Text

Proposed Change to Emergency Plan Table 3-1: Classification of Emergencies
	ML042170294
Person / Time
Site:	Saint Lucie
Issue date:	08/02/2004
From:	Jefferson W; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2004-178
	Download: ML042170294 (42)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach. FL 34957 August 2, 2004 FPL L-2004-178 10 CFR 50.54(q)

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE: St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Proposed Change to Emergency Plan Table 3-1: Classification of Emergencies Pursuant to 10 CFR 50.54(q), Florida Power & Light Company (FPL) requests approval prior to implementation, of a change to the St. Lucie Plant Radiological Emergency Plan.

FPL requests NRC approval for changing the Initiating Condition (IC)for an (Notification of) Unusual Event (UE) due to reactor coolant system (RCS) leakage to resolve an unintended consequence of the 1996 IC change. FPL seeks NRC approval to: (1) reestablish the link to the RCS Technical Specification applicability; and (2) add the condition "ability to isolate" when determining the occurrence of RCS leakage. The current IC and the proposed IC remain consistent with the NUREG 0654 scheme of classification. The ICs for Alert and above remain unchanged by this request.

The proposed change has been reviewed by the Facility Review Group (FRG) on July 29, 2004. On receipt of approval from the NRC of the proposed change, FPL will revise the St. Lucie Plant Emergency Plan and Emergency Plan Implementing Procedures to implement the revised UE declaration threshold. FPL will obtain the concurrence of state and local governments prior to implementing this change.

FPL discussed the attached proposed IC change and the schedule with the NRC Project Manager NRC Staff on July 23, 2004. FPL requests the NRC to complete the review by November 1, 2004, to support training for implementation prior to the fall 2004 Unit 2 refueling outage (SL2-15) which is currently scheduled to start in late November.

Please contact George Madden at 772-467-7155 if there are any questions about this submittal.

ili Vice Prsd-t St. Lucie Plant WJ/GRM Attachments (3) an FPL Group company

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 1 Page 1 Attachment I FPL requests NRC approval for changing the Initiating Condition (IC) for an (Notification of) Unusual Event due to Reactor Coolant System (RCS) leakage to resolve an unintended consequence of the 1996 IC change. FPL seeks NRC approval to: (1) reestablish the link to RCS Technical Specification applicability; and (2) add the condition "ability to isolate" when determining the occurrence of RCS leakage. The current IC and the proposed IC remain consistent with the NUREG 0654 scheme of classification. The contents of this submittal package are as follows:

A. Background B. Comparison of current IC/EAL to proposed IC/EAL C. Basis and justification for the change D. State/local government review/concurrence E. Supporting References A. BACKGROUND In 1996, St. Lucie Plant requested and received NRC approval for changes submitted for the Initiating Condition (IC)for an (Notification of) Unusual Event (TAC Nos. M96274 and M96275). That change allowed an Unusual Event (UE), due to reactor coolant system (RCS) leakage, to be determined solely on the basis of the quantity of the leak and not whether the leakage was identified or unidentified. As a consequence of that change, the RCS Emergency Action Levels (EAL) were no longer tied to the leak rates defined in the Unit 1 or Unit 2 Technical Specifications and therefore, not dependent on the mode relationship within those specifications. The change inappropriately broadened the applicability of the EALs to all modes defined in Technical Specifications.

The EALs used at St. Lucie Plant are based on NUREG 0654. The IC change approved in 1996 was an acceptable alternative to the NUREG 0654 IC previously in place. This proposed change seeks to revise the original acceptable alternative by re-establishing a link to the Technical Specifications and therefore mode dependence as originally defined in the NUREG 0654 RCS leakage IC.

B. COMPARISON OF CURRENT IC/EAL TO PROPOSED IC/EAL Refer to Attachment 2.

C. BASIS AND JUSTIFICATION FOR THE CHANGE St. Lucie currently uses the NUREG 0654 scheme of classification. The NRC safety evaluation dated October 17, 1996 (TAC Nos. M96274 and M96275) allowed the RCS leakage Unusual Event to be characterized as >10 GPM in an effort to avoid the time consuming evolution of doing a mass balance prior to determining if an emergency condition exists. When that EAL was changed, the reference to Technical Specification was removed, feeling that we were not using the Technical Specification leak rate as

V-St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 1 Page 2 defined in the specification. That deletion resulted in no longer bounding RCS leakage to conditions described in the specification, specifically it now includes all reactor operating modes, an approach inconsistent with NUREG-0654 guidance and more in keeping with the NEI/NUMARC scheme of emergency action levels (CU-1 for Mode 5).

St. Lucie has historically focused on the reactor coolant system as a liquid in lieu of the physical system, making it difficult to determine if a challenge to the RCS barrier had occurred. As such, it was necessary to establish a clear definition of the physical system called the reactor coolant system. To that end, a review was performed of the Updated Final Safety Analysis Report (UFSAR), NUREG-1432, The Standard Technical Specifications for Combustion Engineering Plants, the Code of Federal Regulations, NEI 99-01 Revision 4, Methodology for Developing Emergency Action Levels, and St.

Lucie's current Technical Specifications and their bases.

From UFSAR Section 5.1, the Reactor Coolant System (RCS) circulates water in a closed cycle, "to remove heat from the reactor core and transfers it to a secondary (steam generating) system. ...The major components of the system are the reactor vessel; two parallel heat transfer loops, each containing one steam generator and two reactor coolant pumps; a pressurizer connected to one of the reactor vessel outlet pipes; and associated piping. All components are located inside containment."

NUREG-1432, The Standard Technical Specifications for Combustion Engineering Plants (STS) defines the reactor coolant system as components that contain or transport the coolant to or from the reactor core and the Technical Specification covers Modes 1 through 4 specifically.

With regards to RCS leakage, the STS states:

"In Modes 1, 2, 3, and 4, the potential for reactor coolant pressure boundary leakage is greatest when the RCS is pressurized. In Modes 5 and 6, leakage limits are not required because the reactor coolant pressure is far lower, resulting in low stresses and reduced potentials for leakage."

The Code of Federal Regulations, Title 10 section 50.2, includes the following definition of Reactor Coolant Pressure Boundary:

"Reactor coolant pressure boundary means all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves, which are:

(1) Part of the reactor coolant system, or (2) Connected to the reactor coolant system, up to and including any and all of the following:

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 1 Page 3 (i) The outermost containment isolation valve in system piping which penetrates primary reactor containment, (ii) The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor containment, (iii) The reactor coolant system safety and relief valves."

In NEI 99-01 Revision 4, Methodology for Developing Emergency Action Levels, the definition for the reactor coolant system is as follows:

"The RCS Barrier includes the RCS primary side and its connections up to and including the Pressurizer safety and relief valves, and other connections up to and including the primary isolation valves. "

The St. Lucie Technical Specification for RCS leakage (T.S. 3.4.6.2) contains the following Limiting Condition for Operation (LCO) for Modes 1, 2, 3, and 4 (Unit 2 specific is in brackets):

Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through steam generators,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for each Reactor Coolant System Pressure Isolation Valve identified in Table 3.4.6-1.

[1 gpm leakage (except as noted in Table 3.4-1) at a Reactor Coolant System pressure of 2235 + 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.)

As a result of the above review, St. Lucie has incorporated the following definition for RCS for Emergency Plan use:

"RCS includes any component (pipe, vessel, valve, etc.) which is used to contain or transport the reactor coolant to or from the reactor core. This definition includes any component beyond the RCS pressure boundary, which remains open to the RCS."

