Text

Application for Amend to License DPR-72,consisting of Tech Spec Change Request 174 Re Proposed Heatup,Cooldown & Inservice Leak & Hydrostatic Testing Pressure/Temp Limits. Rept Re Low Temp Overpressure Protection Withheld
	ML19325E810
Person / Time
Site:	Crystal River
Issue date:	10/31/1989
From:	Boldt G; FLORIDA POWER CORP.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19292J545	List: 3F1089-23, Application for Amend to License DPR-72,consisting of Tech Spec Change Request 174 Re Proposed Heatup,Cooldown & Inservice Leak & Hydrostatic Testing Pressure/Temp Limits. Rept Re Low Temp Overpressure Protection Withheld; ML19325E811;
References
	RTR-REGGD-01.099, RTR-REGGD-1.099 3F1089-23, GL-88-11, NUDOCS 8911090034
	Download: ML19325E810 (100)
	v • d • e

!

r

=

C 0 h P 0 ft A14 0 N October 31, 1989 3F1089-23 U.S. Nuclear Regulatory Commission Attnt Document Control Desk Washington, D.C.

20555

Subject:

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Technical Specification Change Request No. 174 Pressure /Terparature Limits Generic Istter 88-11 Submittal Dear Sirt Florida Power Corporation (FPC) hereby submits Technical Specification Change Request No. (TSCRN) 174 Revision 0, requesting amendment to Appendix A of Operating License No. DPR-72.

Proposed replacement pages for both the current CR-3 Technical Specifications and the Improved Technical Specifications are provided.

Replacement pages for the associated bases are also provided.

This change request proposes Heatup, Cooldown and Inservice Leak and Hydrostatic Testing pressure / temperature limits based on 15 effective full' power years (EFPY) of reactor operation.

This submittal also proposes a low temperature overpressurization protection (LTOP) features Technical Specification for Crystal River Unit 3 (CR-3).

Previous FPC commitments for administrative controls regarding LTOP are superseded by this submittal. The LTOP approach, discussed with the NRC staff in a May 16, 1989 meeting between FPC, the NRC, and Babcock and Wilcox (B&W), has been developed utilizing non-10CFR50 Appendix G methodology.

The non-Appendix G approach is based on the low probability of occurrence for an LTOP-type' event at CR-3 and is consistent with the NRC position in Generic Letter 88-11.

The attached engineering document that supports the use of the non-Appendix G methodology has been determined to be proprietary to B&W in accordance with 10CFR2.790.

The Affidavit of Mr. James H.

Taylor, B&W's Manager of Licensing Services, which identifies the summary report as B&W proprietary, has been enclosed.

ggt hopoggggggggggjo2 k d' M ~

p PDC 9i Q^

POST OFFICE Box 219

CRYSTAL RIVER, FLORIDA 326290219 * (904) 563 2943

, l $*ff

A Florlds Progress Company

i October 31, 1989 r

l 3F1089-23 j

This submittal completes FPC's commitment on Generic Letter 88-11 "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations".

(FPC letter 3F1188-I 15 to the NRC, dated November 23, 1988; modified in FPC letter 3F0889-10 to the NRC, dated August 14, 1989).

In the Generic Letter 88-11 respons.e, FPC committed to the use of Regulatory Guide i

1.99 Revision 2,

in the development of the 15 EFPY pressure /

temperature limits for CR-3.

Both the Appendix G pressure /

temperature limits and the LTOP pressure / temperature limit curve have been calculated utilizing the methods described in Regulatory Guide 1.99 Revision 2 to predict the effect of neutron irradiation on reactor vessel material properties.

FPC requires NRC review and approval of the proposed changes by January 1991 in order to continue to operate CR-3, as current 8 t

l EFPY pressure / temperature limits are projected to expire at that time.

However, the current CR-3 Technical Specification l

Improvement Program lead plant schedule may necessitate NRC 4

approval of the LTOP Technical Specification by June 1990.

FPC i

i requests this amendment become effective 30 days after issuance in order to allow for procedure changes and training, Sincerely, I

r Gary Boldt, Vice President Nuclear Production f

GLB/BPW j

i Attachment xct Regional Administrator, Region II Senior Resident Inspector l

l 1

o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION t

IN THE MATTER

)

DOCKET No.

50-302 FLORIDA POWER CORPORATION

)

CERTIFICATE OF SERVICE Gary Boldt deposes and says that the following has been served on the Designated State Representative and Chief Executive of Citrus County, Florida, by deposit in the United States mail, addressed as follows:

Chairman, Administrator Board of County Commisrioners Radiological Health Services of Citrus County Department of Health and Citrus County Courthouse Rehabilatative Servicca Inverness, FL 32650 1123 Winewood Blvd.

Tallahassee, FL 32301 A copy of Technical Specification Change Request No. 174, Revision 0, requesting Amendment to Appendix A of Operating License No. DPR-72.

FLORIDA POWER CORPORATION

/

Gary Boldt, Vice President I

Nuclear Production r

SWORN TO AND SUBSCRIBED BEFORE ME THIS 31ST DAY OF OCTOBER 1989.

I A

(

W Wotary Public Notary Public, State of Florida at Large My Commission Expires: /p//9 /fp conu ruaut sun onumen n umt in CDMW%IDfM APmIS OCL 19.1MC f

ov.aaomaan *>um axacy oc i

l

n I

I STATE OF FLORIDA l

COUNTY OF CITRUS I

cary Boldt states that he is the Vice President, Nuclear Production for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and' matters set forth therein are true and correct to the best of his knowledge, information, and belief.

I Gary 8oldt, Vice President Nuclear Production Subscribed and sworn to me, a Notary Public in and for the State and County above named, this 31st day of October, 1989.

b h M d h' d w.A-Notary Public i

Notary Public, State of Florfda at Large My Commission Expirent /o// 9/9 u Not ARY PUBLIC. $1 ATE OF FL0010A AT LAM 0f i

MY COMMi$5 ION (KPIP45 001.19.1M m otoi e awonaetw;vmc

I L

v i

PLORIDR POWER 00RPORATION CRYSTRL RIVER UNIT 3 DOCRET NO. 50-302/ LICENSB NO. DPR-72 REQUEST NO. 174, REVISION 0 l

PRESSURE /TEMPERRTURE LIMITS R.

LICENSE DOCUMENT INVOLVED:

Technical Specifications PORTION:

Index, Page V

/ TSIP Index Not affected Index. Page IX / TSIP Index Not affected 3.4.12

/ TSIP Specification 3.3.8 3.5.2

/ TSIP Specification 3.4.2 3.5.3

/ TSIP Specification 3.4.3 3.4.3.2

/ TSIP Specification 3.3.10 Not Affected 3.1.2.4.1

/ TSIP Specification (N/A)

N/A

/ TSIP Specification 3.

3.6 DESCRIPTION

OF REQUEST The proposed change would add a Limiting Condition for Operability (LCO) for Low Temperature Overpressurization Protection Features to Technical Specifications.

The change also allows two high pressure injection pumps to be deactivated in MODE 3 and MODE 4 in accordance with Limiting Condition for Operation (LCO) 3.4.12 Low Temperature Overpressure Protection Features. The submittal proposes adding a note to the MODE applicability of LCO 3.1.2.4.1 to allow two makeup / HPI pumps to be deactivated in accordance with LCO 3.4.12 Low Temperature Overpressure Protection Features.

A note is also provided to LCO 3.4.3.2 to alert the operator that Pressure Operated Relief Valve (PORV) OPERABILITY for RCS pressure control is also required by LCO 3.4.12 Low Temperature Overpressure Protection Features.

The proposed change also re-defines the MODE APPLICABILITY section of Technical Specification Improvement Program (TSIP) LCO 3.3.6 Pressurizer Water Level to read MODES 1, 2,

and MODE 3 with RCS temperature > 283'F.

RER8oM Pom REQUESTt Low temperature overpressurization protection (LTOP) features ensure the reactor coolant pressure boundary (RCPB) is adequately protected at low reactor coolant system (RCS) temperatures.

The RCPB is one of the primary boundaries to fission product release l

and protective features that insure its' integrity should be included in Technical Specifications.

Additionally, proposed LTOP 1

l

e f

features are based on non-10CFR50 Appendix G methodolgy that i

utilize reduced margins as part of the analysis.

To ensure the plant is operated within the bounds of the analysis, the critical i

analysis assumptions are preserved in Technical Specifications.

With reactor coolant system (RCS) temperature less than 2 8 3*F,

inadvertent high pressure injection (HPI) must be administrative 1y precluded for low temperature overpressure protection (LTOP).

HPI l

actuation could potentially pressurize the RCS in excess of the i

allowable LTOP fracture mechanica limits.

Low temperature overpressure protection requires that two HPI trains be deactivated whenever RCS temperature is less than or equal to 283'F. This is in conflict with LCO 3.1.2.4.1 which requires two OPERABLE makeup pumps when RCS temperature is greater than 280'F.

j i

The LTOP LCO also contains requirements on PORV OPERABILITY for RCS I

pressure control which must be preserved and appropriate actions i

to be taken in the event the requirements are not met.

The PORV is one of the required redundant LTOP features and must be OPERABLE to provide overpressure protection in an LTOP-type event.

At RCS temperatures less than or equal to 2 8 3'F,

LCO 3.4.12 requires additional restrictions on the pressurizer water level l

than provided in LCO 3.3.6.

This is necessary to ensure the low l

temperature overpressure protection features are maintained.

EVALUATION OF REQtlESTt In Generic Letter 88-11, the Nuclear Regulatory Commission

)

indicated it would consider the use of non-10CFR50 Appendix G methodology for developing low temperature overpressurization protection (LTOP) limits. The reduced analysis margins used in the non-Appendix G methodology were intended to make the amount of 9

conservatism in the LTOP limits more representative of the risk.

t Florida Power Corporation (FPC) pursued this opportunity because l

relaxed LTOP limits result in more operational flexibility and

)

l generally lower operator burden.

To justify use of the non-s l

Appendix G method, FPC has 1) demonstrated the frequency of occurrence for a low temperature overpressure event that would exceed Appendix G limits is less than one per reactor lifetime, and l

2) developed adequate safety justification in regards to the appropriate fracture toughness limits.

Crystal River Unit 3 (CR-3) has two design features / operating practices which contribute to the low probability of occurrence for an LTOP-type event.

The first of these is the operating practice t

of maintaining a gas or steam bubble in the pressurizer at all times (except system hydrotest).

This provides a surge volume, l

which unlike the " water solid"

system, can accommodate most

(

pressure transients.

The " water solid" system has the potential for almost instantaneous pressure increase for mass addition 2

~

i

)

transients s.nd significantly faster pressure increases for energy

)

addition transients.

Operating experience throughout the nuclear

]

p power industry has shown that most all LTOP-type events occurred during heatup and cooldown during a " water solid" condition.

Secondly, at low RCS pressures, other pressurized water reactor (PWR) designs are configured so that an inadvertent isolation of the Residual Heat Removsl System will terminate letdown flow from the RCS.

RCS pressure increases as the makeup pump continues to operate, and eventually the RCS is overpressurized.

The CR-3 i

design does not route letdown flow through the Decay Heat Removal l

system, and is not susceptible to this type of event.

i The basis for LCO 3.3.6 Pressurizor Water Level (TSIP format)

{

requiring MODE 3 and MODE 4 with RCS temperature greater than or e

equal to 275'F is to prevent solid water RCS operation during plant heatup and cooldown.

The cushion effect of this steam space l

prevents rapid RCS pressure rises due to normal plant perturbations.

The revision to this LCO will require a more restrictive limit on pressurizer water level over the range of RCS temperatures from 283*F to 276'F and continue to maintain the I

remainder of the existing requirements. Lowering pressurizer water level to that required by LCO 3.4.12 (220 inches), at the RCS temperaturo of 2 8 3'F, results in a larger margin to solid water operation and a larger steam space to cushion the RCS against overpressure transients.

i The LTOP LCO was written to ensure redundant LTOP subsystems are operational when the LTOP features are required for protection of the RCPB.

The primary means of LTOP is operator action to I

terminate the event.

For conservative purposes, the analysis 3

demonstrates, and the plant is operated, such that the operator has at least 10 minutes from the time an alarm is received until RCS pressure exceeds the LTOP fracture toughness (non-Appendix G) limits.

As a backup to operator action, the pressure operated relief valve (PORV), with the reduced pressure setpoint selected, is set to relieve at an RCS pressure less than that corresponding r

to the minimum LTOP fracture toughness limits.

The LCO is written to 1) ensure the PORV is OPERABLE, and 2) place restrictions on plant operations that ensure 10 minutes is available for operator action.

Remedial actions have been provided to ensure that in the event the LTOP LCO is violated, specific steps will to taken to place the plant in a non-applicable condition.

Surveillance Requirements ensure the limitations are maintained.

With the PORV inoperable in MODES 1,2 or 3, LCO 3.4.3.2 action (a) directs the operator to restore the PORV to OPERABLE within one hour or close and remove power to the block valve.

Failing this, the plant shall be placed in HOT STANDBY in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN in the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

This action places the plant into the LTOP LCO which has separate requirements for PORV OPERABILITY and appropriate actions to take if the PORV is 3

~

i i

inoperable.

The note added to the PORV LCO ?rovides a cross-reference between the two specifications.

Provnding the operator with the note on other PORV operability requirements ensures the operator will expeditiously proceed to LCO 3.4.12 Low Temperature i

Overpressurization Protection Features and carry out the appropriate actions for an inoperable PORV in these MODES.

10CFR50 Appendix G

requires the operator to maintain RCS temperature and pressure within the limits of LCO 3.4.9 Pressure /

Temperature Limits at all times.

Tnese limits are based on ASME Section III, Appendix G assumptions for normal occurrences and the i

plant must be operated within these limits.

The non-Appendix G curve is only used to determine the 10 minute " window" and set PORV setpoint.

The curve is contained in the engineering document supporting the non-Appenix G limits and does not appear in procedures or within Technical Specifications. This will eliminate any confusion which might occur as a result of two pressure /

temperature limit curves.

Two analyzed overpressure transients resulted in RCS pressure 1

increases that exceed LTOP limits in less than 10 minutes.

These transients are high pressure injection (HPI) and core flood tank (CFT) actuation. The two events must be administrative 1y precluded from occurring, and this is the bases for requiring both trains of each system to be deactivated in the LTOP region.

With the HPI and CFT actuation precluded, the limiting analysis case becomes the full-open failure of the makeup contro2. valve.

This event is the J

basis for PORV setpoint, pressurizer level, and allowable makeup l

tank level.

]

i Technical Specification LCO 3.1.2.4.1 requires at least two 1

OPERABLE makeup ) pumps in MODE 3 (RCS temperature greater than or equal to 280 F.

LCO 3.4.12 Low Temperature Overpressure Protection Features requires two trains of HPI be deactivated whenever RCS temperature is less than or equal to 283'F to ensure 1

an inadvertent Engineered Safeguards (ES) actuation does not result l

in overpressurizing the RCS.

This creates a conflict between the two LCOs since the HPI pumps also serve as the normal makeup pumps, whereby deactivating the two trains of HPI leaves only one OPERABLE t

makeup pump.

Deactivating the two trains of HPI at 283'F, also causes the operator to enter the action statement for LCO i

3.1.2.4.1.

This LCO provides 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the operator to restore at least two makeup pumps to OPERABLE or take action to place the plant in a safe MODE of operation where the LCO does not apply.

1 The allowable out-of-service period ensures that minor component l

repair or corrective action may be completed without undue risk to overall facility safety from injection system failure during the repair period.

The amount of time the plant remains between the temperatures of 283'F and 200*F is expected to be much less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and once the plant is in MODE 4 (RCS temperature less than 2 8 0*F) only one makeup pump is required, and LCO 3.1.2.4.1 is no 4

l l

Current Technical Specifications allow the HPI injection isolation valve power supply breakers to be " racked out" in MODE 4.

Based on the updated LTOP ana3ysis, HPI should now be deactivated whenever RCS temperature is equal to or less than 283'F (MODE 3) and the reactor vessel head is not fully detensioned. " Racking out" the power supplieo for the HPI isolation injection valves still allows manual initiation of HPI by " racking in" the breakers and then operating the talves (The footnote for current CR-3 LCO 3.5.3 has been reworded to clarify the meaning of " racking out" as it e.pplies to the actual plant practice for deactivating HPI isolation injection valves). However, HPI would not be immediately available since operator action outside the control room is required.

FPC has evaluated alternative methods for deactivating HPI that do not require " racking out" power supply breakers.

These methods would provide the operator with the capability of restoring HPI without leaving the control room and could be used at the higher LTOP temperatures to make HPI more readily available in the event it is needed.

To further balance the possible need for core cooling, LCO 3.4.12 does not require the makeup system to be deactivated.

At the lower temperatures associated with LTOP, and the expected decay heat levels, the makeup system can provide adequate flow via the makeup control valve, or it can be used in the interim until HPI can be re-activated.

Raising the RCS temperature at which HPI is deactivated from 280'F to 283 P is considered acceptable.

