Text

Proposed Tech Specs Relocating pressure-temp Curves, Predicted Radiation Induced NDTT Shift Curve & LTOP Limits to FCS Unit 1 RCS pressure-temp Limits Rept
	ML20195B444
Person / Time
Site:	Fort Calhoun
Issue date:	05/26/1999
From:	; OMAHA PUBLIC POWER DISTRICT
To:	;
Shared Package
ML20195B434	List: LIC-99-0045, Forwards Application for Amend to License DPR-40,proposing to Relocate pressure-temp Curves,Predicated NDTT Shift Curve & LTOP Limits & Values from FCS TS to OPPD Controlled Document.Rev 3 to TR CE NPSD-683 & Supporting TSs Encl; ML20195B437; ML20195B444; ML20195B452; ML20195B458;
References
	NUDOCS 9906010031
	Download: ML20195B444 (31)
	v • d • e

g TABLE OF CONTENTS (Continued) f.BRt 4.3 Nuclear Steam Supply System

.................4-3 4.3.1 Reactor Coolant Sy em (RCS)......................................... 4-3 4.3.2 Reactor Core and C trol..

......................................4-3 4.3.3 Emergency Core Co ing...................... -...................

4-3 4.4 Fuel Storage.................

4-4 4.4.1 New Fuel S torage.................................................... 4-4 4.4.2 Spent Fuel Storage................

.................................4-4 4.5 Seismic Design for Class I Systems......

.......... 4-5 5.0 ADMINISTRATIVE CONTROLS...................

..................5-1 5.1 Responsibility.......

................................................5-1 5.2 Org an iza tion........................................................

..... 5-1 5.3 Facility Staff Qualifications........................................... 5-la 5.4 Training..............................................................

. 5-3 5.5 Review and Audit........

. 5-3 5.5.1 Plant Review Committee (PRC)................................... 5-3 5.5.2 Safety Audit and Review Committee (SARC).............................. 5-5 5.6 Reportable Event Action....................

....................5-9 5.7 Safety Limit Violation......................

.. 5-9 5.8 Procedures...

5-9

{

5.9 Reporting Requirements...

5-10 l

5.9.1 Routine Reports...........................

...............5-10 5.9.2 Reportable Events..... -

... 5-12

.3 ep 5.9.

Unique Reporting equirements..............

.....................5-15 5.9.5 Core Operating Limits Report

....... 5-17a 5.9.6 RCS Pressure-Temperature Limits Report (PTLR);. g.... J. rgyiW.R...Q 5 17.b P

5.10 Re rds Retentio

. 5-18 5.11 adi ion Pr tio Program 5-5.12 DELE i

5.13 Secondary Water Chemistry.........................

...... 5-20 5.14 Systems integrity........

5-21 5.15 Post-Accident Radiological Sampling and Monitoring......

5-21 5.16 Radiological Effluents and Environmental Monitoring Programs.................... 5-22 5.16.1 Radioactive Effluent Controls Program.................

5-22 5.16.2 Radiological Environmental Monitoring Program

.......... 5-23 5.17 Offsite Dose Calculation Manual (ODCM)..................................... 5-24 5.18 Process ControI Program (PCP).........

.................. 5 -2 5 5.19 Containment Leakage Rate Testing Program........

......................5-26 i

6.0 INTERIM SPECI AL TECIINICAL SPECIFICATIONS 6-1 i

6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED

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l 9906010031 990526 PDR ADOCK 05000205 8

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7 TFCilNICAL SPECIFICATIONS - FIGURES TABLE OF CONTENTS PAGE WHICH FIGURE DESCRIPTION FIGURE FOLLOWS fety Li mp O i

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n.,co n.

r..,,..,s...,.......,,,,,....,........................................--2-6 s.

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ST Level v tored BAST Concentration......

1 2-12 Boric Acid Solubility in Water................................................ 2-19h 2-10 Spent Fuel Pool Region 2 Storage Criteria.

........................ 2-3 9 e 2-8 Flux Peaking Augmentation Factors............................................. 2-53 i

viii Amendment No.,.,.,,,,.,,<,,,..,,...,3,..,,,,..., n,,,,,., 1 8 8

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F:

DEFINITIONS t

Dose Equivalent I-131 (pCi/gm)

= pCi/gm ofI-131

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- + 0.0361 x pCi/gm ofI-132 l

+ 0.270 x pCi/gm ofI-133

+ 0.0169 x Ci/gm ofI-134

+ 0.0838 x pCi/gm ofI-135 l

B - Average Disintegration Enerav i

]

l B is the average (weighted in proportion to the concentration of each radionuclide in the reactor i

coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration, in MEV, for isotopes, other than iodines, with halflives greater than 15 minutes making up at least 95%

~ f the total non-iodine radioactivity in the coolant.-

o

' Offsite Dose Calculation Manual (ODCM) j The document (s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring Warn /High (trip) Alarm setpoints, and in the conduct of the Environmental

. Radiological Monitoring Program. The ODCM shall also contain:

i

.1)'

The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2)

Descriptions of the information that should be included in the Annual Radiological i

l Environmental Operating Reports and Annual Radioactive Efiluent Release Reports required by Specifications 5.9.4.a and 5.9.4.b.

