at least once per 24 haurs thereafter, veri fy that the pump speed in the operating loop is < 80K of the rated pump speed.

4. 4. 1. 1. 2. 2 At, least 50K of the requi~ed LPRM upscale alarms shall be determined OPERABLE by performance of the following on each LPRM upscale alarm.

1) CHANNEL FUNCTIONAL TEST at least ance per 92 days, and

2) CHANNEL CALIBRATION at least once per 184 days.

4. 4. 1. l. 2. 3 Within 15 minutes prior '.o either THERMAL POWER increase resulting fram a control rod withdrawal or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is < 30K"""" of POWER or the recirculation loop fTow in the operating RATEO'HERMAL recirculation loop is < 50K""" 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 < 504F between the reactor coalant within the loop not in operation and operating loop.

4. 4. 1. l. 2. 4 The pump discharge valve and bypass valve in both loops shall l be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel du~ing each startup"" prior to THERMAL POWER exceeding 25% of RAT'ED THERMAL POWER.

4. 4. 1. 1. 2. 5 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.

m vo Pa RhSLE

4. 4. 1. l. 2. 6 Ouring single recirculation oop operation, all jet pumps, including those in the laop, shall be demonstrated OPERABLE 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.

SUS(UEHANNA - UNIT 2 3P4 4-le Amendment No. 60

REACTOR COOL'ANT SYSTEM SURVEILLANCE RE UIREHENTS (Continued

c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from estab-lished single recirculation loop patterns by more than 10%

4.4.1.1.2. 7 The SURVEILLANCE REQUIREMENTS associated with the specifications referenced in 3.4. 1. 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.

See Specification 3. 4. 1. 1. 1 ~or two loop operation requirements.

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

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 ac.uai 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 OPERASLE to meet this specification in OPERATIONAL CONOITION 2.

indicated total core flow differs value by more than 10K

b. The flow from single from the established total core recirculation loop flow measurements.

SUSQUEHANNA - UNIT 2 3/4 4-.1f'mendment No. 60

l

\I 0

REACTIVITY CONTROL SYSTEMS BASES REACTIVITY ANOMALIES (Continued)

Since the comparisons are easily done, frequent checks are not an imposi-tion on normal operation. A lX deviation in reactivity from that of the pre-dicted is larger than expected for normal operation, and therefore should be thoroughly evaluated. A deviation as large as lX would not exceed the design conditions of the reactor.

3/4. 1. 3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are .consistent with those used in the accident analysis, and (3) limit, the potential effects of the rod drop accident. The ACTION statements permit variations from the basic re-quiriments but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The re-quirements fot the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

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

0 Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTINM MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inopdrable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is desi ned to brin t actor subcritical at a rate fast enough to pr en HCP rom ecom g le tha he 1 it s eci-f o 2.1.2 urin he re w e tr sient analy d in he cy eci fi tra ient alys rep t. is an ysis shows hat e ne tiv ac vity tes esult g f the cram ith t ave ge re ons of th dr es a give in t spec icati s, ovide e uired rot tio and PR emai rea r th the mit ecifi d in on .1. . e occurrence o scram t mes onger then those spec f ed should be viewe as an indication of a systematic problem with the rod drives and therefore the surveillance inter-val fs reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept dischargo water from the control rods during a SUSQUEHANNA - UNIT 2 B 3(4 1-2 Amendment No. 31

1 REACTIVITY CONTROL SYSTBlS BASES CONTROL ROD PROGSN CONTROLS (Continued)

The RSCS and RW provide automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.

Parametric Control Rod Drop Accident analyses have shown that for a wide range of key reactor parameters (wh'ich envelope the operating ranges of these variables), the fuel enthalpy rise during a postulated control rod drop acci-dent remains considerably lower than the 280 cal/gm limit. For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient, effective delayed neutron fraction, and maximum four bundle local peaking factor are compared with the inputs to the parametric analyses to de-termine the peak fuel rod enthalpy rise. This value is then Compared against the 280 cal/gm design limit to demonstrate compliance for each operating cycle.

If cycle-specific values of the above parameters are outside the range assumed in the parametric analyses, an extension .of the analysis or a cycle-specific analysis may be required. Conservatism present in the analysis, results of the parametric studies, and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in XN-NF-80-19 Volume l.