This conclusion is based on 1) the low probability of a loss of coolant accident requiring immediate HPI flow in this region of plant operation,

2) the limited amount of time the plant is actually operated between the RCS temperatures of 280 and 283F,

3) the ready accessibility of RCS makeup flow and HPI, and 4) the improved LTOP protection.

l The LTOP fracture toughness limits are based on 21 Effective Full Power Years (EFPY) of reactor operation for determining neutron fluence values for the reactor vessel.

Similar to the Appendix G heatup and cooldown limits, the vessel is the limiting RCS l

component in terms of LTOP fracture toughness.

Regulatory Guide 1.99 Revision 2 was utilized to determine the shift in reference nil ductility transition temperature due to neutron irradiation.

This method was endorsed in Generic Letter 88-11 as an acceptable l

method for predicting the shif t in material behavior.

The analysis used the heatup and cooldown rates assumed in generating the 15 EFPY Appendix G pressure / temperature limit curves (listed in Pressure Temperture Limits LCO 3.4.9).

l 1

l 5

l

)

SNOLLY RYALUATION OF REQUESTt i

Florida Power Corporation (FPC) proposes the addition of a Low

)

Temperature overpressurization Protection (LTOP) Features Technical Specification does not involve a significant hazards consideration.

The addition of a Technical Specification requirement to maintain the LTOP features ensures the reactor coolant pressure boundary is protected against a non-ductile failure at low reactor coolant temperatures.

Based on the above, FPC concludes this change will not 1.

Involve a

significant increase in the probability or consequence of an accident previously evaluated since there are currently no LTOP requirements in Technical Specifications.

This change represents additional requirements necessary to preclude an LTOP event from occurring.

These additional requirements also provide protection for all pressure and temperature combinations for which an LTOP event may be postulated.

Overall, these requirements provide a level of protection greater than or equivalent to existing requirements.

2.

Create the possibility of a new or different kind of accident from any previously evaluated because the addition of an LTOP l

Technical Specification does not require modification to the plant nor does it create a new mode of plant operation, with the exception of the small increase in reactor coolant temperature at which the high pressure injection system is deactivated. This is considered acceptable based on the small amount of time the plant is operated in this temperature

~

region and the low probability of a loss of coolant accident (14CA) requiring immediate high pressure injection flow at the reactor coolant temperatures in this region. In the unlikely event a LOCA does occur, high pressure injection would be l

available following operator restoration of the system.

Reactor coolant makeup flow would be available in the interim to provide core cooling requirements.

t 3.

Irivolve a significant reduction in the margin of safety. Any reduction in the margin of safety will be insignificant and offset by the safety benefit gained through the additional requirements placed on plant operation to preclude a low temperature overpressurization event.

l i

7.._

l I

i i

TSCRN 174A CURRENT TECHNICAL SPECIFICATION FORMAT i

t

)

l I

1 i

l 1

l l

t 1

i e

I t

l

c I

INDEX o

LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS l

SECTION Pf.A[

3/4.4.4 PRESSURIZER,................

3/4 4-5 3/4.4.5 STEAM GENERATORS...............

3/4 4 6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGL Leakage Detection Systems..........

3/4 4 13 Operational Leakage.............

3/4 4-15 3/4.4.7 CHEMISTRY..................

3/4 4 17 3/4.4.8 SPECIFIC ACTIVITY..............

3/4 4 20 j

3/4.4.9 PRESSURE / TEMPERATURE LIMITS l

Reactor Coolant System............

3/4 4 24 Pressurizer.................

3/4 4 30 l

3/4 4.10 STRUCTURAL INTEGRITY............

3/4 4-31 3/4.4.11 REACTOR COOLANT SYSTEM VENTS,........

3/4 4-33 3/4.4.12 LOW TEMPERATURE OVERPRESSURIZATION PROTECTION FEATURES...................

3/4 4-35 j

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) i 3/4.5.1 CORE FLOODING TANKS.............

3/4 5-1 l

t 3/4.5.2 ECCS SUBSYSTEMS - Tavg 2 280'F........

3/4 5-3 4

3/4.5.3 ECCS SUBSYSTEMS - Tavg < 280'F........

3/4 5 6 3/4.5.4 BORATED WATER STORAGE TANK..........

3/4 5 7 I

i e

CRYSTAL RIVER - UNIT 3 V

Amendment No.

g i

l l

180f1 BASES 1L01108 EAE 3/4.0 APPLICABILITY B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS l

3/4.1.1 BORATION CONTROL B 3/4 1-1 j

3/4 1.2 BORAT10N SYSTEMS B 3/4 1-2 f

3/4 1.3 MOVABLE CONTROL ASSEMBLIES..........

B3/413 3/4.2 POWER DISTRIBUTION LIMITS............

B 3/8 21 l

jf4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION..

B 3/4 3 1 3/4 3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION..............

B 3/4 3 1 3/4.3.3 MONITORING INSTRUMENTATION.........

B 3/4 3 2 3/4.4.

REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3 3/4 4-1 1

3/4.4.2 AND 3/4.4.3 SAFETY VALVES............

B 3/4 4-1 3/4.4.4 PRESSURIZER..................

B 3/4 4 2 3/4.4.5 STEAM GENERATORS B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE........

B 3/4 4-4 3/4.4.7 CHEMISTRY...................

B 3/4 4 5 1

3/4.4.8 SPECIFIC ACTIVITY...............

B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..........

B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY.............

B 3/4 4 13 j

i 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.........

B 3/4 4-14 3/4.4.12 LOW TEMPERATURE OVERPRESSURIZATION PROTECTION FEATURES B 3/4 4-14 CRYSTAL RIVER - UNIT 3 IX Amendment No.

1

e

)

g1

[

L t't REACTIVITY CONTR01. SYSTEMS t

MAKEUP PUMPS - OPERATING

?

t LIMITING CONDITION'FOR OPERATION

.3.1.2.4.1 LAt least two makeup pumps shall be OPERABLE.

o-

' APPLICABILITY: MODES 1 2, and 3*

i 33 8fl39N' 1

Wiih.citly one makeup pump OPERABLE, restore at least two makeup pumps to 9'

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% Ak/k at 200*F within the next 6-('

hours; restore at least two makeup pumps to OPERABLE status within the next 7 days or be in COLD SHU1DOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

h-i SURVEILLANCE RE0VIREMENTS

'4.1.2.4.1 No additional Surveillance Requirements other than those required by. Specification 4.0.5.

h I

s g

Two makeup /HPI pumps may be deactivated in accordance with Specification o

3.4.12.

CRYSTAL RIVER - UNIT 3 3/4 1-10 Amendment No, i

l m

C..., -

i 7

F REACTOR COOLANT SYSTEM' F

POWER OPERATED RELIEF VALVES (JMITING CONDITION FOR OPERATION 4.6

- 3.4.3.2 The power operated relief valve (PORV) and its associated block valve shall be OPERABLE.*

l APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

With the PORV inoperable, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove )ower from the block valve; otherwise, be in at least HOT STANDBY wit 11n the.next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

t.

With the block valve inoperable, within I hour either restore the block valve to OPL'ABLE status or close the block valve and remove power from the block valve or close the PORV and remove power from the associated solenoid valve; otherwise; be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30

hours, c.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS t

4.4.3.2.1 In addition to the requirements of Specification 3.0.5, the PORV shall be demonstrated OPERABLE at least one per 18 months by. performance of a CHANNEL CAllBRATION.

4.< 3.2.2 The block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.

PORV 0PERABILITY for pressure control is required by Specification 3.4.12.

CRYSTAL RIVER - UNIT 3 3/4 4-4a Amendments Nos.

4

REACTOR COOLANT SYSTEM l

L LOW TEMPERATURE.0VERPRES$URIZATION PROTECTION FEATURES a

LIMITING CONDITION FOR OPERATION L

3.4.12 Low Temperature Overpressurization Protection (LTOP) features shall be L

OPERABLE and shall be comprised of:

a.

Pressurizer level less than or equal to 220 inches, b.

OPERABLE power operated relief valve (PORV) with a setpoint of less than or equal to 555 psig, i

c.

Two trains of High Pressure Injection (HPI) deactivated, and d.

Two core flood tanks (CFT) isolated with the isolation valve closed I

and the power supply breakers fixed in the open position whenever CFT pressure is greater than or equal to the maximum allowable reactor coolant (RC) pressure for the existing RC temperature (per PT limits shown in Figures 3.4-2 and 3.4-3).

APPLICABILITY:

MODE 3 with RCS temperature 1283*F, MODE 4, MODE 5, and MODE 6 with the reactor vessel head not completely detensioned.

ACTION:

]

a.

With pressurizer level greater than 220 inches, restore pressurizer level to less than or equal to 220 inches within one hour, or close and maintain closed the makeup control valve and its associated isolation valve and stop plant heatup within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

]

b.

With the PORV inoperable, restore the PORV to OPERABLE status within one hour or reduce makeup tank level to less t.han or equal to 70 inches and deactivate Low-Low makeup tank level interlock to the BWST suction valves within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

With one or two HPI trains active, deactivate the HPI trains within one hour or close and remove power from the HPI injection valves (MUV-23,MUV-24,MUV-25, and MUV-26) within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , d.

With one or two CFTs not isolated when CFT pressure is greater than the maximum allowable RCS pressure for the ex'. sting RCS temperature, isolate the affected CFT(s) within one hour or within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , increase RCS temperature above 175'F or depressurize CFT to less than 555 psig.

e.

With the LTOP features inoperable for reasons other than above, restore the LTOP features within one hour or depressurize the RCS to atmospheric pressure and establish an RCS vent equivalent to an orifice area of 0.75 square inches and verify both trains of HPI are deactivated and verify both CFTs are isolated with the isolation valve closed and the power supply breaker fixed in the open position within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

f.

The provisions of Specification 3.0.3 are not applicable.

CRYSTAL RIVER-UNIT 3 3/4 4-35 AMENDMENT NO.

c :r 3

s 1

SURVEILLANCE REQUIREMENTS 4.4.12 The LTOP features shall be demonstrated OPERABLE:

a.

By verifying pressurizer level is less than or equal to 220 inches at r

least once per 30 minutes during heatup and cooldown, otherwise verify pressurizer level is less than or equal to 220 inches at least once e

per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

),

b.

By verifying the PORV block valve is open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , c.

By verifying two HPI trains deactivated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d.

By verifying two CFTs are isolated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , e.

By performing a CHANNEL FUNCTIONAL TEST of the PORV at least once per 31-days.

f.

By performing a CHANNEL CALIBRATION of the PORV at least once per 18 months.

g.

By verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that an RCS vent is open when required by ACTION "e" above.

I l

l:

1 L

l^

i r

)

CRYSTAL RIVER-UNIT 3 3/4 4-36 AMENDMENT NO.

r --

]

l j

EMERGENCY CORE COOLING SYSTEMS 1

l ECCS SUBSYSTEMS - Tavg > 280*F LIMITING CONDITION FOR OPERATION I

3.5.2 Two independent ECCS subsystems shall be OPERABLE with each subsystem comprised of:

a.

One OPERABLE high pressure injection (HPI) pump,*

b.

One OPERABLE low pressure injection (LPI) pump, f;

c.

One OPERABLE decay heat cooler, and d.

An OPERABLE flow path capable of taking suction from the borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.

With.one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In-the svent the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and a

submitted to the Commission pursuant to the Specification 6.9.2 i

within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles ~to date.

i Two high pressure injection pumps may be deactivated in accordance with Specification 3.4.12 N

p-L 1

1 t

CRYSTAL RIVER - UNIT 3 3/4 5-3 Amendment No.

- ~.

C h,

j 1

I g

.g

'\\

v, ECCS SUBSYSTEMS - Tavg > 280'F t

LIMITING CONDITION FOR OPERATION j

I 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall c

be OPERABLE:

i a.

One OPERABLE high pressure injection (HPI) pump,*

l I

b.

One OPERABLE low pressure injection (LPI) pump,

'c.

One OPERABLE decay heat cooler, and L.

d.

An' 0PERABLE flow path ** capable of taking suction from the 7

borated water storage tank (BWST) on a safety injection signal and manually transferring suction to the containment sump during e

the recirculation phase of operation.

APPLICABILITY: MODE 4.

7r ACTION:

i g

a.

With one ECCS subsystem OPERABLE because of the inoperability of either the HPI pump or the flow path from the borated water stor-age tank, restore at least one ECCS subsystem to the OPERABLE status within one hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the decay heat cooler or LPI pump, restored at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavg less than 280'F by use of alternate heat removal 4-methods

(

c.

In the event the ECCS is actuated and injects water into the reactor coolant system, a Special Report shall be prepared and in 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE RE0VIREMENTS 4.5.3 The ECCS subsystems shall be demonstrated OPERABLE per the applic-able Surveillance Requirements of 4.5.2 Two high pressure injection pumps may be deactivated in accordance with Specification 3.4.12.

The high pressure injection isolation valves may be closed with their power supply breakers fixed in the open position in MODE 4.

l CRYSTAL RIVER - UNIT 3 3/4 5-6 Amendment No.

t' J.

t

~

r

-v

j j

REACTOR COOLANT SYSTEM (continued)

BASES 3/4.4.11 Reactor Coolant System Vents The operability and surveillance requirements for the Reactor Coolant System (RCS) Vents ensure that gases which could inhibit core cooling during natural circulation may be vented from the RCS.

This system was installed as a result of NUREG-0737, Item II.B.1.

3/4.4.12 Low Temperature Overoressurization Protection Features I

Low temperature overpressurization protection (LTOP) features ensure adequate overpressure protection for the reactor vessel at low reactor coolant system (RCS) temperatures and pressures.

This is especially important since low carbon steels typical of reactor vessel materials, have reduced fracture toughness and are more susceptible to non-ductile failures at lower RCS temperatures.

Traditionally, 10CFR50 Appendix G heatup and cooldown pressure /

l temperature limits have served as the LTOP pressure / temperature limits as well.

This meant LTOP administrative restraints were based on normal heatup and cooldown pressure / temperature limits.

10CFR50 requires fracture toughness limits be established that provide protection of the reactor coolant pressure boundary during any condition of normal operation, including events anticipated to occur one or more times during the reactor lifetime.

This includes normal heatup and cooldown of the RCS and as noted above, also included LTOP.

A review of CR-3 plant design, plant operation, and historical data revealed the frequency of occurrence for a low temperature overpressure event that would exceed 10CFR50 Appendix G limits was considerably less than one per reactor lifetime.

Based on the frequency of occurrence, the use of non-Appendix G l

limits (utilizing reduced design margins) was justified in order to establish appropriate plant limitations for LTOP.

The RCS temperature where low temperature overpressurization is required is determined based on the minimum pressurization temperature.

Above this temperature, LTOP protection is not required since an overpressurization transient will be terminated i

by the pressurizer safety valves (PSV) before the LTOP pressure /

temperature limits are exceeded.

Below the minimum pressurization temperature, LTOP pressure / temperature limits would be exceeded before the PSV relieve.

Reactor coolant pressure boundary protection is based on LTOP features in this region.

Additionally, L

LTOP protection le not required when the reactor vessel head is l

completely detensioned, as overpressurization of an "open" RCS is not considered credible.

CRYSTAL RIVER UNIT 3 B3/4 4-14 Amendment No.

l l

REACTOR COOLANT SYSTEM (continued)

BASES 1

Operator action is assumed as the primary means for terminating RCS pressure increase due to a low temperature overpressurization event.

To ensure a conservative amount of time for the operator to take action, administrative limits on plant operation are implemented which provide 10 minutes prior to exceeding LTOP pressure / temperature limits.

As a backup to operator action, the power operated relief valve (PORV) with reduced pressure setpoint, actuates to relieve RCS pressure increases before exceeding LTOP limits.

CR-3 is designed, wpt for system hydrotest, to be operated with e

a gas or steam bubble in the pressurizer at all times.

The gas or steam space acts as a surge volume and limits the rate of RCS pressure increase in the event of an overpressurization event.

Pressurizer level is a direct indication of the amount of steam space available and must be limited to ensure a 10 minute operator action time is preserved.

l The maximum allowed PORV setpoint for LTOP is derived from the LTOP pressure / temperature limits.

Operation with a setpoint less than t

or equal to the minimum LTOP pressure / temperature limit setpoint ensures that the non-Appendix G criteria will not be violated.

l System pressure overshoot, that is, increase in RCS pressure after pressure reaches the PORV setpoint, does not occur due to the rapid action of the PORV and the relatively slow rates of pressure i

l increase due to the pressurizer steam bubble.

1 Two-of the analyzed overpressurization transients must be l

administratively precluded, since the resultant RCS pressure increase exceeds the LTOP limits in less than 10 minutes.

The two events are inadvertent actuation of high pressure injection (HPI) and core flood tank (CFT) actuation.

With these transients l

precluded as credible events, a full open failure of the makeup l

control valve becomes the limiting overpressurization event on which the LTOP features are based.

The LCO permits HPI surveillance testing for components since pump or valve testing can proceed by alternating the system deactivation from pumps to Valves.

Testing must not permit HPI injection flow to enter the RCS.

CRYSTAL RIVER-UNIT 3 B 3/4 4-15 Amendment No.

J REACTOR COOLANT SYSTEM (continued)

-bases j

l The LTOP pressure / temperature limits are based on 21 Effective Full Power Years (EFPY) of reactor operation and will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens, as required by 10CFR50 Appendix H.