~

1 Unrestricted Area-Any area at or bevond the site boundary access to which is not controlled by the licensee for purposes of protection ofindividuals from exposure to radiation and radioactive materials.

Core Operatine Limits Reoort (COLR) l I.

l

- The Core Operating Limits Report (COLR) is a Fort Calhoun Station Unit No. I specific document that provides core operating limits for the current operating cycle. These cycle-specific core operating limits sh'all be determined for each reload cycle in accordance with Section 5.9.5. Plant operation within these operating limits is addressed in the individual specifications.

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RC$1ressure-Temocrature Limits Report (PTLR) i lhe RCS PRESS _URE-TEMPERATURE LIMITS REPORT (PTLR) is,a, fluence dependent report providing Limiting Conditi.ons'.for, Operation for heatup, cooldown, inservice hydrostatig and leak testing, and core criticali.ty limits in the form of Pressure-Temperaturej (P;T) limits to ensure prevention of brittle fracture!?In addition, this report. establishes Limiting Conditions for, Operation Mich provide Low Temperature Oveipressure Protection (LTOP) t(assure the PK lim.its are not ' exceeded durityg the mostilimiting LTOP eventi!The P-T limits and LTOP criteria in the PTLR are applicable through,the Effective Full Poyer Years (EEPY),specified in,the PT_LRs NRC and ASM,E approved methodologies are used as the basis for the4,COs provided in the PTLR;j References (1) USAR, Section 7.2 (2) USAR, Section 7.3 8

Amendment No. 57,05,141,152,164 l

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2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Comoonents (Continued)

(5)

DELETED (6)

Both steam generators shall be filled above the low' steam generator water level trip set point and available to remove decay heat whenever the average temperature of the reactor coolant is above 300 F. Each steam generator shall be demonstrated operable by performance of the inservice inspection program specified in Section 3.17 prior to exceeding a reactor coolant temperature of 300 F.

(7)

Maximum reactor coolant system hydrostatic test pressure shall be 3125 psia. A maximum of 10 cycles of 3125 psia hydrostatic tests are allowed.

(8)

Reactor co ant system leak and by static est all be conducted within the limitations f the P,TLR Figma 2-1 A and 2-113.

(9)

Maximum sec a' -

t s e shall not exceed 1250 psia. A minimum measured temperature of 73 F is required. Only 10 cycles are permitted.

(10)

Maximum steam generator steam side leak test pressure shall not exceed 1000 psia. A minimum measured temperature of 73 F is r uired.

(11)

Ifno reactor coolant pumps operating, a non-operating reactor c p shall not be started while T, is be w 3652F ti,te tenm

-w listed in the PTLR unl ss at e

least one of the following c nditions is niet:

2-2a Amendment No. 3,55,55,71,11,125, M b188 I

2.0 I.IMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Operable Components [@W (a)

A minimum pressurizer steam space as specified in thJ PTLR of 53S0 by volume or gicatcr (50.5St er Icas actui! 151) exists, or l

j (b)

The temperature' difference between the steam generator secondary side and 1

the reactor coolant system cold leg 2...graturc is less than @e; magnitude ~

specified in the PTLR 30 T ibevc iat efic scacter ceci-at sysicm cold icg.

(12)

Reactor Coolant System Pressure solation Valves (a)

The integrity of all pressure isolation valves listed in Table 2.9 shall be demonstrated, except as specified in (b). Valve leakage shall not exceed the amounts indicated.

(b)

In the event that the integrity of any pressure isolation valve specified in Table 2-9 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a neafunctional valve are in and remain in the mode corresponding to the isolated condition. Manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supply deenergized.

(c)

If Specifications (a) and (b) above cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Basis The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation and maintain DNBR above 1.18 during all normal operations and anticipated transients.

