ML20198P121

Proposed Tech Specs to Incorporate Revised Heatup & Cooldown Limit Curves & Revised Cold Overpressure Mitigation Sys (COMS) PORV

ML20198P121
Person / Time
Site:	Wolf Creek
Issue date:	12/29/1998
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20198P119	List: ML20198P116 ML20198P121
References
NUDOCS 9901060328
Download: ML20198P121 (29)

Text

. .. - . - - - .- . - ~ - - - . . . . = . . . . . . ~ . - . - . - , _ - - - . - , . .

l Attachment V to WO 98-0104 Page 1'of 16 l

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ATTACHMENT V 2 PROPOSED TECHNICAL SPECIFICATION CHANGES i

l CURRENT TECHNICAL SPECIFICATIONS j 1

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I l 9901060328 991229 9 i PDR ADOCK 05000482 1

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l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE i TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. . . . . . 3/4 4-21 3/4.4.7 CHEMISTRY . . . .. ... . . DELETED TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS . . .. . . . . DELETED TABLE 4.4-3 REAC TOR COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS . . . . . . DELETED 3/4.4.8 SPECIFIC ACTIVITY. . . . . 3/4 4-25 i

FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY

>1 Ci/ GRAM DOSE EQUIVALENT l 131.. .. . . 3/4 4-27 l l

TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM . . 3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System . . 3/4 4-29 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS  !

APPLICABLE UP TO $3420 EFPY . . . . . 3/4 4-30 l l l

FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO $3420 EFPY . .. . . 3/4 4-31 l TABLE 4.4-5 DELETED Pressurizer . . .. . DELETED Overpressure Protection Systems . . 3/4 4-34 FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM . .. . 3/4 4-36 3/4.4.10 STRUCTURAL INTEGRITY .. .. . DELETED 3/4.4.11 REACTOR COOLANT SYSTEM VENTS . DELETED 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS . . . 3/4 5-1 WOLF CREEK- UNIT 1 Vlli Amendment No. 40, E, 74,8% l l

Ap/tru MATERIAL PROPERTY BASIS Wli k Controlling Material:

Copper Content:

RV Lower Shell , ) p (, .

0.07 Weight %

, Nickel Content: 0.62 Weight %

initial RTNOT: 40'F Limiting ART after 13.6 EFPY: 1/4T,89 'F 3/4T,79 'F 3.000.- *

- CRITICAUTY IT .......

. . . . . . . . _ . . . . . . . .. . .. mASED ON FMR . . . .

.. . . . . . . . . . . . . . . . . . . . HEATUP C . , , , .

2.500 - - - - - --  !

-/.- . . .

LEAK TEST LIMIT . . . . . ..

\ . . . . .

9 2.000

. v: -

u) e

~

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- CRITICALITYLIMIT

.[.

$ 1.500 60 'FMR mASED ON 100*FMR

c. -

HEATUP CURVE . . HEATUP CURVE ..

C w . . . . .

q -

, g -

gS,oog

~ ~ ~ '

CRITICALITY LIMIT

~ "-"

• 9ASED ON INSERVICE 500 - - -- '

3ng . ppg HYDROSTATIC TEST HEATUP CURVE

[ TEMPERATURE (222*F)

FOR THE SERVICE

-. . . . _.. . _.... . . .. PERIOD UP TO 13.6 EFPY . .. . .

0 '

0 100 200' 300 400 500 L INDICATED TEMPERATURE (DEG. F)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 13.6 EFPY W01.F CREEK - UNIT 1 3/4 4-30 Amendment No. 40,71 L ._ _ . . _ .

l MATERIAL PROPERTY BASIS Controlling Material: RV Lower Shell Copper Content: 0.087 Weight %

Nickel Content: 0.576 Weight %  ;

Initial RTNDT: 40*F l l Limiting ART after 20 EFPY: 1/4T, 90*F 3/4T, 80*F l 2500 l LEAK TEST LIM'T -

2250 - .

i 2000 - . CRimun tiMiT

" BASED ON 60*F/HR ^

_ HEATUP CURVE O '

{ l750 -

CRmCAUTY UMIT' w . BASED ON 100*F/HR i 1,u . HEATUP CURVE

$ 1500 - -

)

m \

W 60*F/HR -

HEATUP CURVE E 1250 - l Q

< 1000 -

0 Q ,

z 750 -  !

i 500 - -

CRiTiCAuTY tJMIT BASED ON INSERVICE HYDROSTATIC TEST-100*F/HR TEMPERATURE (223*F) 250 - "HEATUP CURVE' FOR THE SERVICE PERIOD UP YO 20 EFPY 0 , , , , , , , , ,

0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG F.)

FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 20 EFPY l

WOLF CREEK - UNIT 1 3/4 4-30 Amendment No. 40, M, _.,

i

, . .- - _. _ _ . _ . _ . -- -- - - - ~~'

\

t L MATERIAL PROPERTY BASIS pk Controlling Materi l:

Copper Content:

RV Lower Shell 0.07 Weight %

[

i Nickel Content:

Initia; RTwot:

0.62 Weight % [ k.d (4 40'F 0 {t Limiting ' TT after 13 6 EFPY: 1/4T,89 'F 3/4T, 79 'F 3,000 2.500' - - - - -

9 m

2.000 -

~

C_. -

Lu Cf D

m _

m W 1,500 -

c: - -

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3 COO N N ES Z 1 000

( *F!HR) 0

.. 20 .

40 g 60+ ,

500 -

3 o0 + [~~ - -

0 ' '

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[ 100 200 300 400 500 INDICATED TEMPERATURE (DEG. F)

FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE UP TO 13.6 EFPY WOLF CREEK - UNIT 1 3/4 4-31 Amendment No. 40,71

MATERIAL PROPERTY BASIS

Controlling Material

RV Lower Shell

, Copper Content: 0.087 Weight %

l Nickel Content: 0.576 Weight %

! Initial RTNOT: 40'F l Limiting ART after 20 EFPY: 1/4T, 90*F 3/4T, 80'F i

2500 ^

- ~

2250 - -

1 2000 - - - --

e 1750 -

in -

Lu g 1500 - -

E 1250 - -

CL.

O '

1000 - ~,

- MCOOLDOWN RATES

@ 750 - - ("F/HR)-

500 - '$

= . .

250 - -

0 , , , , , , , , ,

0 50 100 150 200 250 300 350 400 450 500 l

INDICATED TEMPERATURE (DEG F.)

l FIGURE 3.4-3 l REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS l APPLICABLE UP TO 20 EFPY l

l WOi.F CREEK - UNIT 1 3/4 4-31 Amendment No. 40; 1, ___

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, 2,750 -

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2,500 -- . .

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- . . . 97 485 .. ...... . .

127 485 . ~ . - - .- .

' 2,000 -. 177 das . . .

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227 545 4

-o 277 745 .

327 1310 l  ! 1.750 372. 2335

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1,000 -

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750 -

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500 -. . -

250 - -

-* = , . . .

0 100 200 300 400 500 MEASURED RTD TEMPERATURE (DEG. F)

FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM WOLF CREEK - UNIT I 3/4 4-36 Amendment No. 40,71

2500 2000 -- -

TRTD PSMAX

(*F) (PSIG) 60 469 78 469

^

88 493

$ 118 488 W 158 488 E1500 168 529 F- 218 540 Z 268 650 0 318 800 Q. 343 910 y) 368 418 1127 2350

> 1000 - - -

0:

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500 -

0 0 50 100 150 200 250 300 350 400 450 i MEASURED RTD TEMPERATURE (*F) 2 RCPs running below 100*F 4 RCPs running above 100*F FIGURE 3.4-4 MAXIMUM ALLOWED PORV SETPOINT FOR THE COLD OVERPRESSURE MITIGATION SYSTEM i

WOLF CREEK UNIT 1 3/4 4-36 Amendment No. 4 7+, __ l

REACTOR COOLANT SYSTEM BASES PRESSURFJTEMPERATI 'RE LIMITS (Continued)

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below.

3. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

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g ...:.6 .wg 4 gy unm.7, a aag.,93 .c em-.:gm tu g reko a enag om:rm, 7ma o,,,~ g,7 si g ..mt ggag.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of.13.S 20 effective full power years (EFPY) of service life. The .13.6 20 EFPY service life period is chosen such that the 1:miting RTNDT at the 1/4T location in the core region is greater than the RTNDT of the limiting unirradiated material. The selection of such a limiting RTNDT assures that all components in the Reactor Coolant System will be operated c nservatively in accordance with applicable Code requirements. )

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.41. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTNDT. Therefore, an adjusted reference temperature, based upon the fluence and copper content and nickel content of the material in question, can be predicted using Figure B 3/4.41 and the largest value of ARTNDT computed by Regulatory Guide 1.99, Revision 2," Radiation Embrittlement of Reactor Vessel Materials." The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNDT at the end of.13,6 20 EFPY. er c" er edprt e-*r 'e pc~ e e~^r !" '"e pr-~" e e-d te pe etu e re r!";!"rt- e"*r These limit curves were generated without margins for instrumentation errors and are consistent with the methodology presented in WCAP-14040-NP-A, Revision 2.

l

!- WOLF CREEK , UNIT 1 Amendment No. 40,Z1,89 B 3/4 4-7 g

l l

l REACTOR COOLANT SYSTEM l

BASES 1 PRESSURE / TEMPERATURE LIMITS (Continued) .