The RCS heatup and cooldown rates used to develop the limits are the same as listed in LCO 3.4.9 " Pressure Temperature Limits".

I l

l CRYSTAL RIVER UNIT 3 B3/4 4-16 Amendment No.

1

c'&._

m m.

A s.-

e k

1.,

l 1.

i.

1 TSCRN 174A TSIP FORMAT I

i 7

l t

t I

I

e4<

^

Pressurizer Water level m"

3.3.6 i

3.3 REACTOR COOLANT SYSTEM (RCS) n 3.3.6 Pressurizer Water Level LC0 3.3.6 :

Pressurizer water level shall be 5 290 inches.

APFLICABILITY:

MODES 1, 2 and MODE 3 with RCS temperature > 283'F ACTIONS

+

CONDITION REQUIRED ACTION COMPLETION TIME l

l A.

Pressurizer water level A.1 Restore pressurizer 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

> 290 inches, water level to within limit.

B.

Required Action RQI B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

MQ B.2 Be in MODE 3 with RCS 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months temperature s 283'F.

p 1

I SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY SR 3.3.6.1 Verify pressurizer water level 5 290 inches.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ll h

l -

- i p

l' 1

E Crystal River Unit 3 3.3-8 Amendment io.n -

h

,'b l

b. l""

ir U<

LTOP Features

(

3.3.8 r

13.3' REACTOR COOLANT SYSTEM (RCS) i-.

3.3.8 Low Temoerature Overoressurization Protection ~ Features t

(

I, LCO 3.3.8 Low temperature overpressurization protection (LTOP) features shall be OPERABLE and shall be comprised of:

a.

Pressurizer level s 220 inches, and r

i b.

OPERABLE power operated relief valve (PORV) with a setpoint of 1 555 psig, and-c.

Two trains of high pressure injection (HPI) deactivated, and d.

Two core flood tanks (CFT) isolated with 'the isolation

?

valve closed and the power supply breakers fixed in the l

open position.

.......................------N0TE-----------------------------

Provisions of LCO 3.0.3 are not applicable.

3 I

APPLICABILITY:

RCS temperature s 283*F and'the reactor vessel head not completely detensioned.

............................NC CFT isolation only required when CFT pressure is 1 to the maximum allowable reactor coolant (RC) pressure for the existing l

RC temperature (in accordance with PT Limit Curves provided in the PRESSURE / TEMPERATURE LIMITS REPORT).

1 l'

ACTIONS CONDITION REQUIRED ACTION COMPLF" ION LIME A.

LTOP features inoperable A.1 Restore pressurizer 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months due to pressurizer level level to s 220 inches.

> 220 inches.

l (continued) f i

l' l-ll' l

Crystal River Unit 3 3.3-10 Amendment

u LTOP Features 3.3.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Required Action A.1 HQI B.1 Close and maintain 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

-met within required closed the makeup Completion' Time, control valve and its associated isolation valve.

AND B.2 Stop plant heatup.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

-C.

LTOP features inoperable C.1 Restore PORV to I hour due to PORV inopera-OPERABLE status.

I bility.

D.

Required Action C.1 D.1 Reduce makeup tank 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months HQI met within required level to 1 70 inches, Completion Time.

AMQ D.2 Deactivate Low-Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months makeup tank level interlock to the BWST suction valves.

E.

LTOP features inoperable E.1 ' Deactivate HPI trains.

I hour due to one or two HPI trains active.

F.

Required Action E.1 HQI F.1 Close and remove power 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months met within required from HPI injection Completion Time, valves (MUV-23, MUV-24, MVV-25, and MUV-26.)

(continued)

Crystal River Unit 3 3.3-11 Amendment

r -

LTOP Features 3.3.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G.- LTOP features inoperable G.1 Isolate affected CFTs.

I hour due to 1 or 2 CFTs not isolated when CFT pressure is ~> the maximum allowable RCS pressure for existing temperature.

H.

Required Action G.1 H.1 Increase RCS tempera-12 hours tiQI met within required ture above 175'F.

Completion Time.

l l

H.2 Depressurize CFTs to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l-

< 555 psig.

I.

LTOP features inoperable I.1 Restore LTOP features.

I hour for reasons other than above.

M 1

1.2.1 Depressurize RCS to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months atmospheric pressure and establish an RCS vent equivalent to an orifice area of 0.75 square inches.

8!iD i

l 1.2.2 Verify two trains of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months HPI deactivated.

AND I.2.3 Verify two CFTs are 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months isolated with the isolation valve closed and the power supply breaker fixed in the open q

position.

Crystal River Unit 3 3.3-12 Amendment 1

4-I

b' LTOP Features 3.3.8 SURVEILLANCE REQUIREMENTS L'

SURVEILLANCE FREQUENCY r

SR 3.3.8.1 Verify pressurizer level s 220 inches 30 minutes during RCS heatup and cooldown A!fD 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.8.2' Verify PORV block valve is open.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.8.3-Verify HPI deactivated.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.8.4 Verify CFTs isolated.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.8.5

NOTE------------------------

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Required only when complying with Required Action I.2.

Verify RCS vent open.

SR 3.3.8.6 Perform CHANNEL FUNCTIONAL TEST of PORV.

31 days l

SR 3.3.8.7 Perform CHANNEL CALIBRATION of PORV.

18 months l

\\

i i

4 l

?

1 Crystal River Unit 3 3.3-13 Amendment

s.

P'ressurizer Safety Valves 3.3.9 l

3.3 REACTOR COGLANT SYSTEM _(RCS) 3.3.9 Pressurizer Safety Valves LCO' 3.3.9 Two pressurizer code safety valves shall' be OPERABLE with lif t settings 1 2475 psig and 1 2525 psig.

c:

APPLICABILITY:-

MODES 1, 2, and 3.

c ACTIONS CONDITION REQUIRED ACTION COMPLETION ~ TIME A.

One pressurizer code ~

A.1 Restore valve to 15 minutes safety valve inoper-OPERABLE status, able.

l f-B.-

Required Action 1101 B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

?

met within required Completion Time.

8!iQ B.2 Be in MODE 4 12~ hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Verify pressurizer code safety valves In accordance OPERABLE in accordance with SR 3.0.5.

with SR 3.0.5.

l Crystal River Unit 3 3.3-14 Amendment l

Relief Valves 3.3.10 y.

g,..

3.3. REACTOR COOLANT SYSTEM (RCS) 3.3.10 Relief Valves LC0 3.3.10l One power operated relief valve (PORV) and its block valve shall be OPERABLE.

t APPLICABILITY:

MODES 1, 2, and 3.

[

ACTIONS j

CONDITION REQUIRED ACTION COMPLETION TIME L.

A.

PORV inoperable.

A.1 Restore PORV to I hour OPEPABLE status.

98 A. P.

Close the block valve.

I hour B.

Block valve inoperable.

B.1 Restore affected I hour 7

components to OPERABLE l:

QB status.

98 PORV and block valve inoperable.

B.2.1 Close block valve.

I hour MD B.2.2 Remove power from I hour block valve.

i B.3.1 Close PORV.

I hour l

MD F

B.3.2 Remove power from I hour PORV.

i (continued)

Crystal River Unit 3 3.3-15 Amendment i

r Relief Valves hk 3.3.10 i

f ACTIONS (continued).

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action HQI C.1-Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months mat within required Completion Time.

AND C.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.10.1

NOTE-------------------

Surveillance NQI required with block valve closed in accordance with the Required Actions of LCO 3.3.10.

Operate block valve through one complete 92 days cycle of travel.

SR 3.3.10.2 Perform CHANNEL CAllBRATION for PORV.

18 months

.i Crystal River Unit 3 3.3-16 Amendment h

i l-RCS Leakage 3.3.11

)

.l.

3.3 REACTOR COOLANT SYSTEM (RCS) 3.3.11 RCS Leakaae o

'LCO 3.3.11 Reactor coolant system leakage shall be limited to.

a.

No PRESSURE BOUNDARY LEAKAGE; a

b.

I gpm UNIDENTIFIED LEAKAGE; c.

10 gpm IDENTIFIED LEAKAGE from the RCS.

............................--N0TE-----------------------------

Primary to secondary leakage addressed in LCO 3.3.12 is included in the 10 gpm IDENTIFIED LEAKAGE.from the RCS.

t APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1

A.

RCS leakage outside A.1 Reduce leakage to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits for reasons

limits, other than PRESSURE BOUNDARY' LEAKAGE.

B.

Required Action A.1 HQI B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

AND B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months t

C.

PRESSURE B0UNDARY C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months LEAKAGE exists.

AND C.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months s

Crystal River Unit 3 3.3-17 Amendment

~

n.

e e-an.,

a.w

i.'

L RCS Leakage i

3.3.11 l

p SURVEILLANCE REQUIREMENTS

'l

[

l SURVEILLANCE FREQUENCY j

l-SR 3.3.11,1

NOTE----

Provisions of SR 3.0.4 are not applicable.

L Perform reactor coolant system water 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

[

inventory balance, during steady state operation.

t.

t-P h

(

t Crystal River Unit 3 3.3-18 Amendment s-

p

.)

Prirary to Secondary Leakage-i u.' '

3.3.12 4

n 3.3 REACTOR COOLANT SYSTEM (RCS)

I L

3.3.12 RCS Primary to Secondary Leakaae -

ne

'LC013.3.12 RCS primary to secondary leakage shall be s 1 gpm total through both steam generators.

1 APPLICABILITY:

MODES 1, 2, 3, and 4.

i i

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1

.t l

A.

RCS primary to'second-A.1 Reduce. leakage rate to

'4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ary leakage > 1 gpm.'

within limit.

i -

B.

Required Action HQI B.1 Be in' MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

aHD B.2 Be in MODE S.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

?

SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY r

SR 3.3.12.1

NOTE--------------------

Provisions of SR 3.0.4 are not applicable.

L; Verify primary to secondary leakage 51 gpm.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady n'

L, state opera-L tion.

l l~

l Crystal River Unit 3 3.3-19 Amendment i

't f

w.

.e w

g--

w

%+'

n RCS PIV~ Leakage 3.3.13

. 3.3. REACTOR. COOLANT SYSTEM (RCS).

D' 3.3.13 RCS Pressure Isolatica Valve (PlV) Leakaae f1

.LCO.3.3.13 Leakage for each RCS PIV listed below shall be 15.0 gpm.

1.

CFV-1 2.

CFV-3 3.

DHV-1 L.

4.

DHV-2 f

~,

APPLICABILITY:

MODES I, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.:

Leakage for one or A.1 Reduce leakage rate 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months more RCS PIV > 5.0-within' limit.

gpm.

- c.

B..

Required Action @l B.I Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

-t met within required Completion Time.

AE B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months l

y.

1:,

l^

k

.o I,

l l

Crystal River Unit 3 3.3-20 Amendment

?,y

~

o d'

RCS PlV Leakage 3.3.13 SURVEILLANCE' REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.13.1

NOTE----------------

Provisions of SR 3.0.4 are not applicable for entry into MODES 3 and 4 for the purpose of testing the isolation check g

valves.

Verify leakage for each PIV 15.0 gpm.

Prior to enter-ing MODE 2 when-ever the plant has been in MODE 5 for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more, and if leakage testing has not been performed in the previous 9 months.

MQ In accordance with SR 3.0.5.

l i

i l

i 1

Crystal River Unit 3 3.3-21 Amendment 1

f (g

Leakage Detection Instrumentation 3.3.14 j

3.3 ' REACTOR COOLANT SYSTEM (RCS)

L' 3.3.14' RCS Leakaae Detection Instrumentation a

LCO 3.3.14 The following RCS leakage detection instruments shall be OPERABLE:

l a.

Containment atmosphere activity monitor, and b.

Containment sump level monitor.

Lj' APPLICABILITY:

MODES 1, 2, 3, and 4.

1 i

[

ACTIONS l

CONDITION REQUIRED ACTION COMPLETION TIME L

A.

Required containment A.1 Take and analyze grab Once per 24 atmosphere activity samples.

hours.

j' monitor inoperable.

~

b6 A.2 Restore monitor to 30 days OPERABLE status.

B.

Containment sump level B.1 Perform SR 3.3.11.1.

Once per 24 monitor inoperable.

hours.

MQ B.2 Restore monitor to 30 days OPERABLE status.

C.

Required Action RQI C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

MQ C.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

. Crystal River Unit 3 3.3-22 Amendment

'r.

m Leekage Detection Instrumentation 3.3.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.14.1 Monitor containment atmosphere activity.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.14.2 Monitor containment sump level.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.14.3 Perform a CHANNEL CHECK of containment 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months atmosphere activity monitor.

SR 3.3.14.4 Perform a CHANNEL FUNCTIONAL TEST of con-31 days tainment atmosphere activity monitor.

SR 3.3.14.5 Perform a CHANNEL CALIBRATION of contain-18 months ment atmosphere activity monitor.

SR 3.3.14.6 Perform a CHANNEL CALIBRATION of containment 18 months sump level monitor.

l-Crystal River Unit 3 3.3-23 Amendment

)

7 n-Specific Activity 3.3.15 t

m 3.3 REACTOR COOLANT SYSTEM (RCS) s 3.3.15 Specific Activity b

l-LCO' 3.3.15 The specific activity of the primary coolant shall be s 1.0 microcurie / gram DOSE EQUIVALENT I-131, and s 100/E microcuries/

gram, p

i

^

APPLICABILITY:

MODES I and 2.

MODE 3 with T,y 1 500'F.

................................ NOTE----------------------------

[

Provisions of LCO 3.0.4 are not applicable.

p ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

S)ecific activity of A.1-Sample and perform isoto-Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months tie primary coolant pic analysis for iodine

> 1.0 microcurie / gram including I-131, I-133, DOSE EQUIVALENT I-131 and I-135.

but within the accept-able operation region AND of Figure 3.3.15-1.

A.2 Restore specific activity 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to within limit.

B.

Required Action A.2 B.1 Sample and perform isoto-Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 80.1 met within the pic analysis for iodine, required Completion including I-131, I-133, Time.

and I-135.

DB AND Specific activity in B.2 Be in MODE 3 with Tavg 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months unacceptable operation

< 500*F.

i region of Figure 3.3.15-1.

L, (continued) l Crystal River Unit 3 3.3-24 Amendment

]

Specific Activity 3.3.15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Specific activity C.1 Sample and perform iso-

_Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of the primary coolant topic analysis for

> 100/E.microcuries/

iodine, including 1-131, gram.

1-133, and 1-135.

AND i,

C.2 Be in MODE 3 with T 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

< 500'F.

avg D.

S>ecific activity of 0.1 Sample and perform Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months tie primary coolant isotopic analysis for

> 1.0 microcurie / gram iodine, including 1-131, DOSE EQUIVALENT I-131 1-133 and 1-135.

and > 100/E micro-curies / gram.

AMQ D.2 Be in MODE 3 with Tavg 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

< 500'F.

I 1

Crystal River Unit 3 3.3-25 Amendment

l-

]1 L'

p.

Specific Activity

')

3.3.15 j

g SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY-SR 3.3.15.1 Verify specific activity of the primary 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months coolant s 1.0 microcuries/ gram DOSE EQUIVALENT'I-131.

1

'SR 3.3.15.2

NOTE---------------------

m Provisions of SR 3.0.4 are not applicable..

Verify specific activity of the primary

NOTE------

coolant-s 100/E microcuries/ gram.

Sample to be l

taken after a minimum of 2

'EFPD and 20 days of power operation have 1

elapsed since' the reactor was i

last suberitical a

for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

184 days

'r i

l l

Crystal River Unit 3 3.3-26 Amendment

1 Specific Activity l

3.3.15 Figure 3.3.15-1 (Page 1 of 1) l DOSE EQUIVALENT I 131 Primary Coolant Activity Versus

)

Percent of RATED THERMAL POWER With Primary Coolant Specific Activity > 1.0 microcurie / gram DOSE EQUIVALENT l 131.

t

=...

=

1 l

l A

\\

L am i

u

\\

'UNACCEPTAB LE 4

OPERATION 1

i g

T

>T;--;

s D

.+

i.'

1.

._,,. \\

n

150 E

1

~~

3 i

x

\\'

(

~

\\

E

_ _.y

- t 5

i

r. 100

\\

E

\\

m s

s i

t.

1 P

z ACCEPTABLE x

OPERATION i

~~

g I

i w

.e e-0 N

30 40 50 00 70 30 to 100 i

PERCENT OF RATED THERMAL POWER P

Crystal River Unit 3 3.3-27 Amendment

e ECCS Trains o MODES 1, 2 and 3 i

3.4.2

},, <

l L

3.4 EMERGENCY CORE COOLING SYSTEMS (ECCS) 1 3.4.2~

ECCS Trains MODES 1. 2 and 3 1

I.'

LCO 3.4.2 Two ECCS trains shall be OPERABLE, teith each train comprised of:

1

a. One OPERABLE high pressure injectlon (HP1) train,

b. One OPERABLE low pressure injection (LPI) train, and i

c. One OPERABLE decay heat (DH) cooler.

t

........................N01E l

Two HPI pum>s may be deactivated in accordance with LCO 3.3.8,

.ow Temperature Overpressurization Protection i

features.

t APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLET!ON TIME A.

One ECCS train A.1 Restore train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

OPERABLE status.

B.