When Specification 2.1.l(2) is applicable, the reactor coolant pumps (RCPs) are used to provide forced circulation heat removal during heatup and cooldown. Under these conditions, decay heat removal requirements are low enough that a single reactor coolant system (RCS) loop with one RCP is sufficient to remove core decay heat. However, two RCS loops are required to be OPERABLE to provide redundant paths for decay heat removal. Only one RCP needs to be OPERABLE to declare the associated RCS loop operable. Reactor coolant natural circulation is not normally used but is sufficient for core cooling. However, natural circulation does not provide turbulent flow conditions. Therefore, boron reduction in natural circulation is prohibited because mixing to obtain a homogeneous concentration in all portions j

of the RCS cannot be assured.

2-2b Amendment.No. 55,70,77,92,151,188 i

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r 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Onerable Components (Continued)

When Specification 2.1.l(3) is applicable, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be operable. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two shutdown cooling pumps to be OPERABLE.

One of the conditions for which Specification 2.1.1(3) is applicable is when the RCS temperature (T a) is less than 210 F, fuel is in the reactor, and all reactor vessel closure bolts eoi are fully tensioned. As soon as a reactor vessel head closure bolt is loosened, Specification 2.1.1(3) no longer applies, and Specification 2.8 is applicable. Specification 2.8 also requires two shutdown cooling loops to be operable if there is less than 23 feet of water above the top of the core.

The restrictions on availability of the containment spray pumps for shutdown cooling service ensure that the SI/CS pumps' suction header piping is not subjected to an unanalyzed condition in this mode. Analysis has determined that the minimum required RCS vent area is 47 in. This 2

requirement may be met by removal of the pressurizer manway which has a cross-sectional area greater than 47 in.

2 When reactor coolant boron concentration is being changed, the process must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the reactor coolant is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The low pressure safety injection pump will circulate the reactor coolant system volume in less than 35 minutes when operated at rated capacity. The pressurizer volume is relatively inactive; therefore, it will tend to have a boron concentration higher than the rest of the reactor coolant system during a dilution operation. Administrative procedures will j

provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the reactor coolant system during the addition of boron.W Both steam generators are required to be filled above the low steam generator water level trip set point whenever the temperature of the reactor coolant is greater than the design temperature

. heshutdowrrecolings '

. to ass - redundant heat removal system for the reactor.

l The LTOP enable temperature is documented in the PTLR has becn establishcd at T, - 3S5 *F.

The pressure transient analyses demonstrate that a single PORV is capable of mitigating overpressure events. Additional uncertainties have been applied to the Pressure-Temperature i

(P-T) limits to account for the case where a PORV is not available, (T,>3S5 *F) which is the rcasan for the discontinuity in thc P T Figsrcs. -Thc curvcs have bccn conscrvativcly smoothed for opcretions use-2-2c Amendment No. 55,71,136,161,188

2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.1 Ooerable Comoonents (Continued)

' The design cyclic transients for the reactor system are given in USAR Section 4.2.2. In addition, the steam generators are designed for additional conditions listed in USAR Section 4.3A. Flooded and pressurized conditions on the steam side assure minimum tube sheet temperature differential during leak testing. The minimum temperature for pressurizing the steam generator steam side is 70 F; in measuring this temperature, the instrument accuracy must be added to the 70 F; limit to determine the actual measured limit. The measured temperature limit will be 73 F based upon use of an instrument with a maximum inaccuracy of 12 F and an na 1*

' gi Formation o a 53% steam space of the magnitude,specifiedin,the sures that the f

resulting pr ssure mcrease would not result m any overpressurizais,PTIJL on s uld the first reactor coolant pum e

ed hen the rator crature is greater than that of the RCS cold eg.

eam space requirement is not applicable to the start of a react lant ump ' one or more pu are '

tio For the case in which the pressurizer steam space is less than the magnitude aposi6ed in;the PTLR 63%, limitation of the steam generator secondary side /RCS' cold leg AT id Jess than t_he magnitude specified in the PTLR-392F ensures that a single low setpoint PORV would preve an overpressurization due to actuation of the first reactor coolant pump. This requirement i ot app ' able to the start of a reactor coolant pump if one or more pumps are operating.

References (1) USAR Section 4.3.7 l

2-2d Amendment No. 55,71,135,151,188

J 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1.2 Heatuo and Cooldown Rate Anolicability Applies to the temperature change rates and pressure of the reactor coolant sys em (RCS).

Obiective To specify limiting conditions of the reactor coolant system heatup and cooldown rates.

Snecification The scacter cocian; y.casurc shall bc linsted d.d. g pl nt eg. Jon in accord nce veith Fig-e 2-1 A and 2-10 and as fellsvia: The combination of RCS' pressure; RCS temperature and RCS heatupland cooldown, rates lshall be, maintained,within the limitsispecified in the,PJLR andg designated below; p

=

(+)al Allowable combinations of pressure and temperature (T,) fo specific he at shall

~

to the right of the applicable limit lines as sho in the PTLR en e

w-F;gurc 2-1 A.