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..a ..,~gt,.; .- p gru g 4 a s:, ,3 n :'-- Capsules wenLbe are removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can ,

be used to predict the future radiation damage to the reactor vessel material I by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods erived from Appendix G in Section lil Of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section ill as the reference flaw, amply exceed the current capabilities ofinservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against nonductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility

. reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K ,

for the metal temperature at that time. K,is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K, curve is given by the equation:

K, = 26.78 + 1.223 exp [0.0145(T-RT, + 160)) (1)

! WOLF CREEK - UNIT 1 B 3/4 4-8 Amendment No. 40,43 l

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kbV'k W -$

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INNER SURFACE

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I 3/4 T 6.69 x 10" Lu .

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O 5 10 15 20 25 30 32 35 SERVICE UFE (EFPY)

FIGURE B 3/4.41 FAST EUTRON FLUENCE (E>1MeV) AS A FUNCTION OF EFFECTIVE FULL POWER LIFE WOLF CREEK - UNIT 1 B 3/4 4 11 Amendment No. 44,71

1.0E+20 Fu hNER Supppcy k

e so 1*0E+19  :

1 ............. f/4 T = y 38E+g y

........... 3I4 Y n 4.syg,,y 2 "'"

o 1DE+18  :

W 2

Actual Projected 1.0E+17 0 5 10 15 20 25 30 35 40 SERVICE LIFE (EFPY)

FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCITON OF EFFECTIVE FULL POWER LIFE l

WOLF CREEK- UNIT 1 B 3/4 4-11 Amendment No. 4tT,7T, _

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) enm.a ; .se ;e e.e.:~ e< p.. .. ..

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The steady-state heatup and cooldown curves are used as the basis for developing the COMS setpoints. This approach is acceptable since COMS events are most likely to occur when the reactor vessel is at isothermal conditions. In addition, adjustments for possible errors In the pressure and temperature sensing instruments are accounted for in the determination of the COMS PORV Setpoint Limit Curve (Figure 3.4-4) This approach is also consistent with the approved method documented in WCAP-14040-NP A, Revision 2.

The OPERABILITY of two PORVs, or two RHR suction relief valves, or an RCS vent opening of at least 2 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 368'F, Either PORV or either RHR suction relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50 F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and/or the normal charging pump and its injection into a water solid RCS l In addition to opening RCS vents to meet the requirement of Specification 3.4.9.3c., it is acceptable to remove a pressurizer Code safety valve, open a PORV block valve and remove power from the valve operator in conjunction with disassembly of a PORV and removal of its intemals, or otherwise open the RCS.

l WOLF CREEK- UNIT 1 B 3/4 4-13 Amendment No.19, '9, 99, l

4 l

l

l 1

REACTOR COOLANT SYSTEM BASES COLD OVERPRESSURE The Maximum Allowed PORV Setpoint for thn Cold Overpressure Mitigation System (COMS)is derived by analysis which mod mformance of the COMS assuming various mass input and heat input tranwn Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for: (1) process and instru-mentation uncertainties; and (2) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of both Safety injection pumps and all but one centrifugal charging pump and the normal charging pump while in MODES 4,5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary coolant temperature is more than 50'F l above reactor coolant temperature. Exceptions to these requirements are accept-able as described below.

Operation above 350 F but less than 375 F with the normal charging pump and only one centrifugal charging pump OPERABLE and no Safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

As shown by analysis LOCAs occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time duration and the condition of hav;ng only one centrifugal charging pump OPERABLE is allowed and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed.

Operation below 350 F but greater than 325 F with the normal charging pump and all centrifugal charging and Safety Injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During low pressure, low temperature operation all automatic Safety injection actuation signals except Containment Pressure - High are blocked. In normal conditions a single failure of the ESF actuation circuitry will result in the starting of at most one train of Safety injection (one centrifugal charging pump, and one Safety injection pump). For temperatures above 325 F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORVs without exceeding Appendix G limit. Given the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed.

Initiation of both trains of Safety injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents.