Required Action NO.I B.1 Be in MODE 3, 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met within required Completion Time.

ANQ B.2 Be in MODE 4.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s

f' l

I-Crystal River Unit 3 3.4 3 Amendment I

i I

ECCS Trains - MODE 4 3.4.3 3.4 EMERGENCY CORE COOLING SYSTEMS (ECCS) l 3.4.3 ECCS Trains - MODE 4 o

I LCO 3.4.3 One ECCS train shall be OPERABLE with:

a.

An OPERABLE high pressure injection (HPI) train, I

b.

An OPERABLE low pressure injection (LPI) train, and c.

An OPERABLE decay heat (DH) cooler.

............................... NOTE- ---- ------

1. High pressure injection isolation valves may be closed with their power supply breakers locked in the open position.

2. Two HPI pumps may be deactivated in accordance with LC0 3.3.8, low Temperature Overpressurization Protection feature.

APPLICABILITY:

MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i

A.

Required ECCS train A.1 Initiate action to 15 minutes l

l inoperable, restore ECCS train to OPERABLE status, i

m A,2 Maintain MODE 4 by Until restoration available means of heat of required ECCS removal.

train to OPERABLE status.

L g

t A.3 Restore inoperable HPl I hour train to OPERABLE status.

(continued) 1 1

1 1

l Crystal River Unit 3 3.4-7 Amendment

7 Pressurizer Water Level B 3.3.6 BASES APPLICABLE abnormal transient operation, rather than a safety limit, the SAFETY ANALYSIS value for pretsurizer level is nominal and is not adjusted (continued) for instrument error.

Evaluations performed for large break loss of coolant accident (LOCA), which assumed a higher maximum level than assumed for the LOFW event, have been made. The higher pres-surizer lev ~ assumed for the LOCA is the bases for the volume of 1,: for coolant released to the containment. The containmen

.ialysis performed using the mass and energy release demonstrated that the maximum resulting containment pressure was within design limits.

The maximum pressurizer water level limit satisfies the requirements of Selection Criterion 2 of the NRC Interim Policy Statement (Ref. 1) because it prevents exceeding the initial reactor coolant mass which is an input assumption of the ECCS analysis. The maximum water level also permits the pressurizer code safety valves to relieve steam for antici-pated pressure increase transients, preserving their function for mitigation. Thus Selection Criteria 3 is also indirectly applicable.

LCOs The purpose of the LCO is to ensure pressurizer OPERABILITY for pressure control for normal power operation and for anticipated design basis events as previously described.

Compliance with the LCO also ensures that the analysis for LOCA will be met.

APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on reactor coolant system I

temperature resulting in the reatest effect on pressurizer l

1evel and RCS pressure contro.

Thus applicability has been designated for MODES I and 2.

The applicability is also provided for MODE 3 with RCS temperature greater than 283'F.

The purpose is to provent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as RC pump startup.

The temperature of 283'F has been designated as the cutoff for applicability because LC0 3.3.8, low Temperature Overpressurization Protection Features, provides a requirement for pressurizer level at 283'F. With RCS temperature s 283'F, LCO 3.3.8 Low Temperature Overpressurization Protection Features requires further restrictions on maximum pressurizer water level.

LC0 3.3.8 also contains appropriate actions to be taken in the event water level cannot be maintained less than the limit.

(continued)

Crystal River Unit 3 B 3.3-27 Revision

tli Pressurizer Water Level l

B 3.3.6 i

BASES (continued)

ACTIONS Ad With water level in excess nf the maximum limit, action must be taken to restore pressurizer operation to within the bounds assumed in the analysis.

This is done by restoring the pressurizer water level to within the limit.

The one hour Completion Time is based on engineering judgement.

It t

is considered to be a reasonable time for draining excess liquid.

B.1 and B.2 If the water level cannot be restored, reducing core power constrains heat input effects that drive pressurizer insurges i

that could result from an anticipat]d transient.

By reducing power and reactor coolant temperature to at least MODE 3, the l

thermal energy of the reactor coolant mass is reduced which provides compensation for LOCA mass and energy releases.

The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allotted to reach MODE 3 is a reasonable time based on operating experience to reach MODE 3 from full power without challenging safety systems and operators.

Further pressure and temperature reduction to an RCS temperature 1

' i 283*F places the plant into a condition where the LC0 is not applicable.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time to reach the non applicable i

MODE is reasonaole based on operating experience.

L SURVEILLANCE SR 3.3.6.1 REQUIREMENTS i

This surveillance requires pressurizer water level to be i

verified within the maximum limits on a periodic basis. The r

i surveillance is performed by observing indicated level, The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on engineering judgement and i

industry accepted practice.

t REFERENCES 1.

52 FR 3788, NRC Interim Policy Statement on Technical l

Specification Improvements for Nuclear Power Reactors, l

February 6, 1987.

l t

Crystal River Unit 3 B 3.3-28 Revision

I l

LTOP features l'

+

B 3.3.8 l

L B 3.3 REACTOR COOLANT SYSTEM (RCS) j B 3.3.8 Low Temoerature Overoressurization Protection Features

]

BASES BACKGROUND The purpose of the low temperature overpres:ure protection 1

(LTOP) features LCO is to limit reactor coolant pressure at low temperatures to levels which will not compromise reactor L

pressure boundary integrity. The reactor vessel is the i

limiting com)onent for demonstrating that protection is provided. T1e reactor vessel material is less tough at i

reduced temperatures than at normal operating temperature.

As reactor vessel neutron irradiation accumulates, the vessel 1

material becomes less resistant to stress at low temperatures.

Stresses are therefore maintained low and i

increased as temperature increases during plant heatup.

Requirements for all plants were developed by the NRC (Ref.

1) which suggested pressure control measures and required that single equipment failures or single operator errors should not cause the limits of 10 CFR 50, Appendix G (Ref. 2) 4 to be exceeded.

Subsequent to the Appendix G requirement the NRC issued Generic Letter 88 11 (Ref 3) which allows relaxation of the LTOP limits.

System evaluations and stress analysis supporting relaxed LTOP limits (Ref. 4) have been developed for Crystal River Unit 3.

Although the LTOP limits have been relaxed, plant operational maneuvering must be controlled so the 10CFR50 Appendix G heatup and cooldown PT limits are not violated.

i Overpressure protection given by the LC0 is provided by reducing the PORY setpoint, and ensuring that pressurizer level is maintained below a maximum limit. The level provides a vapor space (steam or nitrogen), which can accommodate insurges without rapidly increasing pressure.

The level limit this provides for a time period to allow the operator to stop the cause of the increase.

The PORV, with its reduced setsoint, is the overpressure protection device i

which provides sackup to the operator in terminating increasing pressure events. The approach used to protect the vessel also requires deactivating HPI and CFTs because the PORV and pressurizer level are not fully capable of preventing overpressurization were these systems to be inadvertently actuated.

With HPI deactivated the ability to provide core cooling is restricted.

To balance the possible need for core cooling the LC0 does not require the ma Mup system to be deactivated. At the lower pressures associated with LTOP and the expected decay heat levels, the makeup system can provide adequate flow via the makeup control valve, or it can be used in the interim until HPI can be reactivated.

l l

(continued)

Crystal River Unit 3 8 3.3-33 Revision E

LTOP Features 8 3.3.8 l

s BASES (continued)

APPLICABLE Although analyses of LTOP events are not described in the SAFETY ANALYSIS FSAR, analyses have been performed in response to HRC i

requests to demonstrate that the reactor vessel is adequately protected against overpressurization during shutdown.

l Transients potentially capable of overpressurizing the reactor coolant system have been identified and evaluated.

Postulated transients included inadvertent high pressure

{

injection actuation; opening core flooding system discharge valves; energizing the pressurizer heaters; failing the makeup control valve open; temporary loss of decay heat removal; reactor coolant thermal expansion caused by RC pump start causing heat transfer from hot steam generators; and addition of nitrogen to the pressurizer.

Two transients which could result in exceeding LTOP limits in less than 10 minutes are inadvertent HPI actuation or inadvertent discharge of two core flood tanks. The analyses also show that the PORV cannot maintain RCS pressure below 4

the LTOP limit if more than one HPJ pum) is started, j

Consequently the LCO requires that HPI 3e defeated at low temperatures and 3ressures. The CFTs are also isolated for i

similar reasons,10 wever, the analyses show that the effects of a discharge of two CFIs are important over a narrower i

range (175'F and below) than the range of tl.a LC0 (283'F and below).

For other events, operator action is assumed after 10 minutes to preclude overpressurization.

Evaluations show

[

that time for operator actions is adequate or the events are self limiting (i.e., will not exceed the LTOP limit).

Analyses for uperator response time show that the pressurizer i

must be maintained at or below 220 inches to allow a 10-minute action time for correcting transients.

The fracture mechanics analysis show that the vessel is adequately 3rotected when reactor coolant )ressure is maintained at or

)elow 555 psig.

Consequently tie PORV overpressure protection setpoint has been fixed at 555 psig.

The applicability temperature of 283'F has been established by fracture mechanics analyses. Above this temperature i

reactor vessel pressure protection is provided by the pressurizer code safety valves. The pressure (555 psig) and temperature (283'F) have been determined for the vessel materials with irradiation accumulation equivalent to 21 effective full power years (EFPY) of operation.

(continued) i t

I Crystal River Unit 3 8 3.3-34 Revision

l LTOP Features B 3.3.8 i

BASES j

APPLICABLE The LTOP features are used to prevent pressure increase SAFETY ANALYSIS transients from exceeding allowable limits. Although a low i

(continued) temperature overpressure transient is not a design basis l

accident, Selection Criterion 3 of the NRC Interim Policy j

Statement (Ref. 5) applies because prevention of transient i

ovarpressure events leading to non ductile failure assures i

pressure boundary integrity.

l LCOs The LC0 requires the pressurizer level tc be maintained at or below 220 inches to provide time for operator action to

)

prevent transients from exceeding the overpressure protection i

limit of 555 psig.

The PORV is to be OPERABLE with a j

setpoint at the overpressure limit, and the block valve j

should be open to ensure a clear flow path.

l' Since inadvertent actuation of HPI or CFTs cannot be protected by the PORV in combination with pressurizer level, these systems must be deactivated.

For the HPl system, the j

preferred method of deactivation to be used is to close the l

HPI injection valves and fix their power supply breakers in j

l the open position.

Deactivation of other components is also J

permitted by the LCO, but if powered components are used, power must be removed because this ensures positive i

prevention of inadvertent actuation. The LCO also permits i

HPI surveillance testing for components since pump or valve testing can proceed by alternating the system deactivation I

from pumps to valves. Testing must not permit HPI injection flow to enter the RCS.

The provisions of LCO 3.0.3 are not applicable, if the LCO is violated, changing to a lower MODE may be impractical and a

may not provide improved LTOP protection.

If an LTOP feature t

is inoperable, the correct approach is to restore the LTOP feature or take other specific ACTIONS.

APPLICABil:TY The LCO is applicable whenever the RCS is at low temperature and can be subjected to pressure increases.

Thus the LCO l

does not apply when the reactor vessel head is completely detensioned or removed. The value for the temperature is determined from fracture mechanics analyses.

The LCO is not applicable for operating conditions above the 283*F 6

temperature because the pressurizer code safety valves are able to provide overpressure protection.

(continued)

Crystal River Unit 3 8 3.3-35 Revision

[

1 i.

LTOP features B 3.3.8 I

BASES l

APPLICABILITY CfT isolation is only required when the CFT pressure is (continued) greater than the allowable pressure for the existiiig RC temperature (i.e., heatup and cooldown pressure temperature limits per 10CfR50 A>pendix G and shown in the PRESSURE /

TEMPERATURE LIMITS REPORT). This note permits the CFT J

discharge check valve survelliance to be performed under l

these conditions.

LTOP limits cannot be violated because the

.(

heatup and cooldown PT limits are more restrictive.

l ACTIONS A.I. B.1 and B.2 f

With the pressurizer level greater than 220 in, the time for operator action is reduced, and the postulated transient i

event which is most affected is a failure of the makeup control valve which permits relatively rapid filling of the I

aressurizer.

Pressurizer level restoration is required in 1 l

lour.

If that cannot be accomplished, the makeup control I

valve and its associated isolation valve must be closed and maintained closed within an additional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This Required Action limits makeup, which is not recuired with a high pressurizer level and permits cooldown anc depressurization to continue.

Heatup should be curtailed because heat addition may cause reactor coolant density decrease and increasing pressurizer level. The Completion i

Times are based on engineering judgement and operating experience that these activities can be accomplished in these i

time periods, and on engineering evaluations (Ref. 4) that indicate an event requiring LTOP protection is not likely in the time allowed for the Required Actions.

C.I. D.1. and 0.2 With the PORV inoperable, overpressure relieving capability j

is lost and restoration of the PORV in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is required.

If that cannot be accomplished the ability of the makeup system to add water should be restricted within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

By reducing the makeup tank level to 70 inches and by deactivating the low low level interlock to the BWST, insufficient water volume is available to cause the LTOP limit to be exceeded by an inadvertent makeup control valve opening. The makeup system is not deactivated because it may be needed to continue to manage the RCS inventory. The Completion Times are based on engineering judgement and operating experience that these activities can be accomplished in these time periods, and on engineering l

evaluations (Ref. 4) that indicate an event requiring LTOP protection is not likely in the time allowed for the Required Actions.

(continued)

Crystal River Unit 3 B 3.3-36 Revision

u r

LTOP Features i

L B 3.3.0 BASES

(

ACTIONS C.I. 0.1. and D.2 (Cont'd)

Some PORV testing or maintenance can only be performed at f

plant shutdown.

These activities are permitted provided measures are taken to compensate for the PORV unavailability, i

With HP! and CFT deactivated per the LCO, the limitirg transient which could cause LTOP limits to be exceeded is excessive makeup.

Required Action 0.1 requires restricting the volume of makeup available from the makeup tank or BWST l

to be less than that which could cause the RCS pressure to i

exceed the LTOP limits (due to pressurizer insurge and compression of the vapor space).

E.1 and F.1 l

f With one or both HPI trains active, both actions require deactivation.

Required Action E.1 allows deactivation using i

i l

a variety of methods as may be needed to fit various l

operating configurations of the combined makeup HPl system design.

If powered components are used to accomplish i

deactivation, power should be removed to assure positive I

lockout so that inadvertent ES actuation cannot cause HPl.

If Required Action E.1 cannot be accomplished in the Completion Time of I hour, Required Action F.1 specifically i

requires closing and removing power from the HPI injection e

valves within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Failure to deactivate HPl per the LCO is not expected, however, since inadvertent actuation is the event of greatest significance (causes the

[

greatest pressure increase in the shortest time), emphasis is placed on deactivation by these Required Action statements.

The Completion Times are based on engineering judgement and operating experience that these activities can be accomplished in these time periods, and on engineering evaluations (Ref. 4) that indicate an event requiring LTOP protection is not likely in the time allowed for the Required Actions.

l I

G.1. H.1 and H.2 With the CFis unisolated, Required Action G.1 requires isolation within I hour, if isolation cannot be accomplished Required Action H.1 provides two options, either of which should be accomplished in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The CFT pressure of 600 t

psi cannot cause LTOP limits to be exceeded if the RCS temperature is greater than 175'F with a full discharge of both tanks. Depressurizing the CFTs below the LTOP pressure limit of 555 psig also prevents exceeding the LTOP limits.

(continued)

Crystal River Unit 3 B 3.3-37 Revision

b LTOP Features 4

i B 3.3.8

.I L

q BASES i

l ACTIONS G.1. H.1 and H.2 (Cont'd)

The Completion Times are based on engineering judgement and operating experience that these activities can be accomplished in these time periods, and on engineering evaluations (Ref. 4) that indicate an event requiring LTOP j

protection is not likely in the time allowed for the Required Actions.

1.1. 1.2.1. 1.2.2. and 1.2.3 l

With the LTOP features inoperable for reasons other than j

those cited in the above conditions, the system must be restored to an acceptable condition within I hour or in the next12hourstheRCSmustbedepregsurizedandaventsize l

equivalent to an orifice of 0.75 in must be opened. These actione are principally directed toward conditions where more than one of the LCO conditions is violated. Of the multiple conditions the combination of most interest for these Required Actions are A (pressurizer high level) and C (PORV inoperable).

The LC0 is not applicable when the RCS is adequately vented.

[

Single or multiple vents may be used. A vent size equivalent to an orifice area of 0.75 in2 has been specified.

The vent size has been calculated assuming 100 psi back pressure.

It j

is considered less likely that HPI or the CFTs cannot be deactivated. Because makeup may be required the vent size l

stipulated has been developed to accommodate inadvertent full makeup system operation. A 0.75 in2 vent area is capable of

{

relieving the full flow of one makeup pump with a wide open control valve and preventing the LTOP pressure limit from i

L being exceeded.

The PORV, which has a larger area, may be used for venting by opening and locking it open.

Removing l

the PORV for maintenance or testing also accomplishes ventin.

TheventmustbeaccompaniedbydeactjvatingHPI and CF s since neither the PORY nor the 0.75 in orifice J

equivalent is capable of maintaining pressure below LTOP limits if these systems are inadvertently actuated.

l The Completion Times are based on engineering judgement and operating experience that these activities can be accomplished in these time periods, and on engineering i

evaluations (Ref. 4) that indicate an event requiring LTOP r'

protection is not likely in the time required for the Required Actions.