(9)b. Allowable combinations of pressure and temperature (T,) for a spe '

cool hati oc o and to the right of the applicable limit lines as sho in the PTLR en Figurca 2-10.

(3)c; The heatup rate of the pressurizer shall not exceed 100*F in any one hour period.

(4)d! The cooldown rate of the pressurizer shall not exceed 200 F in any one hour period.

Reg Actions (51) en any of the above limits are exceeded, the following corrective actions shall be taken:

(a)

Immediately initiate action to restore the temperature or pressure to within the limit.

(b)

Perform an analysis to determine the effects of the out oflimit condition on the fracture toughness properties of the reactor coolant system.

(c)

Determine that the reactor coolant s stem remains acceptable for continued n cold i hin 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(5)

Scforc the radia:!cn exposure cf the scactor vcasci excccds he caposure for vehich they apply, Figurca 2-1 A and 2-13 sha!! bc upda;cd in accordar.cc viith the fallsveing cri:cria and proccdurca:

W 2-3 Amendment No. 29 74; 161 7

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2.0 LIMITING CONDITIONS FOR OPERATION l

2.1 Reactor Coolant System (Continued) 2.1.2 Heatun and Cooldown Rate (Continued)

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Basis All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor coolant system temperature and pressure changes.W These cyclic loads are i ro ed b orm' nit 1 - transients, re trips d startu and sh own o eration.

During unit startup and shutdown, the rates of temperature and pressure changes are limited.

The design number of cycles for heatup and cooldown is based upon allowable heatup/cooldown rates and cyclic operation. Cycle' dependent informationlsuch as the pressure-t.emperature (P-T) limit. curves and low temperature overpressure protection (LTOP) system limits are contained in the Fort Calhoun Station RCS Pressurch Temperature Limits Report (PTLR), which was developed'using the meth'odologies of.CE NPSD483,W l

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2.0 LIMITING CONDITIONS FOR OPERATION i

2.1 Reactor Coolant System (Continued) 2.1.2 Heatuo and Cooldown Rate (Continued)

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2-7 Amendment No. -22,47,M,74,' ^0,161 A

g i

1 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) l 2.1.2 ' Heatun and Cooldown Rate Continued This :caperaturc is based on previous NDTT meiads. This ;cmpcreturc caricsponds to the mcasured 10 F NDTT of the rcactor icssci Sangc, which is not subject to redietion damagc, plus 60 F date sca;;ciin NOTT mcasur;nicnts, plus 12 F F

ms;rumen; circe E.

The tempera;urc at which thc hea:ap and cooldown ra cs change in Figurca 2-1 A and 2-1B rcaccis ac poin; at which ic mos; limi:ing hea:np and couldown ratcs wii j

respcc; ;c ic inic :capcre:urc (T,) ch=ge.

References:

)

(1)

USAR, Section 4.2.2 (2)

ASME Boiler and Pressure Ves.cel Code, Section 111 (3)

USAR, Section 4.2.4 (4)

USAF, Section 3.4.6 (5)

Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of Irradiated Capsule W-225, Revision 1, August 1980.

(6)

Technical Specification 2.3(3)

(7)

Article IWB-5000, ASME Boiler and Pressure Vessel Code,Section XI (8)

Omaha Public Power District, Fort Calhoun Station Unit No.1, Evaluation of adia (9)

Fort Calhoun. Station Unit NoJ1 RCS Pressure e Temperature Limits Report (PTLR)

(10)

CE NPSD.-683," Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP. Requirements from the Technical Speedications,'l (Lates_t Approved Revision)

A A

/>

l l

2-7a Amendment No. 22,47,54,74,100,161 L

Vh' 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 -

Teactor Coolant System (continued) -

.2.1.6 Weserizer and Main Steam Safety Valves -

Annlicability' f Applies to the status of the pressurizer and main steam safety valves.

Obiective To specify minimum requirements pertaining to the pressurizer and main steam safety valves.

Specifications I

To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be met:

(1):

The reactor shall not be made critical unless the two pressurizer safety valves are operable with l

their lift settings adjusted to ensure valve opening at 2485 psig 11% and 2530 psig.11%.0) 1 (2)-

Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety valve shall be installed on the pressurizer. However, when in at least the cold shutdown condition, safety valve nozzles may be open to containment atmos performance of safety valve tests or maintenance to satisfy this specification.pher At least four of the five Main Steam Safety Valves (MSSVs)ings shall be at 985 (3) associated with each steam generator shall be OPERABLE in MODES 1 and 2. Lift sett 1000 psig +3/-2%,1010 psig +3/-2%,1025 psig +3/-2%, and 1035 psig +3/-2%.

a..