P L, N F- po - 0 0 g.<+0 The RBN is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided. Tripping one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the written sequence used by the operator for withdrawal of control rods.

3/4. 1.5 STAND8Y LI UID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that none of the withdrawn control rods can be inserted. To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes. A minimum quantity of 4587 gallons of sodium pentaborate solution containing a minimum of 5500 lbs. of sodium pentaborate is required to meet this shutdown require-ment. There is an additional allowance of 165 ppm in the reactor core'to account for imperfect mixing. The time requirement was selected to override the reactivity insertion rate due to cooldown following the Xenon poison peak and the required pumping rate is 41.2 gpm. The minimum storage volume of the solution is established to allow for the portion below the pump'uction that cannot be inserted and'the filling of other piping systems connected to the reactor vessel. The temperature requirement for the sodium penetrate solution is necessary to ensure that the sodium penetaborate remains in solution.

Nth redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.

SUSQUEHANNA - UNIT 2 8 3/4 1"+ Amendment Ho. 31

3/4.2 POWER OISTRIBUTION LIMITS BASES spec'tions this sec 'ssure t the a clad g tern r t e fo stul ed d sign sis loss- -coo ant ccid nt ot exc owing e 22 'imi e p ecified in CFR 0.4 .

11 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit spec-ified in 10 CFR 50.46.

The'peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average he=t generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The Technical Specification APLHGR for ANF fuel is specified to assure the PCT following a postulated LOCA wi 11 not exceed the 2200'F limit. The limiting value for APLHGR is shown in Figure -3. 2. 1-1.

The calculational procedure used to establish the APLHGR shown on Figure 3.2. 1-1 is based on a loss-of-coolant accident analysis. The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. These models are described in XN-NF-80-19, Volumes 2, 2A, 2B and 2C..

, 3/4. 2. 2 APRM SETPOINTS The flow biased simulated thermal power-upscale scram setting and flow biased simulated thermal power-upscale control rod block functions of the APRM instruments limit plant operations to the region covered by the transient and accident analyses.

In addition, the APRM setpoints must be adjusted to ensure that >1% plastic strain and fuel centerline melting do not occur during the worst anticipated operational.

occurrence (AOO), incl.uding transients initiated from partial power operation.

For ANF fuel the T factor used to adjust the APRM setpoints is based on the FLPO calculated by dividing the actual LHGR by the LHGR obtained from Figure 3.2.2- i.

The LHGR versus exposure curve in Figure 3.2.2-1 is based on ANF's Protection Against Fuel Failure (PAFF) line shown in Figure 3.4 of XN-NF-85-67(A), Revision 1.

. Figure 3.2.2-1 corresponds to the ratio of PAFF/1.2 under which cladding and fuei integrity is protected during AOO's.

SUSQUEHANNA " UNIT 2 B 3/4 2-a Amendment No. 58

e

'OWER OISTRIBUTION LIMITS BASES 3/4. 2. 3 HINIMUH CRITICAL POWER RATIO The ified in For an abnormal S eci fication 3. 2. 3 are derived from analy transient analys~s

'f required operating limit MCPRs at steady state o eratin conditions as speci-dition of the reac or ein at the steady state o era in limit it i e uired abnormal operational transients.

with the initial con-that the resulting o n eas w Sa y l M a Ill g e an 'en assuming ins rumen rip se ing given in pec>>ca-tion S hat f cl ntegr' S fety imi is urin T

ny a ure ant'pa d ab e

rma 1 ding opera eter 'ne onal rans'ent,

'ch r ult 'h he st 1'ti ct'tr n ex eded ients in h e en alyz to w TICA OW RA 0 (C ). e type of transients evaluated wei e loss of flow, increase positive reactivity insertion, and in pressure decrease.

an 1 >>n ower, tr sien ie e rges elta li t coolant tern erature CPR. Whe adde pec'cat' o

t

.2.

a ty imi is bt 'd CP ,

d p th req sent d in red m'mum pera ure 3.2.

ng 1 an .2.