I (continued) o l

Crystal River Unit 3 8 3.3-38 Revision 1

i LTOP Features B 3.3.8 f

i BASES i

d SURVEILLANCE SR 3.3.8.1 REQUIREMENTS Verification of pressuri er level by observing control room indications (or equivalent) assures that a steam or nitrogen bubble of sufficient size is available to reduce the rate of t

pressure increase from potential transients. The 30 minute surveillance frequency during heatup and cooldown is to be performed for the LCO applicability period when temperature changes can cause pressurizer level variations and may be i

discontinued when definitions given in plant procedures for-defining the end of these conditions are satisfied.

Thereafter, surveillance is required at 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months intervals.

The surveillance frequencies are based on engineering judgement and industry accepted practice.

SR 3.3.8.2 Verification that the block valve is open ensures an open flow path to the PORV. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on engineering judgement and industry accepted practice.

t SR 3.3.8.3 and SR 3.3.8.4 Verification that HPI is deactivated and the CFTs are i

isolated ensures that inadvertent injection or discharge will not cause a violation of the LTOP pressure limit. The 12 i

hour Frequency is based on engineering judgement and industry l

accepted practice.

I SR 3.3.8.5 The RCS vent is to be verified open for relief protection l

when required per Required Action 1.2.

The verification l

frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on engineering judgement and industry accepted practice.

i SR 3.3.8.6 L

Performance of a CHANNEL FUNCTIONAL TEST is required to ensure the PORV setpoints are proper prior to use of the PORV for LTOP. The Frequency permits testing at any time within i

1 31 days of use and allows testing during cooldown prior to entry into LTOP applicability.

The Frequency is based on 7

engineering judgement and indestry accepted practice.

1 (continued)

L L

Crystal River Unit 3 B 3.3-39 Revision

n i

h i

p LTOP Features

[

B 3.3.8 f

BASES

)

SURVEILLANCE SR 3.3.8.7 REQUIRMENTS (continued)

Surveillance Requirement 3.3.8.6 is the performance of a i

CHANNEL CAllBRATION every 18 months.

The CHANNEL CALIBRATION for the LTOP setpoint ensures that the PORY will be actuated j

at the appropriate RCS pressure by verifying the accuracy of the instrument string. The calibration can only be performed i

during a shutdown.

The Frequency is based on engineering t

judgement and industry accepted practice.

REFERENCES 1.

NRC letter dated October, 1976, J. F. Stolz to J. T.

Rogers of Florida Power Corporation, " Transmittal of j

Analyses of Low Temperature Overpressurization Transients for Crystal River Unit 3."

)

i 2.

10 CFR 50, Appendix G, " Fracture Toughness l

Requirements."

3, Generic letter 88 11, 'NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operation."

i 4.

B&W Summary Report 51-117643101, " Crystal River 3 Reactor Vessel Low Temperature Overpressure Protection (LTOP)," September 1989.

5.

52 FR 3788, NRC Interim Policy Statement on Technical l

Specification Improvements for Nuclear Power Reactors, February 6, 1987.

l l

L 1'

1 4

Crystal River Unit 3 B 3.3-40 Revision l

q w

y M'

Pressurizer Safety Valves F

B 3.3.9

'F i

B ?.3 REACTOR COOLANT SYSTEM (RCS)

\\

8 3.3.9 Pressurizer Safety Valves

,1 BASES i

(

BACKGROUND The purpose of the two spring loaded pressurizer code safety j

valves is to provide reactor coolant system (RCS) overpres-1 sure protection. Operating in conjunction with the reactor i

protection system two valves are used to assure that the i

Safety Limit of 2750 psig is not exceeded for analyzed transients during operation in MODES 1, 2, and 3.

One code j

l safety valve is adequate in MODES 4 and 5.

Overpressure I

protection is also provided by operating procedures and low j

temperature overpressurization protection (LTOP) system q

equipment in MODES 4 and 5.

LTOP is provided by reducing the PORV setpoint and isolating certain functions.

For more j

detail see Bases B 3.3.8 (Low Temperature Overpressurization j

Protection Features).

The code safety valves discharge steam j

from the pressurizer to a drain tank located in the containment. Should the drain tank or connecting piping be unable to accept the code valve discharge, valve operation i

would result in relief to the containment atmosphere. While

)

discharge to the containment atmosphere is highly undesir-j able, reactor coolant pressure boundary integrity would 1

remain protected.

t APPLICABLE All accident analyses in the FSAR which require safety valve SAFETY ANALYSIS actuation assume operation of both pressurizer code safety valves to limit increasing reactor coolant pressure. The overpressure )rotection analysis (Ref.1) is also based on L

operation of )oth code safety valves and assumes that the valves open at the high range of the setting (2500 psig system design pressure plus 1%).

These valves must accom-l modate pressurizer insurges which could occur during the startup, rod withdrawal, ejected rod, loss of main feedwater, and main feedwater line break accidents. The startup l

accident establishes the minimum code safety valve capacity.

The startup accident is assumed to occur at less than 15%

I l

power in MODE I and could occur in the lower bound of MODE 2 l

when the control rod drive trip breakers are closed in the transition from MODE 3 to MODE 2.

Single failure of a code i

safety valve is neither assumed in the accident analysis nor l

required to be addressed by the ASME code. Compliance with this specification is required to assure that the accident e

L analysis and design basis calculations remain valid, 1

(continued) l l

Crystal River Unit 3 B 3.3-41 Revision l

I Pressurizer Safety Valves i

B 3.3.9 j

BASES i

APPLICABLE The pressurizer code safety valves satisfy the requirements L

SAFETY ANALYSIS of Selection Criterion 3 of the Interim Policy Statement (continued)

(Ref. 3) because operation of two valves within their allowed limit setting ensures that they will function to provide RCS overpressure protection for analyzed transients of the design j

basis.

Failure to function could challenge the integrity of a fission product barrier, i

LCOs The two pressurizer code safety valves are set to open at the RCS design pressure (2500 psig) and within the ASME specified i

tolerance to avoid exceeding the maximum RCS design pressure l

Safety Limit, to maintain accident analysis assumptions, and to comply with ASME requirements.

The u)per and lower pressure tolerance limits are based on t1e i 1% tolerance requirements (Ref. 2) for lifting pressures above 1000 psig.

The limit protected by this specification is the reactor coolant pressure boundary Safety Limit of 110% of design L

pressure.

Ino)erability of one or both valves could result

[

in exceeding t1e Safety Limit were a transient to occur.

The consequences of exceeding the ASME pressure limit could j

include damage to one or more RCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation.

l APPLICABILITY OPERABILITY of the two vrives is required in MODES 1, 2, and 3 because their combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents. MODE 3 is included although the listed

[

accidents would not require both code safety valves to function for protection in MODE 3.

It is conservatively l

included because the accident could occur near the lower i

bound of MODE 2.

In MODES 4 and 5 one code safety valve is required to accommodate the pressurizer insurges which could result from all the sources identified (i.e. residual heat, pump energy, pressurizer heaters) and thus the LCO is not applicable in MODES 4 and 5.

Overpressure protection is not required in MODE 6 with the reactor vessel lead removed.

ACTIONS L1 With one pressurizer code safety valve ino)erable, res-toration must take place in 15 minutes.

T1e 15 minutes to restore a safety valve to OPERABILITY is based on engineering judgement.

(continued)

Crystal River Unit 3 8 3.3 42 Revision l

r Pressurizer Safety Valves B 3.3.9 l.

BASES ACTIONS B.1 and B.2 (continued)

If the Required Action cannot be met within the required Completion Time, the plant must be placed in a MODE in which the requirement does not apply.

This is done by placing the 31 ant in at least MODE 3 in six hours and in MODE 4 in 12 1ours. The six hours allotted to reach MODE 3 is a reason-abic time based on operating experience to reach MODE 3 from full power without c1allenging safety systems and operators.

j Siinilarly, the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allotted is a reasonable time to reach MODE 4 considering that a 31 ant can easily cooldown to this MODE in this time frame.

T1e change from MODES 1, 2 or 3 to MODE 4 reduces the RCS energy (core power and pressure),

lowers the potential for large pressurizer insurges and thereby removes the attendant need for overpressure protec-tion by two pressurizer code safety valves.

SURVEILLANCE SR 3.3.9.1 REQUIREMENTS Section XI of the ASME code (SR 3.0.5)least once every 5 requ each pressurizer code safety valve at years. This surveillance includes setpoint testing and demonstration of OPERABILITY through non popping (hydraulic assist) techniques.

Because testing is done at reduced pressure, the test lift setting pressure shall correspond to ambient conditions of the valve at nominal operating tempera-I ture and pressure.

The frequency is based on engineering judgement and industry accepted practice.

REFERENCES 1.

B&W To>ical Report BAW-10043, " Overpressure Protection for Ba) cock & Wilcox Pressurized Water Reactors, J. D.

Carlton, May 1972.

2.

ASME Boiler & Pressure Vessel Code, Section !!!,

Nuclear Vessels," and Section XI " Rules for Inservice Inspection of Nuclear Power Plant Components.

l 3.

52 FR 3788, NRC Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.

1 l

l Crystal River Unit 3 8 3.3-43 Revision l

I l

Relief Yalves i

B 3.3.10 l

B 3.3 REACTOR COOLANT SYSTEM (RCS)

B 3.3.10 Relief Valves e

i BASES BACKGROUND The pressurizer is equipped with three devices for aressure

{

relief functions:

two ASME code safety valves whic1 are i

safety grade components and one power operated relief valve I

(PORV) which is not a safety grade device. The PORV is an electromagnetic pilot operated valve that is automatically opened at a specific set pressure when the pressurizer c

pressure increases and is automatically closed on decreasing pressure.

The PORV may also be manually operated using controls installed in the control room, j

An electric motor o>erated, normally open, block valve is installed between tie pressurizer and the PORV.

The function of the block valve is to isolate the PORV.

Block valve closure is accomplished manually using controls in the control room and may be used to isolate a leaking PORY to permit continued power operation. Most importantly, the i

block valve is used to isolate a stuck open PORV small break loss of coolant accident (SBLOCA) to terminate the reactor coolant system (RCS) depressurization and coolant inventory l

loss.

i t

The PORV, its block valve, and their controls are powered from normal power supplies but are also capable of being powered from emergency supplies.

Power supplies for the PORV

)

are separate from those for the block valve.

Power supply requirements are defined in NUREG 0737, Paragraph III G. I 1

(Ref. 1).

l The PORV setpoint is set greater than the high pressure i

reactor trip setpoint and less than the opening setpoint for i

the pressurizer code safety valves.

This setpoint was required by the NRC in IE Bulletin 79 05B (Raf. 2). The l

purpose of the relationship of these setpoints is to limit the number of transient pressure increase challenges which might open the PORV, which, if opened, could fail in the open L

position.

The P0rV setpoint is greater than the high pressure reactor trip setpoint. Consequently a pressure j

(continued)

Crystal River Unit 3 8 3.3-44 Revision l

i

,l Relief Yalves B 3.3.10 BASES i

v.

i BACKGROUND increase transient would cause a reactor trip, reducing core (continued) energy, and for many expected transients, prevent the l

pressure increase from reaching the PORV setpoint. The PORY r

setpoint thus limits the frequency of challenges from j

transients and limits the possibility of a SBLOCA from a j

failed open PORV.

Placing the setpoint below the pressurizer j

safety valve opening setpoint reduces the frequency of r

i challenges to the safety valves, which unlike the PORV cannot be isolated if they were to fail open. Accurate control of i

the PORV setpoint is therefore important for limiting the i

possibility of SBLOCA.

e The primary purpose of this LCO is to ensure that the PORV, its setpoint, and the block vcive are operating correctly so l

the potential for SBLOCA through the PORV anthway is mini-t mized, or if a SBLOCA were to occur throug1 a failed open j

PORV the block valve could be manually operated to isolate i

the path.

l The PORV may also be manually operated to depressurize the RCS as deemed necessary by the operator in response to normal.

or abnormal transients. The PORY may be used for depres-surization when the pressurizer spray is not available; a' condition that would be encountered during loss of offsite power.

Steam generator tube rupture is one event that may require use of the PORY if the sprays are unavailable.

The PORV may also be used for feed ar.d bleed core cooling for multiple equipment failure events that are not within the

,l design basis, such as total-loss of feedwater.

The PORV functions as an automatic overpressure device and l

limits challenges to the code safety valves. Although the i

PORY acts as an overpressure protection device, safety j

t analyses do not take credit for PORV actuation, but do take credit for the safety valves.

The overpressure protection l

function of the PORY during MODES 1 and 2 is an operational r

l function and is not addressed by Technical Specifications, The PORY also provides low temperature overpressure protec-tion (LTOP) during heatup and cooldown. '400 3.3.8 addresses this function.

1 (continued) i l

l t

Crystal River Unit 3 8 3.3-45 Revision l,

3-.

-e--

w

+-

.m -

ir'

s l

Relief Valves B 3.3.10 6

BASES (continued)

L APPLICABLE There are no explicit FSAR safety analyses of a SBLOCA L

SAFETY ANALYSIS through the PORY path. The PORY SBLOCA is not a design basis e

event, however the break size is bounded by the spectrum of r

piping breaks analyzed for plant licensing. Because the PORY SBLOCA is located at the top of the pressurizer the RCS response characteristics are different from RCS loop )iping L

breaks; analyses have been performed to investigate tiese characteristics.

The possibility of a SBLOCA through The PORV is reduced when the PORV flow path is OPERABLE and tha PORV opening setpoint is verified to be reasonably remote fram expected transient challenges. The possibility is minimized if the flow path is r

isolated.

The PORV opening setpoint has been established in accordance with IE Bulletin 79-05B (Ref. 2).

No specific safety analyses were performed to determine the setpoint, however, it has been set so expected RCS pressure increases from anticipated transients will not challenge the PORV, minimiz-ing the possibility of SBLOCA through the PORV.

Overpressure protection analyses do not take credit for the PORV opening and therefore are not pertinent to the PORV.

l The design basis accidents reported in the FSAR safety 1

l analyses do not take credit for the PORV for mitigation.

However operational analyses that support the emergency operating procedures utilize the PORV to depressurize the RCS for mitigation of steam generator tube rupture (SGTR) when the pressurizer spray system is unavailable (loss of offsite power).

FSAR safety analyses for SGTR have been performed assuming that offsite power is available and thus sprays are available.

j l

The PORV and its block valve do not satisfy the requirements l

of the Selection Criterion of the NRC Interim Policy State-t l

ment (Ref. 3). This specification was evaluated using r

insights gained from reviewing representative probabilistic risk assessments.

The PORV and its block valve are deemed important to risk.

L (continued) i Crystal River Unit 3 B 3.3 46 Revision l,

Relief Valves i

L B 3.3.10 l

t BASES (continued)

LCOs The LCO requires the PORV and its block valve to be OPERABLE.

By ensuring that the PORY opening setpoint is correct the i

PORY is not subject to frequent challenges from possible pressure increase transients and therefore the possibility of a SBLOCA through a failed open PORV is not an event of an l

undesirable frequency. The block valve is required to be OPERABLE so it may be used to isolate the flow path if the j

PORV is not OPERABLE.

If the block valve is not OPERABLE, the PORV may be used for isolation.

APPLICABILITY The PORV will automatically open when the RCS pressure increases to the PORV setpoint.

Imbalances in the energy output of the core and heat removal by the secondary system i

can cause the RCS pressure to increase to the PORV opening setpoint. Pressure increase transients can occur any time the steam generators are used for heat removal.

The most rapid increases will occur at higher operating power and i

pressure conditions of MODES 1 and 2.

Pressure increases are less prominent in MODE 3, because the core input energy is reduced, but the RCS pressure is high.

Therefore the applicability is pertinent to MODES 1, 2, and 3.

When both pressure and core energy are decreased the pressure surges become much less significant and the LC0 is l

not applicable ir. MODE 4, partly because the consequences are less severe and partly because the time spent in heatup and

(

cooldown is short. The PORV setpoint is reduced for low temperature overpressurization protection (LTOP) at lower I

pressures during heatup and cooldown.

LTOP is ap)11 cable i

during MODES 4, 5, and 6 with the reactor vessel lead in place. As such, LC0 3.3.8 (LTOP Features) addresses the PORV requirements in these MODES.

l Anticipated pressure increase transients caused by secondary system upsets which are pertinent to MODES 1, 2, or 3 include:

t t

- Loss of electrical load.

- Turbine trip.

- Loss of main feedwater.

- Loss of condenser vacuum, i

- Inadvertent closure of main steam isolation valve (s).

(continued)

Crystal River Unit 3 B 3.3 47 Revision l

o Relief Valves pp B 3.3.10 BASES (continued) b ACTIONS in keneral, the Required Actions for each of the Conditions (P0 V inoperable, block valve inoperable, or both inoperable) utilize tse same concept:

Restore OPERABILITY, or (if that is not possible) i

- Isolate the flow path (isolation ensures that a transient challenge will not cause the PORV to fail open resulting in a SBLOCA), or (if restoration and isolation are not pos-p sible)

Reduce core power and RCS pressure (by reducing the energy level the pressure increase of a secundary side transient o

l are less likely to challenge the degraded components in the flow path).