With less than four of the five MSSVs associated with each steam generator -

OPERABLE, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .-

(4)l Two power-operated relief valves (PORVs) shall be operable durin heatups and cooldowns when the RCS temperature is less than 515'F, and in Modes 4 te on vessel and the RCS is not vented through a 0.94 are inch or larger vent, to

-l pre,ven,t vi tion of the pressure-temperature limits design ed by in theRJL 1"gu.;; 2-! A j

~ m.

j I

a.

With one PORV inoperable during heatups and cooldowns when CS temperature 1

is less than 515'F, restore the inoperable PORV to operable within 7 days or be in cold shutdown within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b.

With both PORVs inoperable during heatups and cooldowns when the RCS i

temperature is less than 515 *F, be in cold shutdown within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and depressurize and vent the RCS through at least a 0.94 square inch or larger vent within ol(owing 36 ho i

L c.

With one PORV inoperable in Modes 4 or 5, within one hour ensure the pressurizer -

l PTLR pace is, greater than the minimum volume for RCP 'startup as specaded in steam s L

3% m.u...c (50.5"C&lca;hiEt!Tand festoirs the mo rable PORV to

- bperable withi 7 days, ad ua team e canno establis ed within one h

r,

.n o

'o O

op th 2 I

c t

restor in the required time, depressurize and vent the R through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

l 2-15 Amendment No. 3,47,54,145,15:,189 i

L i

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ly

).

' 2.0 LIMITING CONDITIONS FOR OPERATION -

2.2 -

Chemical and Volume Control System

- 2.2.1. Boric Acid Flow Paths - Shutdown Anolicabidtv

~ Applies to the operational status of the boric acid flow paths in MODES 4 and 5 when fuel is in the reactor.-

Obiective -

To assure operability of equipment required to add negative reactivity.

Snecification As a minimum, one of the following boric acid flow paths from an OPERABLE borated water source shallbe OPERABLE:

a.

. A flow path from boric acid storage tank CH-11 A via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System, b.

A flow path from boric acid storage tank CH-11B via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System.

c.

A flow path from both boric acid storage tanks (CH-11 A and CH-11B) via either a boric acid trans ump or gravity feed connection and a charging pump to the Reactor Coolant System.

d.

A flow path from the SIRW tank via eit er a chargmg pump or a high pressure safety injection pump to the Reactor Coolant System. The. flow path fkom the SIRW tank to the'Ranctor Coolant Systemyia a~ single..,HPSI pump shall only,be established,if thepequhements in;tle PTLR;am met; Requir.gl Actions j

(1)

With none of the above boric acid flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

i e

i l

L 2-17 Amendment No.Bt,172 1

2.0 LIMITING CONDITIONS FOR OPERATION 2.2 Chemical and Volume Control System (Continued) l 2.2.3 Chareinz Pumos - Shutdown Anolicability 1

I Applies to the operational status of charging pumps in MODES 4 and 5 when fuel is in the reactor.

Obiective To assure operability of equipment required to add negative reactivity.

Specific l

At least one charging pump or one high pressure safety injection pump in the boric acid flow path l

required to be OPERABLE pursuant to Specification 2.2.1 shall be OPERABLE. The flow p.ath from j

the SIRW tank to the Reactor Coolant Systemyialsingle HPSI pump shall only be established if:the

~

i i

gxplitements in the PTLR are me{ ' '

{

l 4

u Actions

'(1)

With no charging pump or high pressure safety injection pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

1 l

l l

1 l

l l

1 2-19 Amendment No. Bh 172 i

1 l

l 1

i

w 2.0.. LIMITING CONDITIONS FOR OPERATION y

2.2 Chemical and Volume Control System (Continued)

Basis (Continued)

Charging Pumps Whenever the reactor coolant temperature (T,ou) is greater than or equal to 210 F, two charging pumps must 1.

be operable in order to ensure it is possible to inject concentrated boric acid into the reactor coolant system L

with an assumed single failure. With only one pump operable,72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the system to two operable charging pumps. This is consistent with the allowed outage time for the borated water sources and fl:w paths required during these modes.