H of The

'h alu ion f a 'n ti sient gins 'th t syste initi param ers s wn cycle pecifi transi nt ana sis r port at ar input an A ore yna c be ior tr nsient omput prog m. T outp s of s pro am al g w' th initi HCPR f m the 'nput r fur er an yses the t rmall imi 'ng bu le. T e codes nd me odolo to ev uate ressu zatio and no-suri descri in -NF-79 N-NF- 4-105. The p lnci-pr 1 re lt ion e nts of is eva ation ar

'he d ducti 1 an in MC cau d b e tra ent.

Figure 3.2.3-1 defines core flow dependent HCPR operating limits which assure that during a flow increase tran-sien re u ing rom a motor-generator speed control failure. The flow depend-ent MCPR is only calculated for the manual flow control mode. Therefore ef es t automatic flow control o era p er dep dent PR o rat' n is not ermitted'.

lim'hi ass es F gure at e Sa ty 1't M i 1 no be vi ated n t even of eedw er ntro ler i lur , Rod W'dr wal E or, o Lo .Reje Wi out M n T bine ypa oper le i tip(ed ro a red ed P o er c diti Cycle specific analyses are performed for the most limiting local core wide tran-sients to determine'thermal margin. Additional analyses are performed to determine the MCPR operating limit with either the Hain Turbine Bypass inoperable or the EOC-RPT inoperable. Analyses to determine thermal margin with both the EOC-RPT inoperable and Hain Turbine Bypass inoperable have not been performed. Therefore, operation in this condition is not permitted.

SUS(UEHANNA - UNIT 2 .B 3/4 2-2 Amendment No. 58

POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

At THERMAL POWER levels less than or equal to 25K of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content wi',1 be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the re-s'ulting MCPR value is in excess of requirements by a considerable mar in. During initial start-up testing of the plant, a HCPR evaluation made at 25K,o RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin demonstrated such that futur e HCPR evaluation below this power level unnecessary. The daily requirement for calculating HCPR when THERMAL POWER 1s greater than or equal to 25K of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

SUS(UEHANNA - UNIT 2 B 3/4 2-3, Amendment No. 58

I 1

t

3/4.4 REACTOR COOLANT 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 conditions, which r esult in Peak Cladding Temperatures (PCTs) below 2200~F, bound single loop operating conditions.

Single loop operation LOCA analyses using two-loop MAPLHGR imi ts result in 1

lower PCTs. Therefore, the use of two- loop MAPLHGR limits during single loop oper ion assur t the PCT durin a LOCA event remains below 2200 F.

7 ftE'R H AL 4ME'R, The MINIMUM CRITICAL POWER RATIO M ~mits for single loop operation assure that th Sa ety Limit is not exceeded r'or any Anticipated Operational Occurrence (AOO).

For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5%

decrease in recircuTation drive flow to account for the active loop drive fiow 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 r'eactor 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 ope.a-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 n accordance with NRC Bulletin 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors (BWRs)," dated Oecember 30, 1988. The boundaries nf the regions in .Figure 3.4, 1. 1. 1-1 are aetermined using ANF decay ratio calculations and supported by Susquehanna SES stability testing.

LPRM upsca le alarms are required to detect reactor core thermal hydraul ic instability events. The criteria for determining which LPRM upscale alarms are required is based on assignment of these alarms to designated core zones.

These cor'e :ones consist of the level A, 8 and C alarms in 4 or 5 adjacent LPRM str.ings. The number and location of LPRM strings in each zone assure that with 5N or more of the associ.ated LPRM upscale alarms OPERABLE sufficient monitoring capability is availaole to detect core wide and regiona:

osci llations. Operating plant instability data is used to determine the specific LPRM strings assigned to each zone. The core zones and required 'RY upscale alarms in each zone are specified in appropriate procedures.

An inoperaole 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 failure can be detected by moni toring jet pump perfor mance on a prescribed schedule for significant degradation.

SUSqUEHANNA - UNIT 2 8 3/4 4-1 Amendment No.60

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3. 1 The reactor core shall contain 764 fuel assemblies.

79 fuel rods and two water rods. Each fuel rod shall have a nominal active fuel length of 150 inches. Reload fuel shall have a maximum avera e enrichment of 4.0 weight percent U-235.

C TR RO ASS BL

5. 3 2 e r actor ore hall onta '85 ontr rod sse lies eac

'he c si in of a ucif rm ar ay of stain ess s eel bes onta'ng 43 f b ron carbi , B , pow er su ound by cruc'rm hap st 'nle st sh th 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 f'r 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 5?5 F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 22,400 cubic feet at a nominal T ave of 528F.