Although COMS is required to be OPERABLE when RCS temperature is less than 368 F, operation with the normal charging pump and all centrifugal charging pumps and both Safety injection pumps OPERABLE is acceptable when RCS temperature is greater than 350 F. Should l an inadvertent Safety Injection occur above 350*F, a single PORV has sufficient capacity to relieve the combined flow rate of all pumps. Above 350 F two RCPs and all pressure safety valves are required to be OPERABLE. Operation of an WOLF CREEK - UNIT 1 B 3/4 4-14 Amendment No. 40 l

i l

l l BASES-COLD OVERPRESSURE (Continued)

RCP eliminates the possibility of a 50 F difference existing between indicated and actual RCS temperature as a result of heat transport effects. Considering

- instrument uncertainties only, an indicated RCS temperature of 350 F is sufficiently high to allow full RCS pressurization in accordance with l

- Appendix G limitations. Should an overpressure event occur in these

! conditions, the pressurizer safety valves provide acceptable and redundant overpressure protection.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation

' System 6 is updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR l Part 50, Appendix H.

3/4.4.10 DELETED

- 3/4.4.11 DELETED t

WOLF CREEK - UNIT 1 B 3/4 4-15 ' Amendment No. 40,6440 l

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.~.-~ .- . . ___-_ _. . _ _ _ _ _ _ - -. _ ._._ _. - . _ _

J EMERGENCY CORE COOLING SYSTEMS 1

BASES I

-l ECCS SUBSYSTEMS (Continued) ,  !

The limitation insednamunwa of one centrifugal charging pump and the normal charging pump to be .

, OPERABLE and the Surveillance Requirements to verify all charging pumps except l

'3  ;

the required OPERABLE charging pump to be inoperable in MODES 4 and 5 and in - i MODE 6 with the reactor vessel head on, provides assurance that a mass  ;

. addition pressure transient can be relieved by the operation of a single PORV -

l or RHR suction relief valve. In addition, the requirement to verify all j Safety injection pumps to be inoperable in MODE 4, in MODE 5 with the water -

level above the top of the rancier vessel flange, and in MODE 6 with the

' reactor vessel head on and with water level above the top of the reactor vessel flange, provides assurance that the mass addition can be relieved by a single PORV or RHR suction reRef valve.

L With the water level not above the top of the reactor vessel flange and with the vessel head on, Safety injection pumps may be available to mitigate the affects of a loss of decay hett removal during a reduced RCS inventory

' condition.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety ,

. analyses are met and that subsystem OPERABILITY is maintained. Surveillance  !

Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a i

LOCAc Maintenance of proper flow resistance and pressure drop in the piping '

system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance I

- configuration, (2) provide the proper flow split between injection points in

. accordance with the assumptions used in the ECCS-LOCA analyses l and

, (3) provide C acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS LOCA analyses. The Surveillance Requirements for leakage testing of ECCS check valves ensures that a failure of one valve will not cause an intersystem LOCA. The Surveillance -

Requirements to vent the RHR and Si pump casings and accessible, i.e., can be ,

reached without personnel hazard or high radiation dose, ECCS discharge piping  ;

L; ensures against inoperable pumps caused by gas binding or water hammer in ECCS piping.

3/4.5.5 REFUELING WATER STORAGE TANK

~

The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient noply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and

, (2) the reactor will remain subcritical in the cold condition following mixing l . of the RWST and the RCS water volumes assuming all the control rods are out of

. the core. These assumptions are consistent with the LOCA analysesc I

t.

WOLF CREEK- UNIT 1 B 3/4 5-2 Amendment No. ??, ?S, l i.

4 l h.

_.. .. - . _. ~ _- . - .

Attachment VI to WO 98-0104 Page 1 of 9 ATTACHMENT VI PROPOSED TECHNICAL SPECIFICATION CHANGES IMPROVED TECHNICAL SPECIFICATIONS

LTOP System

, B 3.4.12 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls RCS pressure at low temperatures so the integrity of the reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (P/T) limits of 10 CFR 50. Appendix G (Ref.1). The reactor vessel is the limiting RCPB component for demonstrating such protection.

The PTLR provides the maximum allowable actuation logic setpoints for the power operate:. relief valves (PORVs) and the maximum RCS pressure for the existing RCS cold leg temperature during cooldown, shutdown, and heatup to meet the Reference 1 requirements during the LTOP MODES.

The reactor vessel material is less tough at low temperatures than at normal operating temperature. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). RCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only while shutdown: a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vessel. LCO 3.4.3. "RCS Pressure and Temperature (P/T) Limits,"

requires administrative control of RCS pressure and temperature during heatup and cooldown to prevent exceeding the PTLR limits.