In Modes 4 and 5 when fuel is in the reactor, only one charging pump or high pressure safety injection pump l

must be operable. This is consistent with the number of operable borated water sources and flow paths W--

e es u

is required in order to complete an operable flow path to the reactor coolant system, ere are additional restrictions on the use of high pressure safety injection pumps contained in gjeg[]g TM.c.:cel Spince:ier. 2.3 to sure that the reactor vessel is not overpressurized.

l-o account for temperature measurement uncertainty. An administrative procedure to monitor the temperature of the BASTS and boric acid system piping in the Auxiliary Building ensures that the temperature requirements of Figure 2-12 are met. Should the system temperature be unacceptable for operation at the current boric acid concentration, steps will be taken to reduce the boric acid l

concentration or raise the temperature of the system such that the concentration is within the acceptable range

. cf Figure 2-12.'

' The limits on component operability and the time periods for inoperability were selected on the basis of the

. redundancy indicated above and NUREG-0212 Revision 2. The allowed outage times for the various components are connistent such that a suppoit system has the same allowed outage time as the supported system.

. References

(1) USAR Section 9.2 I

2-19h Amendment No.172 e

r 2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emercency Core Cooling System (Continued) i (3)

Protection Acainst Low Temocrature Overnressurizatio

.rWW "W '

Operability requirements for HPSI pumps are provided in the RCS Pressure-Temperature Limits Report (PTLR) and shall be adhered to. If not in compliance with the PTLR, actions j

l shall be taken immediately to restore compliance to the PTLR. ThC felicveing limiting Condition 5 j

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(4) risodium Phosohate (TSPl Dodeca ivdrate l

During operating Modes 1 and 2, the TSP baskets shall contain 2110 fP of active TSP.

i 1

l a.

With the above TSP requirements not within limits, the TSP shall be restored within 72 l

hours.

t b.

With Specification 2.3(4)a required action and completion time not met, the plant shall l

be in hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Basis kC hih&i&ibs Ac

& r ci 5tETsiiig '6 Ink iCdC'6ci 15 'ic 4ii5's liCL'6 'sI.C 11WtGr bccian's '6c iiCiir GpbAii'tisig

^

CnipCis:UTC by ruiiiiing thC iCaC:Gr Cecisnt pump 5. ThC iCsC Or isiliCn iiisdC Critical by icithdreviing CEA'5 sad diluting bcron in-thC rCaC cr Ccciant-With thi5 modC cf 5:arbup, thC Cncigy 5:crCd in the rCaC:or Ccclant during thC epproaCh to Criticality ;5 5ubstantia!!y Cqual to-ths: during povcCr opCration Rnd thCiCfoiC LII CrigisiCCiCd 5LfC!y fettttWC5 &nd LUXiliHry CcGliiig 5y3:Ciii5 arc iCQUliCd-tc DC [UIfy

_____L1_

Ugibh uLin k.

2-22 Amendment No. H,39,43,47,64,H,M, I nn i n, 1,,

,.1,

1,,

,zi A V V, a V.7, A.I I, s r 3, a i i, a u a, no

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2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Emergency Core Cooline System (Continued)

References

- (1) USAR, Section 14.15.1 l

(2) USAR, Section 6.2.3.1 '

t (3) USAR, Section 14.15.3 (4) USAR, Appendix K 5)

Power Dist ' t's Submittal, Dec nb 1976 (6) Tcchical S cdScgion 2.1.2, Figure 2-1B DELETED

(

USAR, Section 4.4.3 1

l a

i l

i 2-23b Amendment No. 47,64,74,179 l

3.0 SURVEILLANCE REOUIREMENTS 3.3 Reactor Coolant System and Other Components Subiect to ASME XI Boiler & Pressure Vessel Code Inspection and Testine Surveillance i

Aonlicability Applies to in-service surveillance of primary system components and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.

Obiective i

To ensure the integrity of the reactor coolant system and other components subject to inspection and testing according to ASME XI Boiler & Pressure Vessel Code.

Snecifications (1)

Surveillance of the ASME Code Class 1,2 and 3 systems, except the steam generator tubes inspection, should be covered by ASME XI Boiler & Pressure Vessel Code.

1 4

In-service inspection of ASME Code Class 1, Class 2, and Class 3 components, a.

including applicable supports, and in-service testing of ASME Code Class 1, Class 2, and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, as required by 10 CFR Part 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a (g)(6)(i).

b.

Surveillance of the reactor coolant pump flywheels shall be performed as indicated in Table 3-6.

. n hanical and c.