According to the STS definition of RCS, systems connected to the reactor coolant pressure boundary (RCPB) should be considered as an extension of the RCS when not isolated from the RCS. During normal plant operations, the RCS extension would include the chemical and volume control system (CVCS) letdown and charging lines since they are open to the RCS. During other conditions, systems that may be included under the definition provided above include shutdown cooling (SDC)/low pressure safety injection (LPSI), high pressure safety injection (HPSI), containment spray (CS),

RCS sample, and some portions of the waste management system. Leakage from one

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 1 Page 4 of these systems, when it remains open to the RCS, would be considered reactor coolant leakage, since it would make it impossible to determine RCS barrier leakage, as such, it would allow an uncontrollable reduction of the RCS inventory being used for core cooling, thus potentially compromising plant safety.

The concept of isolating secondary systems from the physical RCS is evident in the initial NRC guidance provided to the industry in NUREG-0818, Emergency Action Levels for Light Water Reactors. In the category of RCS Technical Specification leakage for the Notification of Unusual Event EAL (page 24), the NRC found the draft EAL submitted by the V. C. Summer Plant acceptable for meeting the NUREG-0654 EAL (allowing for the timeframe provided in Technical Specification for returning the leak into conformance with the specification). This concept is also in place elsewhere in the industry.

FPL is in agreement that isolating interfacing systems is a primary means of determining if there is a true challenge to the reactor coolant system barrier.

Furthermore, allowing plant operations a reasonable amount of time to isolate those interfacing systems is prudent. The initial steps operators take in response to excess RCS leakage would be considered an appropriate timeframe.

The proposed change involves declaration of an Unusual Event due to RCS leakage.

According to the class description, Notification of Unusual Event (NUREG-0654) is indicative of "unusual events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs." The NRC has approved the definition of RCS in NEI 99-01 Revision 4 which states "The RCS Barrier includes the RCS primary side and its connections up to and including the Pressurizer safety and relief valves, and other connections up to and including the primary isolation valves." Additionally, The NRC has stated in Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, that "the safety significance of leaks from the reactor coolant pressure boundary (RCPB) can vary widely depending on the source of the leak as well as the leakage rate and duration."

The two factors of concern regarding RCS leakage, as stipulated in the Regulatory Guide, are: (1) leakage rate, which is addressed by the established threshold of 10 gpm in the current EAL of the IC; and (2) the duration of the leakage, which is addressed by requiring isolation of the leak in the proposed change to the IC. An additional concern stated in both the Regulatory Guide and the NUREG is the location of the leak. RCS leakage detection capabilities within containment are unchanged. Detection of RCS leakage outside of containment is what is impacted by this proposed change. The proposed IC for Unusual Event due to RCS leakage addresses both the duration of leakage and RCS leakage outside of containmentlRCPB. A leak of RCS in excess of 10 gpm must be readily isolable within the bounds of initial operator action or an emergency is declared. A leak in an interfacing system that exceeds 10 gpm and is non-isolable requires an emergency declaration. The Technical Specification allows for

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 1 Page 5 isolating the high pressure portion of the system from the low pressure systems in an effort to understand the leakage. The expectation for isolation/termination of the leak is promptly, with promptly being within the bounds of initial operator actions in off-normal operating procedures or emergency operating procedures.

In accordance with 10 CFR 50.54 (q), a "licensee may make changes to these plans

[Emergency Plan] without Commission approval only if the changes do not decrease the effectiveness of the plans and the plans, as changed, continue to meet the standards of 10 CFR 50.47(b) and the requirements of appendix E to this part." The proposed change seeks to revise the original acceptable alternative by re-establishing a link to the Technical Specifications and therefore mode dependence as originally defined in the NUREG-0654 RCS leakage IC. Implementation of this change would eliminate the basis for entry into the Emergency Plan for conditions (i.e. Mode 5 and 6) that, prior to this change, would have implemented the Emergency Plan. No RCS leakage Technical Specification is provided for the St. Lucie Plant for Modes 5 or 6. In fact, NUREG-0654 does not include a low mode emergency action levels for RCS leakage. The current EAL scheme for St. Lucie addresses the low mode conditions through the IC for the inability to maintain cold shutdown and the loss of sub-cool margin. Technical Specification RCS leakage in excess of 10 gpm that is isolable does not "indicate a potential degradation of the level of safety of the plant." FPL feels that this change to the Emergency Plan enhances the program in that it no longer unnecessarily focuses offsite emergency management attention on a non-emergency condition.

The proposed revision of this IC remains in agreement with the NUREG-0654 scheme of emergency classification and the class description for Unusual Event and continues to meet 10 CFR 50.47 (b) and Appendix E. The revised IC will also continue to provide a logical transition to the IC for Alert within the classification table event/category, "Abnormal Primary Leak Rate.". If a RCS leak were in excess of 50 gpm and unisolable, then conditions would require declaration of Alert.

The proposed change provides an alternate to the existing IC/EAL that is more in line with NUREG-0654 but less restrictive than the current IC/EAL. Therefore, FPL seeks NRC approval prior to implementation of this change to the Emergency Plan.

D. STATE/LOCAL GOVERNMENT REVIEW/CONCURRENCE FPL will obtain the concurrence of state and local governments prior to implementing this change.

E. SUPPORTING REFERENCES The following documents, in total or in part, have been included as Attachment 3 to this submittal:

E.1 Safety Evaluation of Proposed Emergency Action Level Revision for St.

Lucie Plant, Units 1 and 2 (TAC NOS. M96274 and M96275).

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 1 Page 6 E.2 St. Lucie Plant Updated Final Safety Analysis Report, Unit 1 - Chapter 5, Reactor Coolant System and Unit 2 - Section 5.0, Reactor Coolant System and Connected Systems.

E.3 NUREG-1432, Vol.1, Rev. 3.0, Standard Technical Specifications Combustion Engineering Plants, June 2004.

E.4 NRC Regulations, Title 10, Code of Federal Regulations Part 50.2, Definitions.

E.5 NEI 99-01 Rev. 4 (NUMARC/NESP-007), Methodology for Development of Emergency Action Levels, January 2003.

E.6 St. Lucie Plant Technical Specifications Unit 1 - 3.4.6.2, Reactor Coolant System and Unit 2 - 3.4.6.2, Reactor Coolant System.

E.7 NUREG-0818, Emergency Action Levels for Light Water Reactors, October 1981.

E.8 NUREG-0654, FEMA-REP-1, Rev.1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, November 1980.

E.9 Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973.

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 2 Page 1 Attachment 2 Comparison of Current IC/EAL to Proposed IC/EAL NUREG 0654 Current IC/EAL l Proposed IC/EAL

5. Exceeding either Reactor Coolant System (RCS) Leakage Reactor Coolant System (RCS) Leakage primary/secondary leak rate technical specification or 1. RCS leakage GREATER THAN 10 gpm 1.Unisolable Technical Specification RCS leakage primary system leak rate as indicated by: GREATER THAN 10 gpm as indicated by:

technical specification.

A. Control Room observation NOTE OR

If the leak is from an interfacing system (e.g., SDC, B. Inventory balance calculation LPSI, CVCS, etc.) and the leak is readily isolable from OR the Reactor Coolant Pressure Boundary, the leak C. Field observation should not be considered RCS leakage.

OR

To be isolable, personnel must be able to promptly D. Emergency Coordinator Judgment close the valve(s) which isolates the leak within the context of initial operator actions.