This LCO provides RCS overpressure protection by having a minimum coolant input capability and having adequate pressure relief capacity. Limiting c lant '

input capability requires both safety injection pumps and one centrifugal charging pump to be l

- incapable of injection into the RCS and isolating the

,_ g 4g/( t accumulators.3 The pressure relief capacity requires either two l k

Og N ~ redundant RCS relief valves or a depressurized RCS and an RCS vent of sufficient size. One RCS relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.

(continued)

WCGS ITS BASES B 3.4-63 5/15/97

INSERT B3.4-63 The normal charging pump (NCP) flow, in addition to one centrifugal charging pump (CCP) flow, has been included in the analysis of design basis mass input overpressure transient. The term " centrifugal charging pump" or "CCP" refers to the safety-related ECCS pumps only.

Ii

LTOP-System l B 3.4.12 l l

BASES-BACKGROUND With minimin coolant input capability, the ability to provide <

(continued) core coolant addition is restricted. The LCO does not require l

the makeup control system deactivated or the safety injection j

-(SI) actuation circuits blocked. Due to the lower pressures in  ;

the LTOP MODES cnd the expected core decay heat levels, the  !

makeup system can provide adequate flow via the makeup control ,

valve. If conditions require the use of more than one l

centrifugal hagging pump for makeup in the event of loss of  ;

inventory, pimps can be made available throy manual l actions. e gdMo7eWcccf The LTOP System for pressure relief consists of two PORVs with reduced lift settings, or two residual heat removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, or a j depressurized RCS and an RCS vent of sufficient size. Two RCS l

-relief valves are required for redundancy. One RCS relief valve j has adequate relieving capability to prevent overpressurization i for the required coolant input capability.

l The norma charging (NCP) I rendere ncapable injecti j j into t RCS unde adninistr ive contr s, when a RCS col eg  !

t ature is 368'F. is ensur that the c rent LT alysis r .ns boundi .

l PORV Reauiranents l 1

As designed for the LTOP System, each PORV is signaled to open if 4 the RCS pressure appnathes a limit determined by the LTOP l actuation logic. The LTO? actuation logic monitors both RCS temperature and RCS pressure and determines when a condit%n not acceptable with respect to the PTLR limits is approached. The '

wide range RCS temperature indications are auctioneered to select the lowest temperature signal. ,

The lowest temperature signal is processed through a function generator that calculates a pressure limit for that temperature.

l' The calculated pressure limit is then compared with the indicated <

l RCS pressure from a wide range pressure channel. If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.

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WCGS ITS BASES B 3,4 64 5/15/97 l

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LTOP System l

B 3.4.12  ;

1 BASES BACKGROUND PORV Reauirements (continued) 1 The PTLR presents the PORV setpoints for LTOP. The setpoints are l normally staggered so only one valve opens during a low temperature overpressure transient. Having the setpoints of both valves within the limits in the PTLR ensures that the Reference 1 limits will not be exceeded in any analyzed event.

1 When a PORV is opened in an increasing pressure transient, the i release of coolant will cause the pressure increase to slow and l reverse. As the PORV releases coolant, the RCS pressure '

decreases until a reset pressure is reached and the valve is signaled to close. The pressure continues to decrease below the i reset pressure as the valve closes.

RHR Suction Relief Valve Recuirements l During LTOP H0 DES, the RHR System is operated for decay heat I removal and low pressure letdown control. Therefore, the RHR

suction isolation valves are open in the piping from the RCS hot

! legs to the inlets of the RHR pumps. While these valves are open the RHR suction relief valves are exposed to the RCS and are able to relieve pressure transients in the RCS.

l The RHR suction isolation valves must be open to make the RHR suction relief valves OPERABLE for RCS overpressure mitigation.

l The RHR suction relief valves are spring loaded, bellows type water relief valves with pressure tolerances and accumulation

limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3) for Class 2 relief valves.

RCS Vent Recuirements Once the RCS is depressurized, a vent exposed to the containment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent.

Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass or heat input transient.