SUS(UEHANHA " UNIT 2 5-6 Amendment No. 58

4 I 1

INSERTS Insert I SCRAM SPEED FRACTION is a number between 0.0 and 1.0 (inclusive) based on measured core average scram speed. The SCRAM SPEED FRACTION is used to determine the Power-Dependent HCPR value in Specification 3.2.3 and can be obtained from Table 3.2.3-1.

Insert 2

, THERMAL POWER will be limited such that at least 99.9% of the fuel rods are not expected to experience boiling transition.

Insert 3 99.9% of the fuel rods expected'o avoid boiling transition Insert 4 The THERMAL POWER, High Pressure High Flow, Safety Limit is defined to be a core condition such that at least 99.9% of the fuel rods are not expected to experience boiling transition, which represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

Insert 5 The HCPR operating limits assure that, during normal operation and during anticipated operational occurrences the Safety Limit will not be violated. In other words, at least 99.9% of the fuel rods in the core would not be expected to experience boiling transition. The methodology used to establish the HCPR operating limits is described in PL-NF-90-00l.

Insert 6 The HCPR operating limit assures sufficient conservatism in plant operation such that, Insert 7 fraction of fuel rods in boiling transition for various HCPR values.

Insert 8 during an anticipated operational occurrence at least 99.9% of the fuel rods would not be expected to experience boiling transition.

Insert 9 ANF fuel is monitored using the XN-3 Critical Power Correlation. ANF has determined that this correlation provides sufficient conservatism to preclude the need for any penalty due to channel bow. The conservatism has been evaluated by ANF to be greater than the maximum expected hCPR (0.02) due to channel bow in C-lattice plants using channels for only one bundle lifetime.

0 Since Susquehanna SES Unit 2 is a C-lattice plant and uses channels for only one bundle lifetime, monitoring of the HCPR limit with the XN-3 Critical Power Correlation is conservative with respect to channel bow and addresses the concerns of NRC Bulletin No. 90-02 entitled "Loss of Thermal Margin Caused by Channel Box Bow."

j%I pl I 1

Insert 10 assure that at least 99.9% of the fuel rods are not 'expected to experience boiling transition. The HCPR operating limits are adjusted based on measured scram time data in Specification 3.2.3 to assure the validity of the transient analyses.

Insert 11 degradation in thermal margin be such that at least 99.9% of the fuel rods in the core are not expected to experience boiling transition, Insert 12 To assure that at least 99.9% of the fuel rods are not expected to experience boiling transition during any anticipated operational occurrence, the most limiting transients have been analyzed.

Insert 13 The limiting transient yields the largest required HCPR operating limit. The required HCPR operating limits as functions of core power, core flow, and plant equipment availability condition are presented in Figures 3.2.3-1 through 3.2.3-4.

Insert 14 The, transient analyses to determine the HCPR operating limits are performed methods described in PL-NF-90-001. Certain of the pressurization 'sing transients are analyzed statistically assuming a scram insertion versus time curve which is faster than the Technical Specification 3. 1.3.3, limits. The HCPR operating limits are adjusted based on measured scram time data.

Insert 15 at least 99.9% of the fuel rods are not expected to experience boiling transition Insert 16 Figures 3.2.3-2, 3.2.3-3, and 3.2.3-4 define the power dependent HCPR operating limits which assure that at least 99.9% of the fuel rods are not expected to experience boiling transition during the limiting event '(i.e.,

Feedwater Controller'Failure, Rod Withdrawal Error, or Load Rejection Without Hain Turbine Bypass operable) initiated from a reduced power condition..

Insert 17 In addition, the HCPR limits for single-loop operation prote'ct against the effects of the Recirculation Pump Seizure Accident. That is, for operation in single-loop with an operating HCPR limit > 1.30, the radiological consequences of a pump seizure accident from single-loop operating conditions are but a small fraction of 10CFR100 guidelines.

Insert 18 One fuel assembly shall contain 78 fuel rods, one inert rod, and 2 water rods.