OR A. Control Room Observation

6. Failure of a safety or relief 2. Indication of leaking RCS safety or relief OR valve in a safety related valve which causes RCS pressure to drop B. Inventory balance calculation system to close following below SIAS set points: OR reduction of applicable C. Field observation pressure. - Unit 1 - 1600 psia OR

- Unit 2 - 1736 psia D. Emergency Coordinator's judgment OR

2. Indication of leaking RCS safety or relief valve causes RCS pressure to drop below SIAS setpoints:

- Unit 1 - 1600 psia

- Unit 2- 1736 psia

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 1 Attachment 3 Supporting References E.1 Safety Evaluation of Proposed Emergency Action Level Revision for St.

Lucie Plant, Units 1 and 2 (TAC NOS. M96274 and M96275).

E.2 St. Lucie Plant Updated Final Safety Analysis Report, Unit 1 - Chapter 5, Reactor Coolant System and Unit 2 - Section 5.0, Reactor Coolant System and Connected Systems.

E.3 NUREG-1432, Vol.1, Rev. 3.0, Standard Technical Specifications Combustion Engineering Plants, June 2004.

E.4 NRC Regulations, Title 10, Code of Federal Regulations Part 50.2, Definitions.

E.5 NEI 99-01 Rev. 4 (NUMARC/NESP-007), Methodology for Development of Emergency Action Levels, January 2003.

E.6 St. Lucie Plant Technical Specifications Unit 1 - 3.4.6.2, Reactor Coolant System and Unit 2 - 3.4.6.2, Reactor Coolant System.

E.7 NUREG-0818, Emergency Action Levels for Light Water Reactors, October 1981.

E.8 NUREG-0654, FEMA-REP-1, Rev.1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, November 1980.

E.9 Regulatory Guide 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems, May 1973.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 2 Reference E.1 U*ED STATES KUHEIVED NUCLEAR REGULATORY CoMMpsSION OT2 9 Z 0, WAN4INGTOW. Oc. ww we O CT 2 2 199 ICO October 17, 1996 NuclearikensIng Mr. Thomas F. Plunkett President, Nuclear Division Florida Power and Light Cocipany Post Off Ice Box 14000 Juno Beach, Florida 33408-0420

Dear Kr. Plunkett:

SUBJECT:

SAFElY EVALUATION OF PROPOSED EMERGENCY ACTION LEVEL REVISION FOR ST. LUCIE PLANT, UNITS I AND 2 (TAC NOS. H96274 MND 196275)

By letter dated July 25, 1996, Florida Power and Light proposed a revision to Table 3-1 of the St. Lucie Plant Emergency Plan. NRC approval was requested, prior to implementation, of a change to an Emergency Action Level (EAL) regarding reactor coolant system (RCS) leakage. The proposed change revises the declaration threshold for the EAL of an Unusual Event involving RCS leakage.

The NRC staff has completed its review of the proposed change and found that the proposed EAL meets the requirements of 10 CFR 50.47(b)(4) and Appendix E to 10 CFR 50 for emergency classification and action level schemes. Our Safety Evaluation is enclosed.Section IV.B of Appendix E to 10 CFR Part 5Q requires agreement by State and local government authorities to changes to the plant's EALs. In a telephone call with Ceorge Madden of your staff on October 11, 1996, Mr. Madden stated that agreement with State and local officials would be obtained prior to EAL change implementation, or the ihinge would not be imPlemented. Based upon this commitment. you are hereby authorized to implement this change to the St. Lucie Plant Radiological Emergency Plan conditioned upon your obtaining the agreement of appropriate.

State and local officials prior to implementation thereof. This completes our action on TAC Nos. M96274 and 116275.

Sincerely, Leonard A Wiens, Senior Project Manager

,Project Directorate 11-3 Division of Reactor ProJects-I/Il Office of Nuclear Reactor Regulation Docket No. 50-335 .

and 50-389.

Enclosure:

Safety Evaluation cc w/enclosure: See next page

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 3 Reference E.1 UNITE STATES NUCLEAR REGULATORY COMMISSION WAeH5 0t . C.

DI.TO SAFETY EVALUATION B 0fUCLEAR REACTOR REGULATION IFFCE RELATED TO AN EXERGENC ACTION LEVEL REVISTON FIORIDA POWER AND LTCGT COHPANY. FT AL.

ST. LUCIE PLANT. UNIT WOS. I AN 2 DOCKET NOS.' 50-335 AND 5E389 1.O I0TR2O &II(

By letter dated July 25, 1996, Florida Power and Light Cocpany (FPL) proposed a revision to Table 3-1 of their Emergency Plan. They requested approval, prior to implementation, of a change to an Emergency Action Level (EAL) regarding reactor coolant system (RCS) leakage. The proposed EAL change revises the declaration threshold for an Unusual Event involving RCS leakage.

2.0 BACKGROUND

The proposEd EAL change was reviewed -against the requirements in 10 CFR 50.4M7(b)(4) and Appendix E to 10 CFR 50. Section 50.47(b)(4) specifies that onsite emergency plans must meet the following standard: A standard emergency classification and action level scheme, the bases of which include factiity system and effluent parameters, Is in use by the nuclear facility licensee....'

Section IY.C. of Appendix E to 10 CFR 50 specifies that, 'Emergency action levels (based not only on onsite and offsite radiation monitoring Information but also on readings from a number of sensors that Indicate a potential emergency, such as the pressure In containment and the response of the Emergency Core Cooling System) for notification of-offsite agencies shall be described.... The ewergency classes defined shall include; .(1) notification of unusual events, (2) alert, (3) site area emergency, and (4) general emergency.

The current EAL followed the general guidelines for EAls set forth In Appendix I of KUREG-0654, 'Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants* (November 1980). The proposed revision Incorporates the enhancements andoclarifications to the EALs based on the guidelines for EALs set forth in KUKARC/NESP-007, Revision 2, "Methodology for Development of Emergency Action Levels' (January 1992). The NRC endorsed the use of either NUREG-0654 or XUHARC/XESP-007 in Regulatory Guide 1.101, 'Emergency Planning and Preparedness for Nuclear Power Reactors,' Revision 3, August 1992.

RegulatorX Guide 1.101 proyides acceptable methods by which licensees may meet the requirements of 10 CFR 50.47(b)(4) and Appendix E to 10 CFR 50. In Emergency Preparedness Position (EPPOS) Number 1, OEmergency Preparedness Position on Acceptable Deviations from Appendix 1 of HUREG-0654 Based Upon the Staff's Regulatory Analysis of HUMARC/HESP-007 (June 1 1995), the HRC staff recognized -that RUREG-0654-based EALs could be enhanced and clarified by application of the technical bases for NUMARC/NESP-007-based EALs. The staff relied upon the guidance Inthese documents as the basis for its review of the St. Lucie proposed EAL revision.

St. Lucie Units 1 and 2

-Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 4 Reference E.1 2

3.0 EYALUATION The licensue followed the guidance and logic presented in NUHUPC/NESP-007 for determining the level of leakage to declare an Notification of an Unusual Event (NOUE). The licensee revised the threshold for an NOUE and EA1 based upon RCS leakage from two values:

(1) greater than I gpm for unidentified leakage and (2) greater than 10 9pm for identified leakage; to one threshold value:

greater than IO.gp. for all leakage.