(continued)

WCGS ITS BASES B 3.4-65 5/15/97

-- . - . .-- - - - - - . = - . . . . . . . . - . . - - - - - -

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LTOP System B 3.4.12 i-BASES l BACKGROUND RCS Vent Reauirements (continued) and maintaining pressure below the P/T limits. The required vent capacity may be provided by one or more vent paths.

i APPLICABLE Safety analyses (Ref. 4) demonstrate that the reactor vessel SAFETY ANALYSES is adequately protected against exceeding the Reference 1 P/T limits. In H0 DES 1, 2, and 3, the pressurizer safety valves will

! prevent RCS pressure from exceeding the Reference 1 limits. In H0DE 3 (with any RCS cold leg temperature s 368'F) and below.

overpressure prevention falls to two OPERABLE RCS relief valves or to a depressurized RCS and a sufficient sized RCS vent. Each j of these means has a limited overpressure relief capability.

! l l The actual temperature at which the pressure in the P/T limit

curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlemnt. Each time the PTLR curves are revised, the LTOP l System must be re evaluated to ensure its functional requirements '

can still be met using the RCS relief valve method or the depressurized and vented RCS condition.

1 The PTLR contains the acceptance limits that define the LTOP requirements. Any change to the RCS must be evaluated against the Reference 9 analyses to determine the impact of the change on the LTOP acceptance limits.

Transients that are capable of overpressurizing the RCS are i

categorized as either mass or heat input transients, examples of l which follow:

Mass Inout Tvoe Transients

a. Inadvertent safety injection: or l

l b. Charging / letdown flow mismatch.

Heat Inout Tvoe Transients

! a. Insdvertent actuation of pressurizer heaters; i-

b. Loss of RHR cooling: or

(continued)

/ WCGS ITS BASES B 3.4 66 5/15/97 i

LTOP System B 3.4.12 BASES APPLICABLE Heat Inout Tvoe Transients (continued)

SAFETY ANALYSES

c. Reactor coolant ptap (RCP) startup with temperature asymmetry within the RCS or between the RCS and steam generators.

The following are required with exception described below during the LTOP MODES to ensure that mass and heat input transients do not occur, which either of the LTOP overpressure protection means cannot handle: f(the. UcP is aho e#4s/4 bit, hy N L LToP modas ) __

a. Rendering both safety injection pumps nd one centrifugal charging pump incapable of injection:

b. Deactivating the accumulator discharge isolation valves in their closed positions or by venting the affected accumulator: and

c. Precluding start of an RCP if secondary temperature is more than 50'F above primary temperature in any one loop. LCO 3.4.5, "RCS Loops MODE 3," LCO 3.4.6, "RCS Loops-MODE 4." and LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled." provide this protection.

Operation below 350*F but greater than 325'F with all centrifugal charging and safety injection pumps OPERABLE is allowed for up to

%"N i"*'. "

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• 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. ,During low pressure, low temperature operation all automatic safety injection actuation signals except Containment h * *+ h Pressure High are blocked. In normal conditions a single s acp., failure of the ESF actuation circuitry will result in the starting of at most one train of safety injection (one centrifugal charging pump, and one safety injection pump). For temperatures above 325'F. an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both PORV's without exceeding Appendix G limit.

Given the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed. Initiation of both trains of safety injection during this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time frame due to operator error or a single failure l occurring during testing of a redundant channel are not censidered to be credible accidents.

t (continued)

WCGS ITS BASES B 3.4 67 5/15/97

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LTOP System.

B 3.4.12 BASES APPLICABLE Heat Inout Tvoe Transients (continued) M'

' SAFETY ANALYSES #

Although LTOP is required to be OPERABLE when rcd temperature is  ;

less than 368'F. operation with all centrifugal c 1arging pumps '

and both safety injection pumos OPERABLE is accep ;able when RCS temperature is greater than 350*F. Should an inadvertent safety injection occur above 350*F. a single PORV has su 'ficient capacity to relieve the combined flow rate of all pumps. Above 350'F, two RCPs and all pressurizer safety valves are required to

-be OPERABLE. Operation of an RCP eliminates the possibility of a 50'F difference existing between indicated and actual RCS -

temperature as a result of heat transport effects. Considering l instrument uncertainties only, an indicated RCS temperature of I 350*F is sufficiently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pressurizer safety valves provide acceptable and redundant overpressure protection.

Gn awhin 6 %_ dcP)

The Reference 9 analyses demonstrate that either one RCS relief valve or the depressurized RCS and RCS vent can maintain RCS pressure below limits when only one centrifugal charging ptmp is actuated. Thus, the LCO allows only one centrifugal charging ptmp OPERABLE during the LTOP MODES. Since neither one RCS relief valve nor the RCS vent can handle the pressure transient caused by accumulator injection, when RCS temperature is low, the LCO also requires accumulator isolation when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed in the PTLR.

The isolated accumulators must have their discharge valves closed and the valve power supply breakers fixed in their open positions. Fracture mechanics analyses established the temperature of LTOP Applicability at 368'F.