The Required Actions permit continued operation with either or both valves inoperable as long as the flow path is isolated.

A.1 and A.2 With the PORV inoperable, either the PORY must be restored or the flow path isolated within one hour.

In this' instance, as compared to Required Actions for other Conditions, the block valve should be closed but power need nol be removed from the block valve. This Required Action is because the block valve is OPERABLE. The Completion Times are based on engineering judgement and plant operating experience.

B.I. B.2.1. B.2.2. B.3.1. and B.3.2 i

If the block valve is inoperable or the PORV and the block valve are inoperable, the inoperable components must be restored, or the flow path isolated and power supply removed.

The Completion Times are based on engineering judgement and plant operating experience.

C.1 and C.2 If the Required Action cannot be met within the required Completion Time, the plant must be placed in a MODE in which the requirement does not apply.

This is done by placing the I

plant in at least MODE 3 in six hours and in MODE 4 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The six hours allotted to reach MODE 3 is a reason-eble time based on o>erating experience to reach MODE 3 from full power without cla11enging safety systems and operators.

(continued)

L Crystal River Unit 3 B 3.3-48 Revision l

I

[

o 3

I I

Relief Valves I

I B 3.3.10 i'

BASES e

ACTIONS C.1 and C.2 (Cont'd)

(continued)

Similarly, the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allotted is a reasonable time to i

reach MODE 4 considering that a plant can easily cooldown to l

this MODE in such a time frame, in MODE 4 RCS energy (core power and pressure) is reduced to a minimum and decay heat removal can now be provided by the decay heat removal system, I

which eliminates the possibility for secondary plant upsets.

l t

l SURVEILLANCE SR 3.3.10.1 t

REQUIREMENTS Block valve cycling verifies that it can be closed if needed.

The basis for the frequency is ASME XI (Ref. 4).

Block valve cycling is not performed when it is closed for isoittion; cycling could increase the hazard of an existing degraded i

flow path.

SR 3.3.10.2 Surveillance Requirement 3.3.10.2 is the performance of a CHANNEL CAllBRATION every 18 months. The CHANNEL CALIBRATION ensures the PORV setpoint is appropriately established above the RCS high pressure trip setpoint and thus remote from i

transient pressure challenges. The calibration also ensures that the PORY will open below the pressurizer code safety valve setpoint, thus limiting challenges to the code safety valves. The calibration can only be performed during shutdown.

The frequency is based on engineering judgement and industry-accepted practice.

REFERENCES 1.

NUREG 0737

Clarification of TMI Action Plan Require-L ments," November, 1980.

2.

NRC IE Bulletin 79 05B, 4/21/79.

3.

52 FR 3788, NRC Interim Policy Statement on Technical i

Specification improvements for Nuclear Power Reactors, February 6, 1987.

4.

ASME Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components."

t Crystal River Unit 3 8 3.3-49 Revision l

r i

I RCS Leakage B 3.3.11 B 3.3 REACTOR COOLANT SYSTEM (RCS) i l,

B 3.3.11 RCS Leakagg f

BASES l

BACKGROUND The RCS is comprised of components whose joints are made by I

welding, bolting, rolling and pressure loading.

The RCS is j

isolated from other plant systems by valves. During plant life, these interfaces can produce varying amounts of reactor e

coolant leakage, through either normal operational wear or mechanical deterioration. The pur)ose of the RCS leakage LCO is to permit system operation in tie presence of leakage from these sources in amounts which do not compromise safety. The LCO defines the types of leakage and allowable li. nits for L

leakage. This LLO is required to protect the reactor coolant pressure boundary against degradation, which ensures the RCS integrity for maintaining core cooling.

Leakage monitoring is an indicator of RCS integrity and can be performed 1

I frequently during operation.

Leakage monitoring is com-i plementary to inservice inspections which are performed periodically at outages.

Other related LCOs give limits for leakage at specific 3

locations.

LCO 3.3.12 (RCS Primary to Secondary Leakage),

i specifies limits for steam generator tube leakage, and LC0 3.3.13 (RCS Pressure Isolation Valve (PlV) Leakage),

l specifies valve seat leakage limits for certain valves that i

isolate the high pressure RCS from other low pressure systems.

LCO 3.3.14 (Leakage Detection Instrumentation),

specifies the requirements for the monitoring equipment used to detect leakage into the containment, l

J APPLICABLE Except for primary to secondary leakage (LCO 3.3.12) safety

]

SAFETY ANALYSIS analyses for design bases accidents do not address leakage.

Some design basis accidents, particularly those with an j

emphasis for offsite dose calculations such as steam genera-tor tube rupture, assume a pre existing 1 gpm primary to

)

i l

secondary leak and consequent activity in one generator as an i

l initial analysis condition.

j (continued)

{

\\

t l

Crystal River Unit 3 8 3.3-50 Revision ll l

n -

i RCS Leakage i

B 3.3.11 BASES b

i APPLICABLE Leakage is an indication of possible degradation of the RCS SAFETY ANALYSIS boundary. Thus Selection Criterion 3 of the NRC Interim (continued)

Policy Statement (Ref.1) is satisfied.

t LCOs a.

PRESSURE __PGUNDARY LEAKAGE PRESSUP1 BOUNDARY LEAKAGE, defined as leakage throuch a l'

non isolable fault in a RCS component body, pipe, or vessel wall (excluding reactor coolant pump (RCP) shaft seals, packing, and steam generater tube leakage), is not 4

allowed in any amount because it would be indicative of material deterioration.

Leakage of this type is unac-ceptable as the leak itself could cause further deteri-i oration, resulting in higher leakage.

Violation of this LCO could result in continued degradation of the reactor l

coolant pressure boundary, i

b.

UNIDENTIFIED LEAKAGE l

UNIDENTIFIED LEAKAGE is defined as reactor coolant leakage which is not identified. U) to 1 gpm of UNIDEN-TIFIED LEAKAGE is considered allowa)1e on the basis that it is a reasonable minimum detectable level for the I

containment air monitoring and containment sump level monitoring equipment.

l c.

IDENTIFIED LEAKAGE i

IDENTIFIED LEAKAGE is defined as leakage into closed systems connected to the RCS that is captured and recovered.

Up to 10 gpm of IDENTIFIED LEAKAGE is con-i sidered allowable because it provides for leakage from l

known sources which do not interfere with normal opera-L tion and which is well within the capability of the make-u) system to replenish.

This leakage includes leakage to tie containment from sources that are specifically known L

and located, but does not include pipe or vessel wall t

L leakage or RCP controlled bleed off (which is a normal process function and not considered leakage).

Violation of this LC0 could result in impairment of reactor coolant y

inventory control.

Leakage past the seats of pressure isolation valves (LCO 3.3.13) and primary to secondary leakage (LC0 3.3.12) are included in IDENTIFIED LEAKAGE.

(continued) l l

Crystal River Unit 3 3 3.3-51 Revision l

RCS Leakage

)

B 3.3.11 i

BASES (continued)

APPLICABILITY The RCS leakage LCO applies to MODES 1, 2, 3, and 4 to ensure that the applicable accident analysis assumptions remain valid and to minimize pressure boundary degradation from continued leakage. Leakage limits are not provided for MODES 5 and 6 because the reactor coolant pressure is far lower, making leakage less likely and less difficult to control; and because the mechanisms for offsite release have been reduced or eliminated. Accordingly, the potential consequences of reactor coolant leakage are far lower during these MODES.

ACTIONS Ad The general activities associated with the Condition A are:

- Quantifying or verifying the leakage rate.

- Identifying the source of any leakage.

- Determining what repair might be appropriate, and L

determining if the repair can be carried out at pressure or whether shutdown is required.

Of the activities associated with the Condition A, one of great importance is determining if PRESSURE BOUNDARY LEAKAGE exists.

If the leakage source cannot be identified, the leak is UNIDENTIFIED, and if there is doubt of the location a conservative assumption is that the leakage is from the pressure boundary.

l With RCS leakage outside limits for reasons other than PRESSURE BOUNDARY LEAKAGE, the leakage must be reduced. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time may permit repair or isolation depending on the source and the complexity of the re) air, The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on engineering judgement, and tie time i

l permits a reasonably stable period at operating pressure to I

identify the source and verify the quantity of leakage by inventory balance.

Inventory balance calculations require a stable condition; pressure reduction may lessen the ability L

for source identification.

B.1 and B.2 l

If RCS leakage (other than PRESSURE BOUNDARY LEAKAGE) cannot l

be reduced to the permissible limit within the required Completion Time, reducing power and pressure in sequence to MODE 3 and then MODE 5 are required.

By doing so the hazard associated with the leak is reduced.

1 (continued) 1 Crystal River Unit 3 B 3.3-52 Revision l

L 1

l RCS Leakage B 3.3.11 BASES l

l ACTIONS B.1 and B.2 (Cont'd) 3 In this condition, the plant must be placed in a MODE in l

which the LCO does not apply.

This is done by placing the l

)lant in at least MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 in 36 tours. The six hours allotted to reach MODE 3 is a reason-able time based on oserating experience to reach MODE 3 from full power without cla11enging safety systems and operators, i

r Similarly, the 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months allotted is a reasonable time to L

reach MODE 5 considering the plant can easily cooldown in such a time frame on one safety system train.' In MODE 5, pressure is reduced to the lowest possible value and tempera-F ture to below the boiling point at atmospheric 3ressure. The j

Required Action reduces the driving forete for tie leak and eliminates the possibility for contaminated steam leaks to P

the containment atmosphere and the possible deleterious i

L cutting action of steam on the material surrounding the leak.

C.1 and C.2 If any reactor coolant PRESSURE BOUNDARY LEAKAGE is detected.

-[

the reactor must be placed in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE l

5 in the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the leakage, reduces the factors which tend to degrade the pressure

{

l boundary, and most importantly reduces the potential for RCS e

l piping or vessel failure. The Bases for the Completion Times L

are the same as those for Required Actions B.1 and B.2.

As l

such, the Bases for Required Actions B.1 and B.2 are appli-cable to Required Actions C.1 and C.2.

l SURVEILLANCE SR 3.3.11.1 i

REQUIREMENTS l

An inventory balance is the most precise method for quantify-ing leakage rate.

Nominal instrument indications of the

[

l L

parpeters used to make the inventory balance calculation are i

I used without adjustments for instrument error. Although Other methods exist, and may be used, they give results that are less certain, and are not required to meet this speci-fication.

Seventy-two hours permits a reasonable interval for trending.

Steady state operation is required to perform l

a proper inventory balance; calculations during maneuvering are not useful and the surveillance is not required unless steady state is established.

For purposes of leakage I

determination by inventory balance, steady state is defined as stable RCS aressure, temperature, power level, pressurizer and makeup tan ( level, constant makeup and letdown and RC L

(continued)

{

Crystal River Unit 3 B 3.5-53 Revision l

RCS Leakage i

B 3.3.11 BASES SURVEILLANCE SR 3.3.11.1 (Cont'd)

REQUIREMENTS i

pump seal injection and return flows. As such, the Comple-i tion Time is based on engineering judgement and industry-l accepted practice, i

The NOTE stovides an exception to 3.0.4 permitting entry into MODES wit 1 stable conditions so the surveillance can produce accurate results, t

In addition to the required periodic steady state surveil-lance, an inventory balance is required to be made if the containment sump level leakage monitoring instrumentation is l

inoperable (Required Action B.1 of LCO 3.3.14).

REFERENCES 1.

52 FR 3788, NRC Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.

I t

i t

i Crystal River Unit 3 B 3.3-54 Revision l

q l

i F

RCS Primary to SecCndary Leakage B 3.3.12

]

B 3.3 REACTOR COOLANT SYSTEM (RCS)

B 3.3.12 RCS Primary to Secondary leakaae BASES BACKGROUND The purpose of the primary to secondary leakage LCO is to permit system operation in the presence of steam generator tube leakage in amounts which do not compromise safety.

Refer to tile Bases for LCO 3.3.11 for definitions of the leakage terms used.

Leakage into the secondary side of the steam generators has two effects:

1) it indicates a degrada-tion of the RCS boundary (experience indicates that in most cases the source of primary to secondary leakage is a single steam generator tube flaw), and 2) reactor coolant fission product activity is present in the secondary system.

Because the hot well is common to all feedwater trains, activity from any primary to secondary leak is mixed and spread throughout the entire secondary system. The value of I gpm allowable leakage has been established because it is a practical limit within the capability for detection and quantification. The importance of existing secondary leakage and activity is that a primary barrier to fission product confinement has been breached and a pathway to public release via the condenser t.ir ejector (s) exists. Monitoring for secondary leakage can be performed during o)eration thus providing an indication of the condition of the >arrier; it is complementary to the inservice tube inspection program that is performed during outages.

Leakage can be detected in a variety of ways including condenser offgas radiation monitors, secondary water chemistry analysis, or steam line radiation monitors.

The 1 gpm allowable primary to secondary leakage is included in the 10 gpm total allowable IDENTIFIED LEAKAGE rate.

It is nol in addition to the 10 gpm permissible IDENTIFIED LEAKAGE rate.

APPLICABLE The limit of I gpm primary to secondary leakage is assumed as SAFETY ANALYSIS a pre-existing condition for design basis event safety ana-lyses.

The leak is assumed to exist in only one generator.

However, because of feedwater train mixing in the hot well, the fission products are spread throughout the secondary icontinued)

Crystal River Unit 3 B 3.3 55 Revision l l

,e p-h i

I, RCSPrimarytoSecondarydeakage j

[

B 3.3.12 l$

.I

?

BASES

]

f.,.

APPLICABLE system. Of the analyses that are performed with the 1 gpm h

- SAFETY ANALYSIS pre-existing leak, the steem line break (outside of contain-L (continued) ment) event is the most limiting for site radiation releases.

t All of the inventory of the generator with the broken steam line and a portion of the inventory of the unaffected L

generator is released (rapid isolation terminates releases from the intact generator). Results show site boundary doses to be within acceptable limits.

Steam generator tube rupture (SGTR) is also analyzed with a concurrent I gpm leak in the opposite steam generator. The tube rupture flow rate is significantly greater than the 1-i-

gpm rata and the effect is relatively inconsequential.

c Safety analyses for operating B&W plants have not been f.

required to assume a loss of the condenser for SGTR.

However best estimate SGTR releases with no condenser available have been evaluated to support emergency operating procedures.

These analyses indicate 'that doses would be within limits and l

industry o)erating experience (such as the Ginna event)

E confirms t1at actual SGTR releases are' much lower than

.K predicted using conservative safety analysis methods.

Stress analyses have been performed for the once through j

steamgenerators'(OTSG).

Predictions are that the higher L;

stresses caused by harsh transients such as the steam line

[.

break will not cause' loss of-tube integrity.even for tubes thinned to the plugging limit.

This specification satisfies Selection Criterion 2 of the NRC r

InterimPolicyStatement(Ref.11)becausethE1..gpmleakrate limit is used as an input assumption for safeth. l. na L

L LCOs Primary to secondary leakage, assumed at the rate of 1 gpm, produced acceptable offsite doses in the steamline failure accident analysis. Violation of this.LC0 could void the t

accident analysis offsite dose calculations for the steamline failure.

The leakage also shows a breach of the pressure boundary exists and indicates that further tube weakening may occur.

(continued) i Crystal River Unit 3 B 3.3-56 Revision l.

4

+ '(' c RCS Primary to Secondary Leakage B 3.3.12 j> BASES.(continued) l' APPLICABILITY The primary to secondary leakage LC0 applies to MODES 1, 2, 3, and 4 to ensure'that leakage is detected and operation terminated in the event steam generator tube leakage exceeds the amount assumed in the steam line failure accident analy-sis and ensures that the accident analysis remains valid. Leakage limits are not:provided for MODES 5 and 6.because the reactor. coolant. pressure is far lower, reducing the primary l to secondary pressure differential and thus the leak rate. Because the potential for offsite release is likely to be much lower in these MODES the potential consequences of primary to secondary rer.etor coolant leakage are far lower. Y ACTIONS 8.1 With primary to secondary leakage in excess of I gpm the Required Action is to reduce the leak rate below 1 gpm in 4 hours. The Completion Time is based on engineering judge-ment. Four hours provides a stable period at operating l pressure to verify the leak quantity, determine which generator is nffected and prepare plans for remedial action. c Activities to assess the leak rate should be done while the plant is in stable steady state conditions and cannot be done-precisely if the plant is maneuvered. B.1 and B.2 l If the leak cannot be reduced, reducing power and pressure in l sequence to MODE 3 and then MODE 5 are required. This I reduces the leakage, and reduces or eliminates the mechanisms for offsite release. In this condition, the plant must be placed in a MODE in which the LC0 does not apply. This is done by placing the plant in at least MODE 3 in 6 hours and in MODE 5 in 36 hours. The six hours allotted to reach MODE 3'is a reasonable time based on operating experience to reacn MODE 3 from full power without challenging safety systems and ) l operators. Similarly, the 36 hours allotted is a reasonable, time to reach MODE 5 considering the plant can easily cooldown in such a time frame on one safety system train. The orderly cooldown also permits tube stresses to be held to i a minimum, limiting the potential for further tube degrada, tion. j (continued) j l Crystal River Unit 3 8 3.3-57 Revision l

p i RCS Primary to Secondary Leakage B 3.3.12 l-? BASES (continued) l L l i rl SURVEILLANCE SR 3.3.12.1 REQUIREMENTS Detection and location of primary to secondary leakage may be t A, done by several methods. The inventory balance is the most i precise method for quantifying leakage and may be done in o conjunction with other methods which may also include analyzing the secondary chemistry and activity. Seventy two hours permits a reasonable interval for trending, and is f consistent with the time period for assessing leakage at other locations. Stable steady state conditions must exist to obtain an [ accurate-indication of the leakage rate using the inventory t conditions. Other methods also require steady state. For. this surveillance steady state means stable' RCS pressure, L temperature, steady reactor. power with equilibrium xenon, t constant boron concentration in the reactor coolant and stable secondary side conditions (temperature, pressure, flow and chemical consistency). Stable and constant makeup, letdown RC pump seal injection and seal return flows and stable makeup tank and pressurizer levels are also required. I 4 The NOTE provides an exception to 3.0.4 permitting entry into MODES with stable conditions so the surveillance can produce accurate results. L REFERENCES 1. 52 FR 3788, NRC Interim Policy Statement on Technical Specification improvements for Nuclear Power Reac-tors, February 3, 1987. ~ l i L L l l J l. Crystal River Unit 3 8 3.3-58 Revision l w ...n.