The new proposed threshold leakage value is higher for unidentified sources than the current EAL and is the saue as leaks from Identified sources. The value for Identified leaks are more conservative than the value shown in HUHARC guidance. However, the licensee indicated that the selected value and the establishment of a single threshold value makes the declaration of an NUE quicker and easier. The determination of the quantity and the scurce of the leak will be simpler and more timely under the proposed EAL. The licensee will not have to confirm the leakage levels through mass inventory balances and the licensee will not have to spend tine to differentiate between identified and unidentified leaks.

Although Regulatory Culde 1.101 admonishes against the nixing of the emergency classification guidance in HUUARC/NESP-007 with that inAppendix 1 to -

XUREG-0654, it is recognized that licensees who continue to utilize the example initiating conditions inAppendix I to UUREC-0654 as the basis for their classification scheme could benefit from the guidance in HUHARC/HESP-007. To that and, licensees could utilize the technical bases under the example EALS inKULARC/NESP-007 to enhance and clarify some of their site-specific EALs developed from KUREG-0654. The chosen classification scheme, whether based on Appendix 1 to HUREG-0654 or HUYARC/HESP-007, must remain internally consistent.

The staff found the proposed revisions and associated justification, provided by the licensee, to be acceptable. The proposed EAL revision for the St. Lucie Plant, is consistent with the guidance provided by HUREG-0654 and allowable deviations, as discussed in EPPOS 1 in accordance with the technical bases for EAls InNUHURC/JESP-007.

4.0 COLUSTO _

As a result of our review, we have concluded that the proposed EAL meets the requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50 for emergency classification and action level schenes. However, in addition to NRC approval of EAL changes, Section 1V.B of Appendix E to 10 CFR Part S0 requires agreement-by State and local government authorities to changes to thi plant's EALs. Therefore, the licensee should obtain such agreement prior to implementation of the proposed EUL.

Principal Contributor: L. Cohen Dated: October 17, 1996

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 5 Reference E.2 - Unit 1 CHAPTER 5 REACTOR COOLANT SYSTEM This chapter was originally prepared to describe the reactor coolant system during the Initial fuel cycle.

Much of the original text is retained for historical record. However, where applicable, changes have been made to reflect the uprating of the unit to a stretch power level of 2700 Mwt. Where Information associated with the higher power level Is not available the existing Information is identified as 'pre-stretch' or 'cycle 1."

5.1

SUMMARY

DESCRIPTION The function of the reactor coolant system Is to remove heat from the reactor core and transfer it to the secondary (steam generating) system. In a pressurized water reactor the steam generators represent the points of separation between the reactor coolant system and the main steam system. The steam generators are vertical U-tube heat exchangers In which heat Is transferred from the reactor coolant to the main steam system. Reactor coolant Isseparated from the boiler water by the steam generator tube sheet. The reactor coolant system Is a closed system which forms a barrier to the release of radioactive materials Into the containment.

Plan and elevation views of the arrangement of the reactor coolant system are shown In Figures 5.1-1 and 5.1-2, respectively. The piping and Instrumentation (P&l) diagram of the reactor coolant system Is shown In Figure 5.1-3. The major components of the system are the reactor vessel, two heat transfer loops, each containing one steam generator and two reactor coolant pumps, a pressurizer connected to the loop 1B reactor vessel outlet pipe; and connecting inlet and outlet, spray and surge line piping. A quench tank Is provided to receive, condense, and cool steam discharges from the pressurizer safety and power operated relief valves. All components are located Inside the containment, and the relationship of the equipment arrangement to the containment structure Is shown in Figure 1.2-7 through 1.2-11.

Table 5.1-3 shows the principal pressures, temperatures, flow rates and coolant volumes of the reactor coolant system components under pre-stretch normal steady state, full power operating conditions by means of numbered locations (See Figure 5.1-3). Figure 5.1-3 has a detailed representation of the reactor coolant system. Instrumentation provided for operation and control of the system is described in Section 7.

System pressure is maintained by regulating the water temperature in the pressurizer where steam and water are held in thermal equilibrium. Steam is either formed by the pressurizer heaters or condensed by the pressurizer spray to limit the pressure variations caused by contraction or expansion of the reactor coolant. The pressurizer Is located with Its base at a higher elevation than the reactor coolant loop piping. This eliminates the need for a separate pressurizer drain, and ensures that the pressurizer is drained before maintenance operations. The average temperature of the reactor coolant varies with power level, and the fluid expands or contracts, changing the pressurizer water level.

5.1-1 Amendment 15, (1/97)

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 6 Reference E.2 - Unit 2 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS This chapter was originally prepared to describe the reactor coolant system during the initial fuel cycle. Much of the original text is retained for historical record. However, where applicable, changes have been made to reflect the uprating of the unit to a stretch power rating of 2700 Mwt.

Where information associated with the higher power level is not available the existing information is identified as Cycle 1.

5.1

SUMMARY

DESCRIPTION The reactor is a pressurized water reactor with two coolant loops. The Reactor Coolant System (RCS) circulates water in a closed cycle, to remove heat from the reactor core and transfers it to a secondary (steam generating) system. The steam generators provide the Interface between the Reactor Coolant (primary) System and the Main Steam (secondary) System. The steam generators are vertical U-tube heat exchangers in which heat is transferred from the reactor coolant to the Main Steam System. Reactor coolant is prevented from mixing with the main steam by the steam generator tubes and the steam generator tube sheet. The RCS is a closed system thus forming a barrier to the release of radioactive materials.

The arrangement of the RCS is shown on Figures 5.1-1 and 5.1-2. The major components of the system are the reactor vessel; two parallel heat transfer loops, each containing one steam generator and two reactor coolant pumps; a pressurizer connected to one of the reactor vessel outlet pipes; and associated piping. All components are located inside containment.

Reactor Coolant System pressure is controlled by the pressurizer, where steam and water are maintained in thermal equilibrium. Steam is formed by energizing immersion heaters in the pressurizer, or is condensed by the pressurizer spray to limit pressure variations caused by contraction or expansion of the reactor coolant. The average temperature of the reactor coolant varies with power level and the fluid expands or contracts, changing the pressurizer water level.

The charging pumps and letdown control valves in the Chemical and Volume Control System (CVCS) are used to maintain a programmed pressurizer water level. A continuous but variable letdown purification flow is maintained to keep the RCS chemistry within prescribed limits. Two charging nozzles and a letdown nozzle are provided on the reactor coolant piping for this operation. The charging flow is also used to alter the boron concentration or correct the chemical content of the reactor coolant.

Other reactor coolant loop penetrations are the pressurizer surge line in one reactor vessel outlet pipe; the four safety injection inlet nozzles, one in each reactor vessel inlet pipe; two outlet nozzles to the Shutdown Cooling System, one in each reactor vessel outlet pipe; two pressurizer spray nozzles; vent and drain connections; and sample and instrument connections.

Overpressure protection for the reactor coolant pressure boundary is provided by three spring-loaded ASME Code pressurizer safety valves connected to the 5.1-1 Amendment No. 13, (05/00)

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 7 Reference E.3 NUREG-1432 Vol 1,Rev.3.0 r_

Standard Technical, Specifications Combustion Engineering Plants Specifications Manuscript Completed March 2004 Date Published: June 2004 Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-C001

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 8 Reference E.3 RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm Identiffed LEAKAGE,

d. 1 gpm total primary to secondary LEAKAGE through all steam generators (SGs), and

e. (720] gallons per day primary to secondary LEAKAGE through any one SG.