A surveillance program to monitor

' tio 'nducall be m,geurtt e impact properties of the reacto essel m eria s aintaine in ac or with art 50 Appendix H. 4 Examination results shall be tised to update,the F

2 P.

R' (2)

Surveil anc of Reactor Coolant System Pressure Isolation Valves Periodic leakage testing

on each valve listed in Table 2 9 shall be accomplis.hed prior a.

to entering the power operation mode every time the plant is placed in the cdd shutdown

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

3-21 Amendment No. 45,75,104,142,157,176

~ k. N h(N 7

5.0 ADMINISTRATIVE CONTROLS 5.9.6 - Reactor Coolant System (RCS) Pressure - Temnerature Limits Report (PTLR)

a. k Reactor Coblant System pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: Technical Specifications 2.1.1, 2.1.2, 2.1.6, 2.2.1, 2.2.3, and 2.3.

b.

The analytical methods used to detennine the RCS pressure and temperature limits and predicted radiation induced NDTT shift shall be those previously reviewed and approved by the NRC, specifically those described in the following documents

1.

10 CFR 50.61," Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events" 2.

10 CFR Part 50, Appendix G," Fracture Toughness Requirements" 3.

10 CFR Part 50, Appendix H," Reactor Vessel Material Surveillance Program Requirements" 4.

Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials,"

Revision 2,05/88 5.

ASME Boiler and Pressure Vessel Code Section III, Appendix G," Protection Against Nonductile Failure,"1986 Edition 6.

CE NPSD-683, " Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications,"

(Latest Approved Revision) 7.

WCAP-14040-NP-A (Section 2.2; Neutron Fluence Calculation), " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996 L

'c.

The PTLR shall be provided to the NRC upon issuance for each reactor wssel fluence period (i.e., the number of EFPY used in the P-T Limit /LTOP analyses) and for any revision or supplement thereto.

i i

5-17b Amendment No.

hC.V

U.S. Nuclear Regulatory Commission LIC-99-0045 ATTACHMENT B i

I

DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATION DISCUSSION AND JUSTIFICATION Omaha Public Power District (OPPD) is proposing revisions to the Fort Calhoun Station Unit No.1 Tecimical Specifications (TS) in accordance with Generic Letter (GL) 96-03," Relocation ofthe Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," dated January 31,1996. The proposed changes are consistent with the recommendations ofCombustion Engineering Owners Group (CEOG) Task 942," Development ofa RCS Pressure and Temperature Limits Report for the Removal ofP-T Limits and LTOP Requirements from the Technical Specifications," CE NPSD-683.

Accordingly, Topical Report CE NPSD-683 is included for NRC approval as a lead CEOG plant submittal. These documents supersede those provided by OPPD's letter (S. K. Gambhir) to the NRC (Document Control Desk) dated January 30,1998 (LIC-98-0013). The neutron fluence analysis methodology employed is described in Attachment C of OPPD's letter (S. K. Gambhir) to the NRC (Document Control Desk) dated January 30,1998 (LIC-98-0009). The Fort Calhoun Station neutron fluence analysis was performed by Westinghouse using the methods ofWCAP-14040-NP-A and Draft Regulatory Guide-1053, with the ENDF/B-VI Cross-Section Library.

The pressure-temperature (P-T) curves (Figures 2-1 A & 2-1 B), h predicted radiation induced NDTT shifi curve (Figure 2-3) and the low temperature overpmssure p totection (LTOP) limits (TS 2.3(3)) are proposed for relocation to a document entitled " Fort Calhoun Station Unit No.1 Reactor Coolant System (RCS) Pressure-Temperature Limits Report (PTLR)." To ensure that the RCS is not over pressurized when the plant is in Mode 4 or Mode 5 with fuel in the reactor, TS 2.2.1.d and TS 2.2.3 are being revised so that PTLR requirements must be met when a flowpath from the safety inj ection refueling water (SIRW) tank to the RCS is established via a single HPSI pump.

A definition ofthe RCS PTLR is being added to the TS as well as a new administrative contml (TS 5.9.6).

TS 5.9.6 establishes the scope ofthe PTLR, the NRC and ASME approved analytical methods utilized inthePTLR,andrequirementsforsubmittingPTLRrevisions. Additionaladministrativerevisionsarealso proposed, which include relocating certain specific values (e.g., minimum pressurizer steam space) to the PTLR, relocating most ofthe Basis ofTS 2.1.2 to the PTLR, relocating statements from the Basis ofTS 2.3 conceming reactor startup using the reactor coolant pumps to the PTLR, and adding the PTLR and Topical Report CE NPSD-683 as references in TS 2.1.2.