(( H RCS Pre u ure Isolation Valve (PIV) Leakage B 3.3.13 h B 3.3. REACTOR COOLANT SYSTEM (RCS) o e B 3.3.13 RCS Prtssure Isolation Valve (PIV) Leakaos p4 I BASES BACKGROUND This specification applies to the four series check valves (two per line) that isolate the RCS from low pressure portions of the decay heat removal (OHR) system outside the containment. A normally closed, high pressure rated, motor operated g3te valve is upstream of the two check valves. The selection of valves is based on information presented in Reference I which requires testing of two in-series check valves used for isolation of high pressure to low pressure systems when leakage of one valve could go undetected for a substantial length of time, j I' The surpose of the RCS PIV LCO is to permit system operation l in tie presence of leakage through these valves in amounts which do not compromise safety. PIV leakage limits (leakage s 5 gpm) apply to leakage rates for individual valves. The total IDENTIFIED LEAKAGE rate of 10 gpm given by LC0 3.3.11 1 also applies and leakage through these valves are to be -l included in the total allowable leakage, i Although the specification provides limits in the form of allowable leakage rates, the important purpose of the i specification is to prevent overpressure failure of the low pressure portions of the DHR system caused by high RCS pressure. The leakage limits are indications that the boundary (check valves) between the RCS and the DHR system is degraded or becoming degraded. Failure of the check valves could lead to overpressure of the DHR piping or ccmponents. Failure consequences could be a LOCA outside of containment, with the possibility of being unable to recirculate from the containment emergency sump after the initial BWST injection is exhausted. A failure of portions of the DHR system can also degrade the ability for low pressure ECCS injection, l although analyses indicate that other remaining injection L paths would successfully maintain core cooling (when injec-tion water is available). (continued) l L l l

g,

Crystal River Unit 3 B 3.3-59 Revision l l

i RCS Pressure Isolation Valve (PlV) Leakage B 3.3.13 r BASES (continued) L APPLICABLE Pressure iso 1a' tion valve leakage is not considered in any SAFETY ANALYSIS design basis accident analyses. This specification provides I for monitoring the condition of the reactor coolant pressure boundary to detect degradation which could lead to accidents. Therefore, Selection Criterion 1 of the NRC Interim Policy Statement (Ref. 2) is satisfied. i !.COs Maximum isolation valve leakage is usually on the order of l drops per minute. The source of the leakage limits is an NRC i-letter resulting from the Reactor Safety Study (WASH 1400) which identified the inter-system loss of coolant accident as a significant contributor to core melt risk (Ref.1). Viola-i tion of this LC0 could result in continued degradation of a pressure isolation valve. APPLICABILITY This LC0 apply to MODES 1, 2, 3, and 4 to ensure that pres-sure isolation valve leakage is detected to minimize pressure boundary degradation from continued leakage. Leakage limits are not provided for MODES 5 and 6 because the reactor coolant pressure is far lower, making leakage less likely. Accordingly, the potential for the consequences of reactor 1 coolant leakage are far lower during these MODES. 1' I ACTIONS 6.d l With leakage in excess of the allowable limits 4 hours are L provided to reduce leakage. The period permits operation l to continue under stable conditions while leakage is assessed and corrective actions are being taken. Leakage assessment requires a stable plant pressure condition. The 4 hour time I is based on engineering judgement that actions to reseat and verify the leakage quantity in-this period can be reasonably performed. The 4 hour period is also based on a subjective judgement that operation for longer periods increases the hazard for overpressurizing the low pressure portions of the DHR outside containment. (continued) R Crystal River Unit 3 8 3.3-60 Revision l

m f;.

ll RCS Pressure Isolation Valve (PIV) ' Leakage U

hf B 3.3.13 i BASES. ACTIONS. B.1 and B.2 (continued) The reactor must be placed in a MODE in which the requirement does not apply if lee.kage cannot be reduced. This is done by-placing-the )lant in MODE 3 within 6 hours and MODE 5 within the next 36 Tours. This action reduces the leakage and also reduces the factors which tend to further degrade the isolation valves. The six hours allotted to reach MODE 3 is a reasonable time based on operating experience to reach MODE 3 from full power without challenging safety systems and operators. Similarly, the 36 hours allottej is reasonable to [ reach MODE 5 considering that a plant can easily couldown in such a time frame on one safety system train. SURVEILLANCE SR 3.3.13.1 REQUIREMENTS Performance of leakage testing on each reactor coolant pressure isolation valve is required to verify that leakage is below the specified limits and to detect leaking valves. t The 5 gpm limit is to be applied to each valve. Testing is performed at least every refueling or prior to startup after a 72 hour outage if a recent test (within 9 months) does not exist. The 72 hour outage allowance is based on engineering judgement. These Surveillance Requirements were specified by the NRC (Items 1 and 2) in Ref. I and are in accordance with ASME XI (Item 3).(Ref. 3). SR 3.0.4 is exempted to permit leak testing at high valve differential pressures with stable conditions than are possible in the lower MODES. ,h i l' l REFERENCES 1. NRC Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves (all plants) dated 4/20/81. Includes Technical Evaluation Report " Primary Coolant System Pressure Isolation Valves," prepared by the Franklin Research Center. 2. 52 FR 3788, NRC Interim Policy Statement on Technical l, Specification Improvements for Nuclear Power Reactors, February 6, 1987. 3. ASME Boiler and Pressure Vessel Code, Section XI, L Subsection IWV, " Inservice Testing of Valves in Nuclear l. Power Plants." l ~ Crystal River Unit 3 B 3.3-61 Revision l

. =.. i Leakage Detection Instrumentation 3.3.14 B'3.3 REACTOR COOLANT SYSTEM (RCS) B 3.3.14 Lgakaae DetSction Instrumentation BASES i j BACKGROUND 3eneral Design Criterion 30, " Quality of Reactor Coolant Piessure Boundary," of Appendix A to 10 CFR 50, " General Design Criteria for Nuclear Power Plants, (Ref.1)" rcquires that means be provided for detecting and, to the extent practical, identifying the' location of the source of reactor coolant leakage. A limited amount of leakage is expected from the reactor coolant system (RCS) and from auxiliary systems within the containment. Some leakage will occur from valve packing, pump seals, vessel / closure head seals and safety and relief valves. If leakage occurs via these paths it is detected, collected to the extent practical, and isolated from the containment atmosphere so as not to mask any potentially L serious leak should it occur. These leakages are IDENTIFIED LEAKAGE and may be piped to tanks or sumps so flow rate can be established and monitored during plant operation. Uncollected leakage to the containment atmosphere from other sources increases the humidity of the containment. The moisture condensed from the atmosphere by air coolers together with any associated liquid leakage to the contain-ment is UNIDENTIFIED LEAKAGE and is collected in tanks or sumps where the flow rate is established and monitored during alant operation. A small amount of UNIDENTIFIED LEAKAGE may t )e impractical to eliminate, but it should be reduced to a small flow rate, to permit the leakage detection systems to positively and rapidly detect a small increase in flow rate. Thus a small UNIDENTIFIED LEAKAGE rate that is of concern will not be masked by a larger acceptable IDENTIFIED LEAKAGE rate. Leakage detection systems should detect significant reactor coolant pressure boundary (RCPB) degradation as soon after occurrence as practical to minimize the potential for a gross pressure boundary failure. Some cracks might develop and penetrate the RCPB wall, exhibit very slow growth, and afford ample time for a safe and orderly plant shutdown. (continued) l I L Crystal River Unit 3 B 3.3-62 Revision l l.

x t Leakage 03tection Instrumentation 3.3.14 BASES l' BACKGROUND The leakage detection monitors used are of two different ( (continued) principles: containment sump level and atmospheric activity monitor. The atmospheric activity monitoring instrumentation detects gaseous and iodine radioactivity. Industry practice has shown that water flow rate changes of from 0.5 to 1.0 gpm can readily be detected in containment sumns by monitoring changes in sump water level, in flow i rate, or in the operating frequency of pumps. Sumps and tanks used to collect UNIDENTIFIED LEAKAGE and air cooler condensato are instrumented to alarm for increases of the normal flow rates. This sensitivity provides an acceptable performance for detecting increases in UNIDENTIFIED LEAKAGE. Reactor coolant activity released to the containment can be i detected by radiation monitoring instrumentation. Instrument sensitivities of 10~8 micro Ci/cc particulate monitoring and of 10', radioactivity for air micro Ci/cc radioactivity for gaseous monitoring are practi::a1 for thepe leakage I detection systems. Typical ranges are 10-10 6 cpm. Radio-activity monitoring systems are included because of their sensitivity and rapid response to leaks from the RCPB. In addition to the instrumentation cited by the LCO, other leakage detection means may be used. Humidity changes or I pressure and temperature changes may provide indications of leakage. I l APPLICABLE The safety significance of leaks from the RCPB vary widely SAFETY ANALYSIS depending on the source of the leak as well as the leakage r rate and duration. Therefore, the detection and monitoring of reactor coolant leakage into the containment area is F necessary. Separating the identified sources of leakage from L unidentified sources is necessary to provide prompt and y quantitative information to the operators to aermit them to J take immediate corrective action should a lea ( occur that is detrimental to the safety of the facility. RCS leakage detection instrumentation satisfies Selection l' Criterion 1 of the NRC Interim Policy Statement (Ref. 2). As such, these variables are retained in the RCS Leakage

u Detection Instrumentation LCO.

(continued) i t 4 3 Crystal River Unit 3 B 3.3-63 Revision l i

p{ i y L Vd Leakage Detection Instrumentation 3.3.14 1 BASES (continued) c m a LCO One method of protection against RCPB leakage failure is the J, ability of instrumentation to rapidly detect extremely small j leaks. This LC0 requires that instrumentation of two diverse j principles be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition when leakage indicates possible RCPB degradation. The LC0 is satisfied when monitors of diverse measurement means are available. Thus the contair. ment sump monitor in combination with the containment atmosphere radioactivity monitor (iodine and gaseous channels) provides an acceptable minimum. APPLICABILITY Leakage detection systems are only required to be OPERABLE in E, MODES 1, 2, 3, and 4. In these MODES the RCS temperature is greater than 200*F and pressure is greater than atmos)heric. With the plant in MODES 5 or 6, temperature is less tian or equal to 200'F'and pressure is low. Below 200'F any leakage would be liquid and atmospheric monitors are less effective. Since the design of the RCPB is eble to withstand tempera-tures and pressures far greater than those allowed in MODES 5 or 6, and the pressure differential across the RCPB is low, leakage is almost impossible. Therefore, the LC0 is not applicable in MODES 5 and 6. ACTIONS A.1 and A.2 With the containment atmosphere radioactivity. monitor (gaseous and iodine activity channels) inoperable, grab samples shall be taken and analyzed to provide alternate c' periodic information. Provided these samples are obtained ^ and analyzed every 24 hours, the plant may continue operation for up to 30 days. The 24 hour sampling interval is based on engineering judgement and plant operating experience. The 30 day Completion Time for restoration is based on engineering y judgement and recognizes that multiple forms of leak detection are available, but extended periods of operation while using alternative monitoring is not prudent since the original design monitoring capability is not being met. i (continued) L ,m Il l l Crystal River Unit 3 B 3.3-64 Revision l

v n ' E Lcakage Detactice Instrumentation oi 3.3.14 I BASES d h -ACTIONS B.1 and B.2 (continued). With the containment sump level monitor inoperable, no form of grab sample could provide the equivalent information. However the atmos)heric activity monitors provide indications of changes in lea (age. Restoration is required to regain the function of the sump monitor. As an alternate to the sump 1 monitor and in conjunction with atmospheric monitors the periodic surveillance, SR 3.3.11.1, for RCS inventory balance is to be acrformed at an increased frequency of once per 24 hours. T1e 24 hours is based on engineering judgement and is compatible with the required Frequency for grab sam)les. The ( 30 day Completion Time for restoration recognizes t1at i multiple forms of leakage detection are available. The Completion' Time is based on engineering judgement and plant . operating experience. C.1 and C.2 i If the Required Action cannot be met within the required i e Completion Time, the plant must be placed in a MODE in which the LC0 does not apply. This is done by placing the plant in i at least MODE 3 in six hours and in MODE 5 in 36 hours. The six hours allotted to reach MODE 3 is a reasonable time based-on operating experience to reach MODE 3 from full power i U without challenging safety systems and operators. Similarly, the 36 hours allotted is a reasonable time to reach MODE 5 considering that a plant can easily cooldown in such a time frame on one safety system train. 1 SURVEILLANCE SR 3.3.14.1 REQUIREMENTS This Surveillance requires the periodic monitoring of containment atmosphere activity. This provides assurance that leakage which would indicate reactor coolant pressure boundary degradation would be detected. The surveillance Frequency (12 hours) is based on engineering judgement and industry-accepted practice, j (continued) t ~ s k i Crystal River Unit 3 B 3.3-65 Revision l = ~ - - .. m

q. lN: Leakage Det'ection Instrumentation 3.3.14 BASES-SURVEILLANCE SR 3.3.14.2 REQUIREMENTS l (continued) This Surveillance requires the periodic monitoring of the containment : ump level. This provides assurance that leakage i which would indicate reactor coolant pressure boundary i degradation would be detected. The surveillance Frequency (12 hours) is based on engineering judgement and industry-accepted practice. l SR 3.3.14.3 ) Surveillance Requirement 3.3.14.3 is the performance of a l CHANNEL CHECK of the containment atmosphere (gaseous'and . iodine) activity monitor The CHANNEL CHECK gives reasonable confidence that th; channels are within specification with . respect to their alarm setpoints. The surveillance Frequency (12 hours), is based on engineering judgement and industry-accepted practice. l. SR 3.3.14.4 Surveillance Requirement 3.3.14.4 is a performance of a j CHANNEL FUNCTIONAL TEST for the containment atmosphere l (gaseous and iodine) activity monitor. This test ensures l' that the monitor can perform its function in the desired manner. The CHANNE: FUNCTIONAL TEST verifies the alarm setpoint and relative accuracy of the instrument strings. The Frequency is based on engineering judgement and industry- ) accepted practice. l SR 3.3.14.5 and SR 3.3.14.6 Surveillance Requirements 3.3.14.5 and 3.3.14.6 are the performance of CHANNEL CALIBRATIONS of the containment i atmosphere activity monitor and containment sump level monitor every 18 months. The CHANNEL CALIBRATION verifies L I the accuracy of the instrument string. The calibration includes the calibration of instruments located inside containment. The Frequency is based on engineering judgement and industry-accepted practice. p [ REFERENCES 1. Title 10 Code of Federal Regulations, Part 50, Appendix L, A, General Design Criteria for Nuclear Power Plants. L 2. 52 FR 3788, NRC Interim Policy Statement on Technical Improvements for Nuclear Power Reactors, February 6, 1987. Crystal River Unit 3 B 3.3-66 Revision l

l, Specific Activity'. B 3.3.15 3.3 REACTOR COOLANT SYSTEM (RCS) B 3.3.15 Specific Activity BASES ' BACKGROUND The purpose of the reactor coolant system (RCS) Specific Activity LCO.is to limit the concentration of radionuclides in the reactor coolant and the resultant offsite dose consequences in the event of a steam generator tube rupture (SGTR). The sequence of this event includes a brief release of steam to the atmosphere through the safety and atmospheric dump valves followed by cooldown and depressurization using tne turbine condenser. This LCO contains both iodine and total specific activity limits. The iodine isotopic activities are expressed in terms of a DOSE EQUIVALENT l-131 per gram of reactor coolant. Total specific reactor coolant activity is limited on the basis.of the weighted average beta and gamma energy levels in the coolant. The allowable levels are intended to limit the 2-hour dose at the site boundary to a small fraction of the 10 CFR 100 limit. APPLICABLE The limitation on the specific activity of the primary SAFETY ANALYSIS coolant ensures that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of the 10 CFR 100 limit following an SGTR with an assumed pre-existing steady state primary to secondary steam generator leakage rate of 1.0 gpm to the unaffected steam generator. The pre-existing leak has the effect of contaminating the secondary side prior to the SGTR. Operation with iodine specific activity levels greater than the LC0 value is al-lowed provided the isotopic concentration does not exceed the limit in Figure 3.3.15-1. The 1.0 microcurie / gram limit may be exceeded temporarily (48 hours) during non-steady state conditions when power or reactor coolant pressure changes cause iodine spiking. Operation outside this limit for a restricted time period provides time to reduce the temporari-O ly increased iodine concentration and DOSE EQUIVALENT l-131 to within its limit. The activity levels allowed by Figure 3.3.15-1 increase the dose at the site boundary by a factor of up to twenty following a postulated SGTR. Use of Figure 3.3.15-1 is believed to be acceptable because the probability of an SGTR occurring during this short time interval is low enough to justify a higher dose limit. (continued) Cryst 1 River Unit 3 8 3.3-67 Revision l e m

m a .I f L Specific Activity i~ B 3.3.15 V BASES' j .j

APPLICABLE LCO 3.3.15, Specific Activity, satisfies the requirements of' SAFETY ANALYSIS Selection Criterion 2 of the NRC Interim Policy Statement' (continued)