APPLICABILITY: MODES 1, 2. 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS LEAKAGE not A.1 Reduce LEAKAGE to within 4hours within limits for reasons limits.

other than pressure boundary LEAKAGE.

S. Required Action and 8.1 Be In MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

8.2 Be In MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

CEOG STS 3.4.1 3-1 Rev. 3.0, 03131/04

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 9 Reference E.3 RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS _

SURVEILLANCE FREQUENCY SR 3.4.13.1 NOTE- -

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify RCS operational LEAKAGE Iswithin limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months performance of RCS water Inventory balance.

SR 3.4.13.2 Verify SG tube Integrity Is In accordance with the In accordance Steam Generator Tube Surveillance Program. with the Steam Generator Tube Surveillance Program CEOG STS 3.4.13-2 Rev. 3.0, 03/31104

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 10 Reference E.3 RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves Isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO Isto limit system operation In the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE Into the containment area Is necessary. Quickly separating the Identified LEAKAGE from the unidentified LEAKAGE Is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur detrimental to the safety of the facility and the public.

A limited amount of leakage Inside containment Is expected.from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and Isolated from the containment atmosphere, if possible, to not Interfere with RCS LEAKAGE detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from Inadequate cooling, In addition to preventing the accident analysis radiation release assumptions from being exceeded. The consequences of violating this LCO Include the possibility of a loss of coolant accident (LOCA).

B 3.4.13-1 Rev. 3.0, 03/31104 CEOG STS CEOG STS B 3.4.13-1 Rev. 3.0, 03131/04

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 11 Reference E.3 RCS Operational LEAKAGE B 3A.13 BASES APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY address operational LEAKAGE. However, other operational LEAKAGE ANALYSES Is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event The safety analysis for an event resulting In steam discharge to the atmosphere assumes a 1 gpm primary to secondary LEAKAGE as the Initial condition.

Primary to secondary LEAKAGE Isa factor in the dose releases outside containment resulting from a steam line break (SLB) accident To a lesser extent, other accidents or transients Involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid Is only briefly released via safety valves and the majority Is steamed to the condenser. The I gpmprimary to secondary LEAKAGE Is relatively Inconsequential.

The SLB Is more limiting for she radiation releases. The safety analysis for the SLi3 accident assumes I gpm primary to secondary LEAKAGE In one generator as an Initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 50 or the staff approved licensing basis (I.e., a small fraction of these limits).

RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE Is allowed, being Indicative of material deterioration. LEAKAGE of this type Is unacceptable as the leak Itself could cause further deterioration, resulting In higher LEAKAGE. Violation of this LCO could result In continued degradation of the RCPB. LEAKAGE past seals and gaskets Is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE Is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result In continued degradation of the RCPB, If the LEAKAGE Is from the pressure boundary.

Rev. 3.0, 03131/04 STS B 3.4.13-2 CEOG CEOG STS B 3.4.13-2 Rev. 3.0, 03/31/04

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 12 Reference E.3 RCS Operational LEAKAGE B 3.4.13 BASES LCO (continued)

c. Identified LEAKAGE Up to 10 gpm of Identified LEAKAGE Is considered allowable because LEAKAGE Isfrom known sources that do not Interfere with detection of unidentified LEAKAGE and Is well within the capability of the RCS makeup system. Identified LEAKAGE includes .LEAKAGE to the containment from specifically known and located sources, but does not Include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result In continued degradation of a component or system.

LCO 3.4.14, 'RCS Pressure Isolation Valve (PIV) Leakage,'

measures leakage through each individual PIV and can Impact this LCO. Of the two PiVs In series In each Isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other Is leaktight. If both valves leak and result In a loss ofnmass from the RCS, the loss must be Included in the allowable identified LEAKAGE.

d. Primary to Secondary LEAKAGE through All Steam Generators (SGs)

Total primary to secondary LEAKAGE amounting to I gpm through all SGs produces acceptable offsfte doses In the SLB accident analysis. Violation of this LCO could exceed the offsite dose limits for this accident analysis. Primary to secondary LEAKAGE must be Included In the total allowable limit for Identified LEAKAGE.

e. Primary to Secondary LEAKAGE through Any One SG The [720] gallon per day limit on primary to secondary LEAKAGE through any one SG allocates the total I gpm allowed primary to secondary LEAKAGE equally between the two generators.

APPLICABILITY In MODES 1, 2,3, and 4, the potential for RCPB LEAKAGE Is greatest when the RCS Ispressurized.

In MODES 5 and B, LEAKAGE limits are not required because the reactor coolant pressure Is far lower, resulting In lower stresses and reduced potentials for LEAKAGE.

CEOG STS B 3.4.13-3 Rev. 3.0, 03/31104

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 13 Reference E.3 RCS Operational LEAKAGE B 3.4.13 BASES ACTIONS A.1 Unidentified LEAKAGE, Identified LEAKAGE, or primary to secondary LEAKAGE In excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage rates and either Identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action Is necessary to prevent further deterioration of the RCPB.

B.1 and B.2 If any pressure boundary LEAKAGE exists or If unidentified, identified, or primary to secondary LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences.

The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

The allowed Completion Times are reasonable, based on operating experience, to reach the required conditions from full power conditions In an orderly manner and without challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration Is much less likely.

SURVEILLANCE SR 3.4,13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB Is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. Unidentified LEAKAGE and Identified LEAKAGE are determined by performance of an RCS water inventory balance. Primary to secondary LEAKAGE is also measured by performance of an RCS water Inventory balance In conjunction with effluent monitoring within the secondary steam and feedwater systems.

The RCS water Inventory balance must be performed with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, [and RCP seal injection and return flows]). Therefore, a Note Is added allowing that this SR Is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

8 3.4.13-4 Rev. 3.0, 03131104 STS CEOG STS 8 3.4.13-4 Rev. 3.0, 03131/04

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 14 Reference E.3 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued)

Steady state operation Is required to perform a proper water Inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water Inventory balance, steady state Is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE Is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level.

These leakage detection systems are specified In LCO 3.4.15, 'RCS Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency Is a reasonable Interval to trend LEAKAGE and recognizes the Importance of early leakage detection In the prevention of accidents.

SR 3.4.13.2 This SR provides the means necessary to determine SG OPERABILITY In an operational MODE. The requirement to demonstrate SG tube integrity In accordance with the Steam Generator Tube Surveillance Program emphasizes the Importance of SG tube Integrity, even though this Survelilance cannot be performed at normal operating conditions.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.

2. Regulatory Guide 1.45, May 1973.

3. FSAR, Section [15].

CEOG STS B 3.4.13-5 Rev. 3.0, 03/31104

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 15 Reference E.4 other entity; and (2) any legal successor, representative, agent, or agency of the foregoing.

Price-cap regulation means the system of rate regulation In which a rate regulatory authority establishes rates electric generator may charge its customers that are based on a specified maximum price of electricity.

Procurement document means, for the purposes of § 50.55(e) of this chapter, a contract that defines the requ which facilities or basic components must meet In order to be considered acceptable by the purchaser.

Produce, when used in relation to special nuclear material, means (1) to manufacture, make, produce, or refin nuclear material; (2) to separate special nuclear material from other substances In which such material may be or (3) to make or to produce new special nuclear material.