All other fuel assemblies shall contain

Insert 19 1'ONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 control rod assemblies consisting of two different designs. The "original equipment" design consists of a cruciform array of stainless steel tubes containing 143 inches of boron carbide (B4C) powder surrounded by a stainless steel sheath. The "replacement" control blade design consists of a cruciform array of stainless steel tubes containing 143 inches of boron carbide (B4C) powder near the center of the cruciform, and 143 inch long solid hafnium rods at the edges of the cruciform, all surrounded by a stainless steel sheath.

NO SIGNIFICANT HAZARDS CONSIDERATIONS The following three questions are addressed for each of the proposed Technical Specification changes:

~ ~

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

3. Does the proposed change involve a significant reduction in a margin of safety?

S ecifications 1.0 Def'nitions and 3 4.2.3 Minimum Critical Power Ratio The changes to these specifications support new HCPR operating limits based on the PP&L reactor analysis methods described in Reload Summary Report Reference

3. The limits calculated for U2C5 will be a function of scram speed.

Therefore, the format for Specification 3/4.2.3 has changed significantly and the new definition is required.

1. No. The HCPR operating limits for U2C5 were generated with the PP&L reactor analysis methods described in PL-NF-90-001 (See Reload Summary Report Reference 3). The U2C5 NCPR operating limits are presented as NCPR versus Percent of Rated Core Flow and NCPR versus Percent Core Thermal Power. These limits cover the allowed operating range of power and flow. As specified in PL-NF-90-001, six major events were analyzed. These events can be divided into two categories: core-wide transients and local transients. The core-wide transient events analyzed were:

1) Generator Load Rejection Without Bypass (GLRWOB),

2) Feedwater Controller Failure (FWCF),

3) Recirculation Flow Controller Failure Increasing Flow (RFCF), and

4) Loss of Feedwater Heating (LOFWH)

As discussed in PL-NF-90-001, the other core-wide transients are non-limiting (i.e., would produce lower calculated ~CPRs than one of the four events analyzed). The local transient events analyzed were:

1) Rod Withdrawal Error (RWE), and

2) Fuel Loading Error (FLE).

The fuel loading error evaluation includes analysis of both rotated and mislocated fuel bundles.

Page 1 of 8

Sufficient analyses were performed to define the HCPR operating limits as a function of core power and core flow. Analyses were also pe}formed to determine HCPR operating limits for three plant equipment availability conditions: 1) Turbine Bypass and EOC-RPT operable, 2) Turbine Bypass inoperable, and 3) EOC-RPT inoperable.

Core-Wide Transients The PP&L RETRAN'odel and methods described in PL-NF-89-005 and PL-NF-90-001 (See Reload Summary Report References 2 and 3) were used to analyze the GLRWOB, FWCF, and RFCF events. The ~CPRs were evaluated using the XN-3 Critical Power Correlation (See Reload Summary Report Reference 26) and the methodology described in PL-NF-90-001 (See Reload Summary Report Reference 3). The GLRWOB and FWCF events were analyzed in two different ways (as described in PL-NF-90-001):

1) Deterministic 'analyses using the. Technical Specification scram speed. (minimum',allowed);

2) Statistical Combination of Uncertainty (SCU) analyses at an average scram speed of 4.2 feet/second.

Thus, the Technical Specification HCPR operating limits calculated for U2C5 will be a function of scram speed.

The LOFWH event was conservatively analyzed by PP&L using the steady state core physics methods and process described in Reload Summary Report References 1 and 3, and the LOFWH event results were found to be bounded by results of the other three core-wide transients. The minimum HCPR operating limit required for the U2C5 LOFWH event is 1. 17.

Results of the GLRWOB, FWCF, and RFCF events are presented in Reload Summary Report Tables 3, 4, and 5, respectively.

Local Transients The fuel loading error (rotated and mislocated bundle) and the Rod Withdrawal Error (RWE) were analyzed using the methodology described in PL-NF-90-001. The results of these analyses apply to all three plant equipment availability conditions previously described, and the results are independent of scram speed. The RWE analysis supports the use of both the Duralife 160C control blades and a Rod Block Honitor setpoint of 108X. The HCPR operating limits that result from the analyses of these events are presented in Reload Summary Report Table 6. These events are non-limiting for U2C5.