PORV Performance The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limit shown in the PTLR. The setpoints are derived by analyses that model the performance of the LTOP System assuming the mass injection transient of one centrifugal charging pump injecting into the RCS and the heat injection transient of starti g an RCP with the RCS F

atms h uce (continued)

WCGS ITS BASES B 3.4 68 5/15/97

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..- . . . - _ - - - - - - . = . - . _ - . . . . - - . . _ .. . -

LTOP System B 3.4.12 BASES APPLICABLE PORV Performance (continued)

SAFETY ANALYSES l

50*F colder than the secondary coolant. These analyses consider pressure overshoot and undershoot beyond the PORV opening and

! closing, resulting from signal processing and valve stroke times.

The PORV setpoints at or below the derived limit ensures the Reference 1 P/T limits will be met. 4 The PORV setpoints in the PTLR will be updated when the revised P/T limits conflict with the LTOP analysis limits. The P/T  ;

limits are periodically modified as the reactor vessel material '

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' toughness decreases due to neutron embrittlement caused by neutron irradiation. Revised limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits,"

discuss these examinations. '

l The PORVs are considered active components. Thus, the failure of  !

l one PORV is assumed to represent the worst case, single active failure.

RHR Suction Relief Valve Performance The RHR suction relief valves do not have variable pressure and temperature lift setpoints like the PORVs. Analyses show that one RHR suction relief valve with a setpoint at or between 436.5 psig and 463.5 psig will pass flow greater than that required for the limiting LTOP transient while maintaining RCS pressure less than the P/T limit curve.

l As the RCS P/T limits are decreased to reflect the loss of toughness in the reactor vessel materials due to neutron embrittlement, the RHR suction relief valves must be analyzed to still accommodate the design basis transients for LTOP.

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The RHR suction relief valves are considered active components.

Thus, the failure of one valve is assumed to represent the worst case single active failure.

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, (continued)

WCGS ITS BASES B 3.4 69 5/15/97

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l. Attachment VII to WO 98-0104 i j Page 1 of 4 I

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l Attachment VII 10 CFR 50.60 Exemption Request i

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1 Attachment VII to WO 98-0134 Page 2 of 4 l

10 CFR 50.60 Exemption Request

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1 Requirement for Which Exemption is Requested l i

Pursuant to 10 CFR 50.12, " Specific Exemptions," enclosed is a request for l

exemption from certain requirements of 10 CFR 50.60, " Acceptance criteria for i fracture prevent measures for light water nuclear power reactors for normal 1 operation," and 10 CFR 50, Appendix G, " Fracture Toughness Requirements," for l

Wolf Creek Generating Station (WCGS). This exemption is requested to allow '

the application of ASME Code Case N-514, " Low Temperature Overpressure Protection," in determining the acceptable low temperature overpressure protection (LTOP) system [ referred to as cold overpressure mitigation system (COMS) at Wolf Creek) enable temperature and power-operated relief valve (PORV) pressure setpoints for WCGS.

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ASME Section XI Code Requirements l ASME Section XI, Appendix G, " Fracture Toughness Criteria for Protection Against Failure," Article A-2000, provides the service limits for pressure vessels, and establishes the allowable vessel loading (internal pressure, external load, thermal stress) versus temperature. The Code requirements are to maintain vessel operating conditions within Article A-2000 requirements.

Code Requirement from Which Exemption is Requested Exemption is requested from 10 CFR 50, Appendix G, and ASME Section XI, Appendix G, requirements for reactor vessel pressure limits at low temperatures.

Basis for Exemption Request Current COMS setpoints produce operational constraints by limiting the '

pressure-temperature range available to the operator to heat up or cool down the plant. The " operating window" through which the operator must heat up and pressurize, or cool down and depressurize the reactor coolant system (RCS) is determined by the difference between the maximum allowable pressere determined by Appendix G of ASME Section XI, and the minimum required pressure for the reactor coolant pump (RCP) seals, adjusted for COMS overshoot and instrument uncertainties. Unoer the pressure-temperature (P/T) requirements of Appendix G of ASME Section XI, COMS can have significant impact on operation by limiting RCP and residual heat removal pump operation at low temperatures.

In addition, the operating pressure window imposed by COMS becomes more and more restrictive with reactor vessel service. Reducing this operating window could patentially have an adverse safety impact by increasing the possibility of inadvertent COMS actuation due to pressure surges associated with normal plant evolutions such as RCP start and shifting operating charging pumps ith the reactor coolant system RCS in a water-solid condition.