Production facility means:

(1) Any nuclear reactor designed or used primarily for the formation of plutonium or uranium-233; or (2) Any facility designed or used for the separation of the Isotopes of plutonium, except laboratory scale faciliti or used for experimental or analytical purposes only; or (3) Any facility designed or used for the processing of Irradiated materials containing special nuclear material, laboratory scale facilities designed or used for experimental or analytical purposes, (ii) facilities In which the on nuclear materials contained In the Irradiated material to be processed are uranium enriched In the Isotope U-23 plutonium produced by the Irradiation, if the material processed contains not more than 106 grams of plutoniu of U-235 and has fission product activity not In excess of 0.25 millicuries of fission products per gram of U-235 facilities In which processing Is conducted pursuant to a license Issued under parts 30 and 70 of this chapter, o regulations of an Agreement State, for the receipt, possession, use, and transfer of irradiated special nuclear m which authorizes the processing of the irradiated material on a batch basis for the separation of selected fission and limits the process batch to not more than 100 grams of uranium enriched In the Isotope 235 and not more grams of any other special nuclear material.

Reactorcoolant pressure boundary means all those pressure-containing components of boiling and pressurized cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves, which are:

(1) Part of the reactor coolant system, or (2) Connected to the reactor coolant system, up to and Including any and all of the following:

(i) The outermost containment Isolation valve In system piping which penetrates primary reactor containment, (ii) The second of two valves normally closed during normal reactor operation in system piping which does not primary reactor containment, (iii) The reactor coolant system safety and relief valves.

For nuclear power reactors of the direct cycle boiling water type, the reactor coolant system extends to and Inc outermost containment Isolation valve In the main steam and feedwater piping.

Research and development means (1) theoretical analysis, exploration, or experimentation; or (2) the extenslo Investigative findings and theories of a scientific or technical nature Into practical application for experimental a demonstration purposes, Including the experimental production and testing of models, devices, equipment, ma processes.

Responsible officer means, for the purposes of § 50.55(e) of this chapter, the president, vice-president, or othe in the organization of a corporation, partnership, or other entity who Is vested with executive authority over ac subject to this part.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 16 Reference E.5 NEI 99-01 Rev. 4 (NUMARC/NESP-007)

Methodology for Development of Emergency Action Levels January 2003 Revision 01/2003

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 17 Reference E.5 The (site-specific) value for the "Potential Loss" EAL corresponds to the top of the active fuel. For sites using CSFSTs, the "Potential Loss EAL is defined by the Core Cooling - ORANGE path. The (site-specific) value In this EAL should be consistent with the CSFST value.

5. Containment Rtadiatlon Monitoring The (site-specific) reading is a value which Indicates the release of reactor coolant, with elevated activity indicative of fuel damage, Into the containment. The reading should be calculated assuming the Instantaneous release and dispersal of the reactor coolant noble gas and Iodine inventory associated with a concentration of 300 .iCVgm dose equivalent 1-131 Into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including Iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value Is higher than that specified for RCS barrier Loss

iAL #4. Thus, this EAL Indicates a loss of both the fuel clad barrier and a loss of RCS barrier.

There Is no "Potential Loss" EAL associated with this item.

6. Other (Site-Specific) Indications This SAL Is to cover other (site-specific) Indications that may Indicate loss or potential loss of the Fuel Clad barrier, Including Indications from containment air monitors or any other (site-specific)

Instrumentation.

7. Emergency Director Judgment This EAL addresses any other factors that are to be used by the Emergency Director In determining whether the Fuel Clad barrier.is lost or potentially lost. In addition, the inability to monitor the barrier should also be Incorporated In this EAL as a factor in Emergency Directorjudgment that the barrier may be considered lost or potentially lost. (See also IC SGI, "Prolonged Loss or All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional Information.)

RCS BARRIER EXAMPLE EALs: (1 or2or3or4or5or6)

The RCS Barrier includes the RCS primary side and Its connections up to and Including the pressurizer safety and relief valves, and other connections up to and Including the primary Isolation valves.

1. Critical Safety Function Status This EAL Is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional restoration procedures. For more Information, please refer to Section 3.9 of this report.

RED path Indicates an extreme challenge to the safety function derived from appropriate instrument readings, and these CSFs Indicate a potential loss of RCS barrier.

There Is no "Loss" EAL associated with this iem.

2. RCS Leak Rate The "Loss" EAL addresses conditions where leakage from the RCS Is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling Is the fundamental Indication that the Inventory control systems are Inadequate In maintaining RCS pressure and Inventory against the mass loss through the leak.

Revision 01/2003 5-F-1 5

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 18 Reference E.6 - Unit 1 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. I GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total primary-to-secondary leakage through steam generators,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. Leakage as specified in Table 3.4.6-1 for each Reactor Coolant System Pressure Isolation Valve identified in Table 3.4.6-1.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 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 .

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and Reactor Coolant System Pressure Isolation Valve leakage, reduce the leak-age rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 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 .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit in 3.4.6.2.e above reactor operation may continue provided that at least two valves, including check valves, in each high pressure line having a non-functional valve are in and remain in the mode corresponding to the isolated con-dition. Motor operated valves shall be placed in the closed posi-tion, and power supplies deenergized. (Note, however, that this may lead to ACTION requirements for systems involved.) Otherwise, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 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 .

SURVEILLANCEREQUIREMENTS 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by.

a. Monitoring the containment atmosphere gaseous and particulate radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT 1 314 4-14 d ted-42OM4 0dor Amendment No. 118

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 19 Reference E.6 - Unit 1 REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE SRVEILLANCE _REOUIREMENT [Continued)

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation except when operating in the shutdown cooling mode,

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

e. Verifying each Reactor Coolant System Pressure Isolation Valve leakage (Table 3.4.6-1) to be within limits:

1. Prior to entering MODE 2 after refueling,

2. Prior to entering MODE 2, whenever the plant has been in COLD SHUTDOWN for 7 days or more if leakage testing has not been performed in the previous 9 months,

3. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

4. The provision of Specification 4.0.4 is not applicable for entry into MODE 3 or 4.

f. Whenever integrity of a pressure isolation valve listed in Table 3.4.6-1 cannot be demonstrated the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in each high pressure line having a leaking valve shall be recorded daily.

ST. LUCIE - UNIT I 3/4 4-14a Amendment No. 133

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 20 Reference E.6 - Unit 1 TABLE 3.4 6-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Check Valve No.

V3227 V3123 V3217 V3113 V3237 V3133 V3247 V3143 V3124 V3114 V3134 V3144 NOTES (a) Maximum Allowable Leakage (each valve):

1. Leakage rates less than or equal to 1.0 gpm are acceptable.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount the reduces the margin between previous measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are unacceptable.

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

(c) Minimum test differential pressure shall not be less than 150 psid.

ST. LUCIE - UNIT 1 314 4-14b Order dated 4120181

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 21 Reference E.6 - Unit 2 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CON21TIOQ FOR OPERATnQN 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. I gpm total primary-to-secondary leakage through steam generators and 720 gallons per day through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. I gpm leakage (except as noted InTable 3.4-1) at a Reactor Coolant System pressure of 2235 +/- 20 psig from any Reactor Coolant System Pressure Isolation Valve specified InTable 3.4-1.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be Inat least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and InCOLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System leakage greater than any one of the limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be Inat 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, Isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be Inat 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With RCS leakage alarmed and confirmed Ina flow path with no flow Indication, commence an RCS water Inventory balance within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to determine the leak rate.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by.

a. Monitoring the containment atmosphere gaseous and particulate radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. Monitoring the containment sump Inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT 2 314 4-19

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 22 Reference E.6 - Unit 2 REACTOR COOLANT SYSTEM SURVEILLANCE REDUIRFMENTS (Continued)

c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve check valve specified InTable 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within Its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been InCOLD SHUTDOWN for 7 days or more and If leakage testing has not been performed Inthe previous 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve,

d. Following valve actuation due to automatic or manual action or flow through the valve:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and

2. Within 31 days by verifying leakage rate.

4.4.6.2.3 Each Reactor Coolant System Pressure Isolation Valve motor-operated valve specified InTable 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within Its limit;

a. At least once per 18 months, and

b. Prior to returning the valve to service following maintenance, repair, or replacement work on the valve.