It should be noted that Topical Report CE NPSD-683 references the use of the ABB/CE codes ROCS /MC. OPPD used the ROCS /MC codes during the time period (i.e.,1990) in which the attached PTLR heatup/cooldown limit curves were generated. However, OPPD has upgraded to the CASMO/ SIMULATE codes as contained in Topical Report OPPD-NA-8302-P, Rev. 04, dated May 1994 (Reload Analysis Neutronics Methodology). NRC approval ofTopical Report OPPD-NA-8302-P, Rev. 04 is documented in a letter dated December 16,1994, from S. D. Bloom (NRC) to T. L. Patterson (OPPD). Therefore, future use ofthe CASMO/ SIMULATE codes by OPPD is considered equivalent to the use of the ROCS /MC codes described in Topical Report CE NPSD-683, t

I 1

DISCUSSION AND JUSTIFICATION (Continued)

As required by GL 96-03, OPPD utilizes NRC approved methodology (ASME Section III, Appendix G) to derive the parameters used to construct the P-T curves and LTOP setpoints. The NDTT shifi curve is also derived using NRC approved methodology (Regulatory Guide 1.99, Revision 2).

As described in OPPD's " Application for Amendment of0perating License"(S. K. Gambhir) to the NRC (Document Control Desk) dated June 1,1992, (LIC-92-157A), ABB-CE performed an analysis for Fort Calhoun Station for P-T limits and LTOP system requirements for continued operation thmugh 20 efrective full poweryears. The P-T limits were calculated to meet the regulations of10 CFR 50 Appendix A, Design I

Criterion 14 and Design Criterion 31. The limits were developed using the requirements of10 CFR 50, Appendix G and ASME Section III, Appendix G. In a letter from the NRC (S. D. Bloom) to OPPD (T.

L. Patterson) dated March 23,1994, the NRC approved the amendment request and issued Amendment 161. Note that OPPD is not requesting an exemption for the use ofeither ASME Code Case 514 or 636 at this time but may elect to do so in the future. Such a change will be submitted to the NRC for approval if OPPD elects to use one of these code cases.

The proposed amendment will reduce the burden on OPPD and NRC resources by eliminating the necessity ofpmcessing an amendment request each time a change is made to P-T limits, NDTT shift cun'es or LTOP setpoints. Future changes to the PTLR will be made using the latest revision ofCE NPSD-683 and will be controlled by the requirements of10 CFR 50.59 (similar to the Core Operating Limits Report).

Thus, future changes to the PTLR will not normally require a license amendment to be effective.

l v

i i

2

g l>

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:

. The proposed changes to the Fort Calhoun Station (FCS) Unit No.1 Technical Specifications (TS) are

in accordance with Generic Letter (GL) 96-03," Relocation ofthe Pressure Temperature Limit Curves and L

' Low Temperature Overpressure Protection System Limits," dated January 31,19%. The proposed j

changes are also consistent with the recommendations ofCombustion Engineering Owners Gmup (CEOG)

Task 942," Development ofa RCS Pressure and Temperature Limits Report for the Removal ofP-T Limits and LTOP Requirements fmm the Technical Specifications," CE NPSD-683. The proposed changes do not involve significant hazards consideration because operation ofFCS in accordance with these changes would not:

(1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

l The proposed changes relocate the reactor coolant system (RCS) pressure-temperature (P-T) curves, the predicted radiation induced NDTT shift curve and the low temperature overpressure protection (LTOP) limits to the Fort Calhoun Station Unit No.1 RCS Pressure-Temperature Limits Report (PTLR). Compliance with these curves and limits continues to be required by the Technical Specifications. Changes to the curves and limits will be contmiled by TS 5.9.6, and must i.

bein accordance with the NRC and ASME approved methodologies listed there and with 10 CFR 50.59.

The FCS PTLR in combination with the limitations imposed by the TS,will ensure the integrity of the reactor vessel pressure boundary. Therefoie, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

(2)'

Create the possibility ofa new or difTerent kind ofaccident from any accident previously evaluated.

There will be no physical alterations to the plant configuration (no new or difTerent equipment is being installed). No changes in operating modes or limits are proposed. The TS retain requirements to maintain the RCS within acceptable operational limits established in accordance with NRC and ASME approved methodologies and assure operability ofthe LTOP system. As such, the TS will continue to requite compliance with the limitations being relocated to the FCS PTLR. Therefore, these proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated, i-3-

q

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

(3)

Involve a significant reduction in a margin of safety.

This proposed change to the FCS TS is administrative in nature relocating the P-T curves, NDTT curve, LTOP limits and associated TS requirements to the FCS PTLR in accordance with GL 96-

03. Future updates ofthe FCS PTLR will be conducted under the 10 CFR 50.59 process utilizing NRC and ASME approved methodologies (as described in FCS Unit No.1 PTLR, Rev. O and CEOG Topical Report CE NPSD-683). Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above considerations, the proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 and the proposed changes will not result in a condition which significantly alters the impact ofthe Station on the environment. Thus, the pmposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.

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