(Ref.1) because it helps ensure that reactor coolant activity will be within the initial conditions assumed in the-accident analysis. LCOs The specific iodine activity is limited to 1.0 microcurie per gram DOSE EQUIVALENT I-131 and the total specific activity in the primary coolant is limited to_the number of microcuries per gram equal to 100 divided by E (average disintegration ,2 energy of the sum of the average beta and gamma energies of the coolant nuclides). These values represent a reasonable operating capability rather than a specific analytical result.. Adherence to these limits is' required to restrict the 2-hour offsite dose following a SGTR to a small fraction of 10CFR100 limit. Violation of this LCO may result in coolant activity levels which exceed the generically a)- plicable dose objective for the SGTR. The DOSE EQUIVA.ENT'l-131 may exceed 1.0 microcurie per gram for up to 48 hours during one continuous time interval, provided that the concentration does not exceed the limit shown on Figure y N 3.3.15-1. This accommodates iodine spiking. t APPLICABILITY LC0 3.3.15, Specific Activity, is applicable in MODES 1 and i 2, and in MODE 3 with T,y 1 500*F. If a SGTR were to occur, in conjunction with a reactor trip, the energy in the primary system is sufficient to cause the secondary safety valves and l the atmospheric dump valves to open in MODE 1, and possibly in MODE 2.~ For conservatism, the applicability has been 1 500'F. At this extended to include MODE 3 with T*7eleases are not possible temperature secondary atmospheric l-because the saturation pressure at 500*F is considerably below the opening setpoint. Thus the applicability is a conservative range that bounds the possibility for site boundary doses for SGTR. The LC0 is not applicable in MODE 3 s 500*F or MODES 4, 5, or 6 because releases cannot with T@ rough the steam supply system flow paths. occur R (continued) l L l L 1 L Crystal River Unit 3 8 3.3-68 Revision l

r q y Specific Activity B 3.3.15 BAS.ES (continued). 1 ACTIONS A.I. A.2. B.I. B.2. C.l. C.2. 0.1. and D.2 7 The Required Actions taken in response to high reactor coolant activity include sampling and analysis for determina-tion of the DOSE EQUIVALENT I-131. Samples are analyzed at least every 4 hours for I-131, I-133, and 1-135 if the limits of LC0 3.3.15 is exceeded. If the limits on Figure 3.3.15-1 are exceeded, or if the LC0 3.3.15 limit on iodine activity is exceeded for more than 48 hours during one continuous time interval, the reactor is placed in MODE 3 (below 500*F) within six hours. The Completion Time of 48 hours to reduce the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with processing systems and is based on engineering judgement. L Similarly, if reactor coolant specific activity exceeds the 100/E limit, the reactor is placed in MODE 3 (below 500*F) within 6 hours. This Required Action reduces the possibility of a release of activity through the main steam safety or at-l mospheric vent valves because 500*F is below the saturation i i temperature required to open these valves. Six hours is a reasonable maximum time based on operating 4 experience to reach MODE 3 from full power without challeng-ing safety systems and operators, and 4 hours is a reasonable L time between reactor coolant sample analyses. As such, the l Completion Times are based on engineering judgement. l l ~ L' SURVEILLANCE SR 3.3.15.1 ' REQUIREMENTS The purpose of this periodic surveillance is to assure the coolant activity remains within the allowable limits.

The l

surveillance is performed by drawing samples of coolant and l performing a radiochemical analysis. By maintaining activity within limits, site boundary doses from a SGTR would be !~ expected to be a sma'il fraction of the 10CFR100 limit. The l' 72 hour Frequency is based on engineering judgement. Performing the surveillance routinely while steady state h conditions exist (constant power and RCS temperature) allows trends of average coolant activity to be evaluated and inferences about fuel condition to be made. n (continued) i Crystal River Unit 3 8 3.3-69 Revision l p

'(', J l Specific Activity B 3.3.15 BASES SURVEILLANCE SR 3.3 1122 - REQUIREMENTS (continued) The purpose of this surveillance is also to assure the L coolant activity remains within limits to ensure that SGTR site boundary doses are a small fraction of the 10CFR100 limit. The surveillance provides more details of the isotopic content than the iodine surveillance (SR 3.3.15.1) / but is performed at a less frequent interval (184 days). The Frequency is based on engineering judgement and industry-accepted practice. The Frequency NOTE has the effect of requiring a stable coolant condition free from any power change effects that would cause activity spikes or anomalies that would not represent the usual condition of the coolant. l l REFERENCES 1. 52 FR 3788, NRC Interim Policy Statement on Technical l Specification Improvements for Nuclear Power Reactors, i l-February 6, 1987. I i + f i -l L l L i I l 1 I Crystal River Unit 3 8 3.3-70 Revision l L

( b ECCS Trains - B 3.4.2 LCOs 3.4.2 and 3.4.3 j BASES (continued) i. c LCOs' These LCOs establish' the minimum equipment required to be available to accomplish the core cooling safety function i L following accidents which render the. steam generators effectively unavailable, such as a large LOCA. -Two indepen-dent (and redundant) ECCS trains are required to.be OPERABLE to ensure that at least one is available assuming a single failure in the other train. Additionally, individual 4 components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents. l During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the BWST to-the RCS via the HPI and LPI pumps and their respective discharge flow paths to each of tk four cold leg injection nozzles and the reactor vessel.- la the long term, this flow path may be manually transferred to take its supply from the containment sump. The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains. In MODE 4, one OPERABLE ECCS' train is acceptable without single failure consideration on the bases of the stable reactivity condition of the reactor and the limited core cooling requirements. In this condition one HPI and one LPI train provide sufficient. injection flow to meet ECCS requirements. [ 4 In order to preclude a low temperature ovepressurization I event in MODE 3 with RCS temperature s 283*F and in MODE 4, j the high pressure injection isolation valves may be closed 1 with their power supply breakers locked in the open position. Two HPI pumps may also be deactivated when low temperature overpressurization is a concern. Regardless of the method used to deactivate HPI, operator action is then required to 1 initiate high pressure injection. This is also considered acceptable on the bases of the stable reactivity condition of i the reactor and the-limited core cooling requirements. The requirements of these LCOs are derived principally from events involving a loss of coolant inventory, and par-I ticularly the Appendix K (Ref. 5) evaluation. Failure to meet these requirements could result in the inability to { h match core heat generation, leading to fuel melting, in-creased clad metal-water reaction, and potential for altera-L tion of core geometry for such low probability, high conse-quence events. 1 (continued) i l Crystal River Unit 3 B 3.4-12 Revision 3 l$'

pw / h ECCS Trains - B'3.4.2 LCOs 3.4.2 and 3.4.3 BASES SURVEILLANCE SR 3.4.'2.6 (Cont'd) REQUIREMENTS-This surveillance has two notes associated with it. Note 1 requires the HPI flow balance surveillance be performed in MODE 3. This surveillance energizes the high pressure injection (HPI) pumps to inject water into the reactor c. vessel. In order to minimize the possibility of overpres-surizing the reactor vessel at low temperatures as a result of HPI operation, this surveillance is performed at plant conditions in which low temperature overpressurization is not a Concern. MODE 6 is not utilized for the testing due to the potential i for overfilling the fuel transfer canal and increasing the reactor building's airborne contamination. As a result of Note 1, Note 2 makes the provision of 3.0.4 not applicable i for the HPI flow balance surveillance to allow it to be I performed in MODE 3 after reaching the applicable MODE. SR 3.4.2.7 This. surveillance ensures that the ECCS flow path from the BWST to the RCS is properly aligned by requiring a verifica-tion of the line up of those valves which could be inadver-i tently repositioned. Failure of one of these valves will affect only one ECCS train. Therefore, a monthly frequency 3 has been established based on engineering judgement, and has i been shown to be acceptable through operating experience. SR 3.4.2.8' This surveillance ensures that the automatic isolation and interlock function' of the DH system from the RCS'will function if challenged. This interlock will prevent exces-sive RCS pressure from being exerted on the DH/LPI system through an open DH suction line from the RCS hot leg. The interlock will prevent opening and will automatically close l isolation valves in the DH line when RCS pressure is above l-the setpoint. Excessive pressures in the DH system poten-p tially could result in a loss of coolant accident outside of the containment. The interlock setpoint is based on prevent-ing excessive pressure from being exerted on the DH/LPI system from the RCS. A surveillance frequency of 18 months has been established based on engineering judgement. This i frequency has been shown to be acceptable through operating i. experience. t l (continued) l. L Crystal River Unit 3 B 3.4-17 Revision n

7: y 1 ~ -4 j. e 1 m I i g., t 4' l ) e TSCRN 174B + i i f i m

i l r i FLORIDA POWER 00RPORATION CRYSTAL RIVER UNIT 3 [ DOCKET No. 50-302/ LICENSE No. DPR-72 l I REQUEST No. 174, REVISION 0 PRESSURE / TEMPERATURE LIMITS 3. LICENSE DOCUMENT INVOLVED Technical specifications R PORTIONS 3.4.9.1 Figure 3.4-2 l Figure 3.4-3 Figure 3.4-4 DESCRIPTION OF REQUEST: This submittal requests that reactor coolant system pressure and temperature limits be revised to allow for reactor operation during 'the first 15 effective full power year (EFPY) service period. The [ maximum' heatup and cooldown rates assumed in the analyses and implemented in Technical Specifications (TS) have been revised. References to the 10CFR50 Appendix G criticality limit curve have been deleted. from. the TS and Bases. The Bases have also been revised to reflect the most recent surveillance capsule results. RT NDT values at the 1T/4 and 3T/4 flaw locations for 21 EFPY and the latest information on chemical composition of the limiting reactor coolant system (RCS) components were added. i REA8ON FOR REQUEST Title 10 of the Code of Federal Regulations, Part 50, Appendix G requires reactor vessel pressure / temperature limits be established l in' order to ensure fracture toughness requirements are satisfied for the current operational service period. Current Crystal River Unit 3'(CR-3) pressure / temperature limit curves are conservative a for the first 8 EFPY. CR-3 is approaching the end of this service I period and is submitting revised pressure / temperature limits to allow for continued operation. + The maximum allowable heatup and cooldown rates assumed in the fracture mechanics analyses have been revised to more accurately L reflect actual plant capabilities. Previously assumed values for 'heatup and cooldown rate were not representative of actual plant ? operating conditions. Administrative changes were required to make the TS and bases i L consistent with the current CR-3 licensing basis. References to I the 10CFR50 Appendix G Criticality Limit curve have been deleted. While this curve is still calculated (as required by 10CFR50 Appendix G) the requirement to include it in TS was deleted by CR-3 License Amendment No. 82, dated September 23, 1985. The Bases are revised to include the most recent reactor vessel surveillance capsule data and reflect the use of Regulatory Guide 1.99 Revision Number 2 in the preparation of the 15 EFPY CR-3 pressure / temperature limit curves.

t i EVALUATION OF REQUEST The Heatup, Cooldown, and Inservice Leak and Hydrostatic Testing pressure / temperature limit curves provide a substantial margin between fracture toughness limits and actual plant operating conditions. This ensures that the reactor coolant pressure boundary (RCPB) is protected against non-ductile failures due to anticipated mass or energy inputs to the RCS. After the first several

EFPY, the reactor vessel becomes the most limiting component of the RCPB in terms of fracture toughness.

Fast-neutron irradiation causes a decrease in the ability of the vessel to absorb stresses and resist fracture. The concern is a crack that is undetected during inservice inspection will propagate and i result in a non-ductile failure of the reactor vessel. This failure of the RCPB would have a serious impact on safety, particularly at rated RCS pressure and temperature. Therefore it is imperative that fracture mechanics limits be provided that are conservative - r the period of operation they are in use. This request submits updated pressure / temperature limit curves for the first 15 EFPY. The 15 EFpY curves (reported in BAW-2091) are calculated for the predicted fluence the vessel will be exposed to at the end of 15 EFPY and are conservative until that time. The . fluence values are based on the latest surveillance capsule results for CR-3 and are published in BAW-2049 " Analysis of Capsule CRS-F l Florida Power Corporation Crystal River Unit-3" September 1988. The pressure / temperature limit curves are prepared in accordance with 10CFR50 Appendix G and an NRC-approved methodology documented in BAW-10046A, Rev. 2 " Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, Appendix G", June 1986. Regulatory Guide 1.99 Revision 2 " Radiation Embrittlement of Reactor Vossel Materials" May 1988 was used to predict the shift in reference temperature (RT NDT) as a function L of fluence and vessel chemistry. This methodology was endorsed by the NRC Staff in Generic Letter 88-11 as an acceptable method for l predicting the effect of neutron irradiation on reactor vessel I materials. The change in methodology necessitates a revision to l the bases to include information pertinent to the new methods and [. remove. superseded data based on previous revisions of Regulatory H Guide 1.99. t l' The maximum allowable heatup and cooldown rates are consistent with the assumptions used in the fracture mechanics analysis. Changes in RCS temperature result in thermal stresses to the vessel. The larger the allowable rete of change of temperature, the larger the l magnitude of the resultant thermal stress. This request reduces l the allowable heatup and cooldown rates to make the rates more consistent with actual plant capabilities. The lower rates result in lower thermal stresses to the vessel. These limits ensure that the operator will not heatup or cooldown the RCS at an unanalyzed rate. I -n-'

I SHOLLY EVALUATION OF REQUEST Florida Power Corporation (FPC) proposes the revision to the pressure / temperature limit curves and the reduction in the maximum i allowable heatup and cooldewn rates does not involve a significant hazard consideration. The revision of the pressure / temperature l limit curves is necessary to ensure conservative protection of the RCPB for the first 15 EFPY of reactor operation. Furthermore, the limit curves were prepared in accordance with an NRC-approved methodology. Technical Specifications will continue to require the plant to be operated within the limits of the curves and allowable heatup and cooldown limitations and include appropriate actions to be taken in the event the limits are violated. FPC concludes this change will not: 1. Involve a significant increase in the probability or consequence of an accident previously evaluated because the revision of the pressure / temperature limit curves and the reduction in allowable heatup and cooldown rates has no influence or impact on the probability of a Design Basis Accident (DBA) occurrence. CR-3 Technical Specifications will continue to require operation.of the RCS within the pressure / temperature and heatup/ cooldown limits. The reduction in the magnitude of the allowable heatup and cooldown rates results in generally lower thermal stresses applied to the reactor vessel than are currently allowed. The methodology used in preparation of the curves has been reviewed and approved by the NRC and is l unchanged from previous pressure / temperature limit revisions (With the exception of the current usage of Regulatory Guide t l.99 Revision 2). Regulatory Guide 1.99 Revision 2 has been endorsed by the NRC Staff as providing improved empirical correlations for predicting the effects of irradiation on the properties of reactor vessel materials. 2. Create the possibility of a new or different kind of accident from any accident previously evaluated because updating the pressure / ter.perature limit curves and reducing the allowable RCS heatup and cooldown rates introduces no new mode of plant l l operation nor does it require a physical modification to the l plant. 1 3. Involve a significant reduction in the margin of safety because the CR-3 15 EFPY pressure / temperature limits are I prepared using the same NRC-approved methodology used to l generate the current 8 EFPY pressure / temperature limit curves. The decrease in reactor vessel fracture toughness 1 (CONTINUED) l l L

i -( r SEOLLY EVALUATION OF REQUEST (CONTINUED) (due to accumulated neutron exposure) has been accounted for in the preparation of the 15 EFPY pressure /. temperature limit curves using the methodology of NRC Regulatory Guide 1.99 Revision 2. This has resulted in lower allowable RCS pressures for given RCS temperatures in order to maintain acceptable levels of stress within the vessel and ensure the i margin of safety is not decreased. e l' l I 1 1 i. l Y 1 r 1 l _}}

3F1089-23, Application for Amend to License DPR-72,consisting of Tech Spec Change Request 174 Re Proposed Heatup,Cooldown & Inservice Leak & Hydrostatic Testing Pressure/Temp Limits. Rept Re Low Temp Overpressure Protection Withheld

Text

Subject:

3.6 DESCRIPTION

Navigation menu

Search