Based on the above, the methodology used to develop the new HCPR operating limits for the Technical Specifications does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Page 2 of 8

2. No. The methodology and results described above can only be evaluated for their effect on the consequences of analyzed events; they cannot create new ones. The consequences of analyzed events were evaluated in l. above.

3. No. Based on l. above, the methodology used to generate the HCPR operating limits for U2C5 is both sufficient and conservative.

Furthermore, although the methodology (PL-NF-90-001) is still undergoing NRC review, PP&L believes it meets all pertinent regulatory criteria for use in this application. Therefore, its use will not result in a significant decrease in any margin of safety.

S ecification 2. 1.2 Thermal Power Hi h Pressure and Hi h Flow

1. No. The PP&L Statistical Combination of Uncertainties (SCU) methods are described in Reload Summary Report Reference 3. When using the SCU methodology, the transient aCPR and traditional HCPR safety limit analyses are combined through a single unified analysis. As a result, the Thermal Power, High Pressure and High Flow safety limit is not represented as a single HCPR value, but rather as a condition such that at least 99.9X of the fuel rods in the core are expected to avoid boiling transition. As described in Appendix B of Reload Summary Report Reference 3, this combined analysis and compliance with the resulting safety limit condition are supported by "HCPR Safety Limit type" calculations. The "HCPR Safety Limit type" calculations were performed by ANF using the same methods and assumptions as the traditional HCPR Safety Limit analysis.

As shown in Reload Summary Report Table 1, a HCPR value of 1.06 in two loop operation assures that less than 0. 1X of the fuel rods are expected to experience boiling transition. The methodology and generic uncertainties used in the "HCPR Safety Limit type" calculations are provided in XN-NF-80-19(P)(A), Volume 4 Revision 1 (Reload Summary Report Reference 6). The uncertainties used for the SSES U2C5 "HCPR Safety Limit type" calculations are the same as for U2C4 and are presented in Reload Summary Report Reference

18. The results are presented in Reload Summary Report Table l.

During U2C5, as in the previous cycle, the ANF 9x9 fuel will be monitored using the XN-3 critical power correlation. ANF has determined that this correlation provides sufficient conservatism to preclude the need for any penalty due to channel bow'during U2C5. Susquehanna SES is a C-lattice plant and uses channels for only one fuel bundle lifetime. The conservatism has been evaluated by ANF to be greater than the maximum expected ~CPR (0.02) due to channel bow in C-lattice plants using channels for only one fuel bundle lifetime. Therefore, the monitoring of the HCPR limit is conservative with respect to channel bow and addresses the concerns of NRC Bulletin No. 90-02. The details of the evaluation performed by ANF have been reported generically to the NRC (Reload Summary Report Reference 17).

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Based on the above, the methodology used to develop the new safety limit condition for the Technical Specification does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. No. The methodology and results described above. can only be evaluated for their effect on the consequences of analyzed events; they cannot create new ones. 'The consequences of analyzed events were evaluated in l. above.

3. No. Based on l. above, the methodology'sed,to'enerate the Thermal ,

Power, High Pressure and High Flow safety limit condition for U2C5 is both sufficient and conservative. Furthermore, although the methodology (PL-NF-90-001) is still undergoing NRC review, PP&L believes it meets all pertinent regulatory criteria for use in this application. Therefore, its use will not result in a significant decrease in any margin of safety.

S ecification 3 4.4. 1 Recirculation S stem Two Loo 0 eration The changes to this specification (i.e., Figure 3/4.1. 1. 1-1) reflect cycle-specific stability analyses.

1. No. COTRAN core stability calculations were performed for Unit 2 Cycle 5 to determine the decay ratios at predetermined power/flow conditions. The resulting decay ratios were used to define operating regions which comply with the interim requirements of NRC Bulletin No. 88-07, Supplement 1 "Power Oscillations in Boiling Water Reactors". As in the previous cycle, Regions B and C of the NRC Bulletin have been combined into a single region (i.e., Region II), and Region A of the NRC Bulletin corresponds to Region I.

Region I has been defined such that the decay ratio for all allowable power/flow conditions outside of the region is less than 0.90. To mitigate or prevent. the consequences of instability, entry into this region requires a manual reactor scram. Region I for Unit 2 Cycle 5 is slightly different than Region I for the previous cycle.