Wolf Creek Nuclear Operating Corporation (WCNOC) has evaluated the impact on the P/T limits and COMS setpoints due to increasing the service period to 20 EFPY based on ASME Section XI, Appendix G, requirements, and has concluded the l requirements for COMS would significantly restrict the ability to perform plant heatup and cooldown, create an unnecessary burden to plant operations, and challenge control of plant evolutions required with COMS enabled. The proposed license amendment limits reactor coolant pump operation to no more than two pumps operating with the RCS less than 100*F.

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Attachment'VII to WO 98-0104 Page 3 of 4 Proposed Alternative WCNOC proposes the use.of ASME Code Case N-514 requirements for reactor vessel i pressure limits at low temperatures as an acceptable alternative to 10 CFR 50, Appendix G, and ASME Section XI, Appendix G, requirements.

Justification for Granting of Relief Pursuant to 10 CFR 50.12', -the Commission may grant exemptions from the requirements of 10 CFR 50 when:

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1) the ( temptions are authorized by law, will 1 not present an. undue risk to public health or safety, and are consistent with l the common defense and security, and 2) when special circumstances are l present. . Special. circumstances are present whenever, according to 10 CFR 50.12 (a) (2) (ii), " Application of the regulation in the particular  ;

circumstances would not serve the underlying purpose of the rule or is not  !

necessary to achieve the underlying purpose of the rule." . l WCNOC believes the underlying purpose of 10 CFR 50.60, Appendix G, is to establish fracture toughness requirements for ferritic materials of pressure retaining components of the reactor coolant pressure boundary.. These requirements provide adequate margins of safety durino any condition of normal l operation,- including unanticipated operational occurrences, to which the pressure boundary may be subjected to- over its service lifetime.

Section IV. A.2 of this appendix requires that the reactor vessel be operated with P/T limits at least as conservative as those obtained by following the methods.of analysis and the required margins of safety of Appendix G of ASME Section XI.  !

Appendix G of Section XI of the 1SME Code requires the. P/T limits be calculated: a) using a safety facto:- af 2 on the principal membrane (pressure) stresses, b) margin added to the reactor vessel RTm in accordance with Regulatory Guide 1.99, Rev. 2, " Radiation Embrittlement of Reactor Vessel Materials," c) assuming a flaw at the surface with a depth of 1/4 of the vessel wall thickness and a length of. 6 times its depth, and d) using a conservative fracture toughness curve that is based on the lower bound of static, dynamic, and crack arrest fracture toughness tests on material similar to the reactor vessel material.

I In determining the setpoints for COMS, WCNOC proposes to apply ASME Code Case N-514. This code case provides for normal operation within the P/T j limits determined in accordance with ASME Section XI, Appendix G, but allows determination of setpoints for COMS events such that the maximum pressure in the vessel would not exceed 110% of the Appendix G limits. The safety margins l provided by application of Code Case N-514 result in a safety factor of 1,8 on  !

the principal membrane (pressure) stresses. All other factors, including assumed flaw size and fracture toughness, remain the same as ASME Section XI, Appendix G, methodology. The basis of ASME Code Case N-514 indicated that, due to the isothermal nature of the COMS events, the margin with respect to )

toughness for a COMS transient is within the range provided by ASME Section XI, Appendix G, for normal heatup and cooldown in the low temperature range. Thus, applying Code Case N-514 will satisfy the intent of 10 CFR 50.60 for fracture toughness requirements. Further, application of Code Case N-514 will relieve operational restrictions for WCGS. This will reduce the

. potential for inadvertent-RCS pressure relief events, thereby improving plant l

safety, and will reduce unnecessary burdens on operators during important

-plant evolutions.

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Attachment VII to WO 98-0104 Page 4 ef 4 Implementation Schedule The exemption request will be implemented within 60 days of approval of the associated license amendment request, and prior to the WCGS reaching 13.6 EFPY.

This request is similar to the following:

Turkey Point Units 3 and 4 exemption request approved by the NRC on May 11, 1993 (exemption from the requirements of 10 CFR 50.60);

e McGuire Units 1 and 2 request for exemption - ASME Code Case N-514, dated June 20, 1994;

• Arkansas Nuclear One, Unit 2 request for exemption - ASME Code Case N-514, dated April 11, 1996;

• Point Beach Nuclear Plant, Units 1 and 2 request for exemption - ASME Code Case N-514, dated January 27, 1997; and

• Diablo Canyon Power Plant, Units 1 and 2 request for exemption from 10 CFR 50.60, dated September 3, 1998.

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