The provisions of Specification 4.0.4 are not applicable for entry Into MODE 3 or 4.

ST. LUCIE - UNIT 2 314 4-20 Amendment No. 72

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 23 Reference E.6 - Unit 2 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES Check Valve No. Motor Onerated Valve No.

V3217 V3525 V3480 V3227 V3524 V34B1 V3237 V3527 V3652 V3247 V3526 V3651 V3259 V3258 V3260 V3261 V3215 V3225 V3235 V3245 NOTES (a) Maximum Allowable Leakage (each valve):

1. Except as noted below leakage rates greater than 1.0 gpm are unacceptable.

2. For motor-operated valves (MOVs) only, leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount the reduces the margin between previous measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

3. For motor-operated valves (MOVs) only, leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are unacceptable.

(b) To satisfy ALARA requirements, leakage may be measured Indirectly (as from the performance of pressure Indicators) If accomplished In accord-ance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

(c) Minimum test differential pressure shall not be less than 200 psid.

ST. LUCIE - UNIT 2 14 4-21

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 24 Reference E.7 NUREG-0818 Emergency Action Levels Emergency Afction Levels for Light Water Reactors Draft Report for Comment VUpmc*%Crad. Auvnt a0l1 DM Pubbho&h0ctoer1981 Divislon of Emergncy Prepareness Office of Inspection and Enforcement U.S. Nuclear Regulatory Commisson Washington, D.C. 20R

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 25 Reference E.7 E E INITIATING CONDITIONS: NOTIFICATION OF UNUSUAL EVENT

1. Emergency Core Cooling Systems (ECCS) initiated and discharge to vessel.

2. Radiological efflueat technical specification limits exceeded.

3. Fuel dage Indication. Examples:

a. H1g0i offgas at 8WR air ejector monitor (greater than 50,000 Pcf/sec; corresponding to 16 isotopes decayed to 30 minutes; or an increase of ItO,000 izci/sec within a 30 minut time period)

b. High coolant activity sample (e.g.. exceeding coolant technical specifications for iodine spike)

c. Failed fuel monitor (PSR) indicated increase greater than 0.1%

equivalent fuel failures within 30 minutes.

4. Abnormal coolant temperature and/o- pressure or abnorm.1 fuel temperatures outside of technical specification limits.

5. Exceeding either primary/secondary leak rate technical specification or primary system leak rate technical specification.

6. Failure of a safety or relief valve In a safety related system to close following reduction ot applicable pressure.

7. Loss of offsite power or loss of onsite AC power capability.

8. Loss of containment integrity requiring shutdown by technical specifications.

9. Loss of engineered safety feature or fire protection system function requiring shutdown be technical specifications (e.g., because of malfunction, personnel error or procedural inadequacy).

10. Fire within the plant lasting more than 10 minutes.

11. Indications or alarms on process or effluent parameters not functional in control roon to an extent requiring plant shutdown or other significant loss of assessment or cocnunication capability (e.g., plant computer.

Safety Paraneter Display System, all meteorological instrunentation).

12. Security threat or attempted entry or attempted sabotage.

13. Natural phenomenon being experienced or projected beyond usual levels:

a. Any earthquake felt in-plant or detected on station seismic instrumentation 11

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 26 Reference E.7

b. 50 yur flood or low witer. turnal, bxrrfcsaf surge, sachoe C. lurtorado sn site

d. Any torr1cane.

14. Other hazards being experienced or projected:

a. Aircraft erash o&-site or ulwsual arcratt activity over facil'ty

b. Train derailsmnt on-site

e. Halr or onsite Seplosliw

d. Hear or onsits toxic or flamable ps reloase.

a. Turbine rotating ceapwMt failure causing rapid plant shutdown.

15. Othei plant cniltiors exist that warrant increased awareness on the part of a pI&At operating staff or State and/or local offiste auth~rlties or require plant slmtdown wnar technical seification requirements or involve other than Rmol 3Ltpolled thutdown (e.g., cooldown rate emsCedfag teClefca1 spCfCeition lWits, pIpe cracking found during

operation).

16. lramportation of contaminated lnJired individual froa sut to offstte 1hospital.

1. Rapid depressuriziiiot of PYR secondary side.

12

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 27 Reference E.7 acceptable value of subcooltng margin will differ depending on %tetherthe n Condition No. 5 ant1itn2 Mxceing either primrylryscondary leak rate techn cal specification or 7 primary system leak rate technical specification. I P raf t DU~

Primary to secondary leak rate greater than 1 9p4 total for more than four hours or greater than 500 gpm per steau generator as Identified by daily RCS leakage evaluation; or Primary system leat rate grtattr than those specified in Technical Specification 3.4.6.2 as identified by daily RCS leakage evaluation.

1. .> 0 pressur.^ boundary leakage

2. > 1 gpm unidentified for rore than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

3. ) 10 gpm Identified RCS leakage for more tnan 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

4. > 30 gp1 controlled leakage (2235 +/- 20 psig) for more enan 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Discussicn Tersonse is adequate.

Failure of a safety or relief valve In a safety related system to closa following reduction of applicable pressure.

Draft EALs Pressurizer or steam generator safety cr relief valve opens and then falls to reset as indicated by:

1. Pressurizer relief valve position 11jht indicates open; or Pressurizer safety valve position Indicator reads greater than lX; or Valid accoustical monitor indication.

24

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 28 Reference E.8 NUREG-0654 FEMA-REP-1 Rev. 1 Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of ink Lear Power Plants Manuscript Completed: October 1980 Date Published: November 1980 U.S. Nuclear Regulatory Commission Federal Emergency Management Agency Washington, D.C. 2005 Washington D.C. 20472

n r- IC (1 CD MaO 0 - Z (

CD -4 0Cr

.r:,. t\

(D a

0*

0 a0 -

0 3C.c X

= M1)

CD 0

State and/or Local Offsite Class Licensee Actions Authority Actions CO03' CD c NOTIFICATION OF UNUSUAL EVENT 1. Promptly inform State and/or local 1. Provide fire or security offsite authorities of nature of assistance if requested CO Class Description unusual condition as soon as discovered* 2. Escalate to a more severe Unusual events are inprocess or class, if appropriate have occurred which indicate a 2. Augment on-shift resources as potential degradation of the level needed 3. Stand by until verbal of safety of the plant. No closeout releases of radioactive material 3. Assess and respond requiring offsite response or monitoring are expected unless 4. Escalate to a more severe class, further degradation of safety if appropriate systems occurs.

or Purpose

5. Close out with verbal summary to Purpose of offsite notification offsite authorities; followed by is to (1)assure that the first written summary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months step In any response later found to be necessary has been carried out, (2)bring the operating staff to a state of readiness, and (3)provide systematic handling of unusual events information and decisionmaking.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 30 Reference E.9