Region II has been defined such that the decay ratio for all allowable power/flow conditions outside of the region (excluding Region I) is less than 0.75. For Unit 2 Cycle 5, Region II must be immediately exited if it is inadvertently entered. Similar to Region I, Region II is slightly different than in the previous cycle.

In addition to the region definitions, PP&L has performed stability tests in SSES Unit 2 during initial startup of Cycles 2, 3 and 4 to demonstrate stable reactor operation with ANF 9x9 fuel.

The test results for U2C2 (See Reload Summary Report Reference 20) show very low decay ratios with a core containing 324 ANF 9x9 fuel assemblies.

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Figure 3/4. 1. 1.1-1 is also referenced by Specification 3/4.4. 1.1.2, which governs Single Loop Operation (SLO). The evaluation above applies under SLO conditions as well.

Based on the above, operation within the limits specified by the proposed changes will ensure that the probability and consequences of unstable operation will not significantly increase.

2. No. The methodology described above can only be evaluated for its effect on the consequences of unstable operation; it cannot create new events. The consequences were evaluated in 1. above.

3. No. PP8L believes that the use of Technical Specifications that comply with NRC Bulletin 88-07, Supplement 1, and the tests and analyses described above, will provide assurance that SSES Unit 2 Cycle 5 will comply with General Design Criteria 12, Suppression of Reactor Power Oscillations. This approach is consistent with the SSES Unit 2 Cycle 4 method for addressing core stability (See Reload Summary Report References 4 and 5).

S ecification 3 4.4. 1 Recirculation S stem Sin le Loo 0 eration The changes to this specification are either evaluated above or are editorial in nature. The reference to Specification 2.1.2 is deleted because the new limit (see Evaluation of Specification 2. 1.2 above) will not change for Single Loop Operation. The additional figures referenced from Specification 3.2.3 are the result of the HCPR operating limit analyses evaluated above.

The other two changes to Surveillance Requirements 4.4. 1. 1.2.6, correct inadvertent typographical errors that occur red during the issuance of Amendment 60 to the Unit 2 Technical Specifications.

1. No. The changes are either evaluated elsewhere in this No Significant Hazards Considerations evaluation, or are entirely editorial in nature.

2. No. See l. above.

3. No. See l. above.

S ecification 5.3. 1 Fuel Assemblies This section has been changed to describe the actual core configuration for U2C5, which includes one inert (i.e., solid zircaloy-2) rod.

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1. No. The inert rod was used to repair a fuel assembly that failed during U2C2. This repaired assembly was analyzed and found to be acceptable in support of U2C4 operation, which was approved by the NRC (See Reload Summary Report Reference 5). Based on the above, use of the repaired assembly does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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3. No. See I above.

S ecification 5.3.2 Control Rod Assemblies The changes to this specification are provided in order to recognize the replacement control blade design being utilized in U2C5.

1. No. The main differences between the replacement Duralife 160C control blades and the original equipment control blades are:

1 a) the Duralife 160C control blades utilize three'solid hafnium rods at each edge of the cruciform to replace the three B~C rods that are most susceptible to cracking and to increase control blade life; 1>

b) the Duralife 160C control blades utilize improved B~C tube material (i.e. high purity stainless steel vs. commercial purity stainless steel) to eliminate cracking in the remaining B4C rods during the lifetime of the control blade; c) the Duralife 160C control blades utilize GE's crevice-free structure design, which includes additional B C tubes in place of the stiffeners, an increased sheath thickness, a full length weld to attach the handle and velocity limiter, and additional coolant holes at the top and bottom of the sheath; d) the Duralife 160C control blades utilize low cobalt-bearing pin and roller materials in place of stellite which was previously utilized; e) the Ouralife 160C control blade handles are longer by approximately 3. 1 inches in order to facilitate fuel moves within the reactor vessel during refueling outages at Susquehanna SES; and f) the Duralife 160C control blades are approximately 16 pounds heavier as a result of the design changes described above.

The Ouralife 160C control blade has been evaluated to assure it has adequate structural margin under loading due to handling, and normal, emergency, and faulted operating modes. The loads evaluated include those due to normal operating transients (scram and jogging), pressure differentials, thermal gradients, seismic deflection, irradiation growth, and all other lateral and vertical loads expected for each condition. The Duralife 160C control blade stresses, strains, and cumulative fatigue have been evaluated and result in an acceptable margin to safety. The control blade insertion capability has been evaluated and found to be acceptable during all modes of plant operation within the Page 6 of 8

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limits of plant analyses. The Duralife 160C control blade coupling mechanism is equivalent to the original equipment coupling mechanism, and is therefore fully compatible with the existing control rod drives in the plant. In addition, the materials used in the Duralife 160C are compatible with the reactor environment. The impact of the increased weight of the control blades on the seismic and hydrodynamic load evaluation of the reactor vessel and internals has been evaluated and found to be negligible.

With the exception of the crevice-free structure and the extended handle, the Duralife 160C blades are equivalent to the NRC approved Hybrid I Control Blade Assembly (See Reload Summary Report Reference 9). The mechanical aspects of the crevice-free structure were approved by the NRC for all control blade designs in Reload Summary Report Reference 10. A neutronics evaluation of the crevice-free str ucture for the Duralife 160C design was performed by GE using the same methodology as was used for the Hybrid I control blades in Reload Summary Report Reference 9.

These calculations were performed for the original equipment control blades and the Duralife 160C control blades described above assuming an infinite array of ANF 9x9 fuel. The Duralife 160C control blade has a slightly higher worth than the original equipment design, but the increase in worth is within the criterion for nuclear interchangeability. The increase in blade worth has been taken into account in the appropriate U2C5 analyses. However, as stated in Reload Summary Report Reference 9, the current practice in the lattice physics methods is to model the original equipment all 6 C control blade as non-depleted. The effects of control blade depIetion on core neutronics during a cycle are small and are inherently taken into account by the generation of a target k-effective for each cycle. As discussed above, the neutronics calculations of the crevice-free structure, show that the non-depleted Dur alife 160C control blade has direct nuclear interchangeability with the non-depleted original equipment all BC design. The Duralife 160C also has the same end-of-life reactivity worth reduction limit as the all B~C design. Therefore, the Duralife 160C can be used without changing the current lattice physics model as previously ap'proved for the Hybrid I control blades (Reload Summary Report Reference 9).

The extended handle and the crevice-free structure features of the Duralife 160C control blades result in a one pound increase in the control blade weight over that of the Hybrid I blades, and a sixteen pound increase over the Susquehanna SES original equipment control blades. In Reload Summary Report Reference 9, the NRC approved the Hybrid I control blade which weighs less (by more than one pound) than the D lattice control blade. The basis of the Control Rod Drop Accident analysis continues to be conservative with respect to control rod drop speed since the Duralife 160C control blade weighs less than the D lattice control blades, and the heavier D lattice control blade speed is used in Page 7 of 8

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the analysis. In addition, GE performed scram time analyses and determined that the Duralife 160C control blade scram times are not significantly different than the original equipment control blade scram times. The current Susquehanna SES measured scram times also have considerable margin to the Technical Specification limits. Since the increase in weight of the Duralife 160C control blades does not significantly increase the measured scram speeds and the safety analyses which involve reactor scrams utilize either the Technical Specification limit scram times or a range of scram times up to and including the Technical Specification scram times, the operating limits are applicable to U2C5 with Duralife "

160C control blades.

Since the Duralife 160C control blades contain solid hafnium rods in locations where the B C tubes have'ailed, and the remaining B<C rods are manufactural with an improved tubing material (high purity stainless steel vs commercial purity stainless steel),

boron loss due to cracking is not expected. Therefore, the requirements of IE Bulletin 79-26, Revision 1 do not apply to the Duralife 160C control blades. However, PP&L plans to continue tracking the depletion of each control blade and discharge any control blade prior to a ten percent loss in reactivity worth.

Based on the discussion above, the new control blades proposed to be utilized in U2C5 do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The replacement blades can only be evaluated for their effectiveness as part of the overall reactivity control system, which is evaluated in terms of analytical consequences in 1.

above. Since they do not cause any significant change in system operation or function, no new events are created.

The analyses described in l. above indicate that the replacement blades meet all pertinent regulatory criteria for use in this application, and are expected to eliminate the boron loss concerns expressed in IE Bulletin 79-26, Revision 1. Therefore, the proposed change does not result in a significant decrease in any margin of safety.

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