&' '~.May 1973

\ U.S. ^TOMIC ENEnGY COMMtS8SION S ~

REGULATC DIRECTORATE OF l;EGULATORY STANnARDS Vu G'IDE REGULATORY GUIDE 1.45 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS A. INTRODUCTION Leakago Separation Gencral Design Criterion 30, "Quality or Reactor. A limited amount of leakage is expected from the Coolant Pressure Boundary,' or Appendix A to !O RCPB and fr6m auxiliary systerns within the C7FR Part 50, "General Design Criteria for Nuclear containment such as from valve stem packing glands.

l'ower Planis,' reqepires that means be. provided for circulating pump 'shaft seals, and other equipment that detecting and. to the extent practical, Identifying the cannot practically be made 100%' leaktight. The reactor location of th: source of reactor coolant Ieakage. This vessel closure seals and safety and relief valves should guide describes acceptable methods of Implementing not leak significantly: however. if leakage occurs 'ia this requirement withi regard lo the selection of leakage these paths or via pump and valve seals. it should be detection systems for the reactor coolant pressure detectable and collectable and, to the extent practical.

boundayrv. This guide applies to lightwvater-cooled isolated from the containment :atmosphere so as not to reactors. The Advisory Committee on Reactor mask any potentially serious leak should it occur.

Safeguards has been consulted concerning t)'it guide These leakages arc known as "identified leakage" and and has concurred in the regulatory position. shou~ld be piped to tanks or sumps so that the flow rate can be established and monitored during plant B. DISCUSSION operation.

The safety significance of leaks from the reactor Uncollected leakage to the containment coolant pressure boundary (RCPB) can vary widely atmosphere from sources such as valve stem packing depending on the source of the leak as well as the glands and other sources that are not collected leakag: rate and duration. Therefore, the detection and increases the humid'ty of the containment. The monitoring of leakage or reactor coolant into the moisture removed from the atmosphere by air coolers containment area Is necessary. In most cases, methods together with any associated liquid leakage to thc for separating the Icakage from an identified source containment Is known as "unidentified leakage" and from the leakage from an unidentified source are should be collected in tanks or sumps where the flow necessary to provide prompt and quantitative rate can be established znd monitored during plant information to the operators to permit them to take operation. A small amount or unidentified leakage may immediate corrective action should a leak be be Impractical to eliminrte, but it should be reduced detrimental to the safety of the facility. Identified to a small flow rate, preferably less than one gallon per lakage is: (I) leakage into closed systems, such as minute (gpm), to permit the leakage detection systems pump scal or valve packin7 leaks that are captured, to detect positively and rapidly a small increase in flow flow metered, and conductLi to a sump or collecting rate. Thus a small unidentified leakage rate that is of tank, or (2) leakage into the containment atmosphere concern will not be masked by a larger acceptable from sources that arc bot.. specifically located and identified leakage rate.

known either not to interfere with the operation of unidentilicl leakage monitoring systcrrTs or not to be Substantial intersystem leakige from the RCPI to from a flaw in the RCI'B. Unidentified leakage is all other systems across passive barriers or valves is not other leakage. expecled. However, should such leakage occur, it may IJSAEC REGOLATORY GUIDES Cw-. of PAUth~hdg,$..h ,y be obt.mned by sqoetl bdimeiln th deehiont ll-i'.d to the U.S. Aitooc Eoe9, Ce.'-....O... We.tNtO6O. D.C. 20!.45.

Fteq.4e'oy Gwtd'e eweis.... gme0.o Ib. e.d m.h. eI~bt to th. P.bleC Altetqlo.. O4,.yte of A.Wtotye. Sv,s,.nb. C~en',.-'S *m i.Iemt fWi

~mted.& pwpblsto theAEC 1te9WtC"y sill.4le~l..tie~ifitcb~,C I" 00-o1. n he-. 904.' el-~Oge~d #od IN0.,,d bes-1. to the5n~t.'V the Co.-.'.on,,,'m sgaite to d.Ii-'see imch~iqe. ed by it- staff I. of the Conmeil.'., U. Aromic E-.v,' Co.,,.elso. Wwhlrp.D.C O. 20'..45.

4-1.teie' "wieclepeaties DI a0tt0Ietd to preed. 9.ldn.~ to ofdet5 Af- i~:chill,PeblicPmd g.v Su!.'.

emiheef. fe,Amte-y Gulde' Mi .-of seb.0tvieto, 1 r g.iWLos "eMomS.'tti w~thOwnmIt sio requi,.1. tAeIPod 0od WOWtt.'Ollf fee, fr"to home "e 0.4 i. The VASd.ee bUM~ Ie the keIoVb~ Ie boed dns the gold. evtl be oemp..ibte It lhey p.0.18 a butl Io, the UIdk'o ,q$ust to S. P~od.c doe bww" ., S.ttso O f 5 0pe't Of.te.,0 by theCoer.tnd..i I. P.., "Oeecie

2. A-chnt a TeV"Atoclon 7. WeCnelooti e s-

a. Fulle .sd U.tewletsFocI1lt0 0. Aetlt-aetR'.,-

,Atehed Skaid iAt b. ee-i-e P-oed.UflV. 0. PWW-M~. tO.,.odt 4. E. SHItiN rce.-tMcd P- - *... sln'ie, pteo. B. Ui..,a i. ymc~. tO. Oee..

ir. .

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 31 Reference E.9 not be detcc.jhlc through the above-mentioned Specific nuciCr power pla31t. Ilowever, since tlhe detec:kioz systems, and other aiarm and detection mctlods differ in sensitivity and response time. prudent miethods should be cmployed. For cxample. steam silection of detection methods :hould include sufficicnt generator liaka:c In pressurized water reactors (PWR s) systems to cssiare effective monitoring during perioads should hc monitored to detect tube or tube s!iect leaks. when some detection systems may he ineffective or inoNrabic. Somc Orf these systems should serve as early Acceptable Detection Methods alarmi systems signaling the operators thaI closer examlination of oller detection systems is necessary to Although monitoring of both identified and determinc thce extent of any corrective action that may unidcntified leakage is important, effectivc systems for be required.

detecting and locating unid.r.tined leakage are also nccdcd. The following pragraphis describe sonic Detector Semitivity acceptable detection methods.

It is essential that leakage detection systems have In addition to monitoring flow rate changes to tanks the capability to detect significant RCI'U degradation as and sumps for liquid collection, other methods should soon arter occurrence as practical to minimizc the Ie included to indicate when and where coolant is potential for a gross boundary failure. It is possible that relc-sed to' the containment atmosphere. For example, some cracks might develop and penetrate the RCPB wall.

these additional detection methods would Indicate exhibit very slow growth, and afford ample time for a andlor mcnitor changes in: sare and orderly plant shutdown after a leak is detected.

a. airborne particulate radioactivity, On the otiher hand, leakage such as that resulting froin

b. airborne gaseous radioactivity, stress-assisted corrosion in stainless steel or from a faw

c. containment atmosphere humidity, at a highl fatigue point in the RCPI would demand rapid

d. confainment atmosphere pressure and detection and probable, plant shutdown. Therefoie. an tclirmCraturc. early warning signal is necesstr-" to permit proper

e. condensate nlow rate from air coolers. evaluation of all unidentified leakage.

Since intersystem leakage does not re;.ase reactor Industry practice has shown that water flow rate coolant to the containment atmosphere, detection changes of froin 0.5 to 1.0 gpm can readily be detected methods rhould include monitoring of water in containment sumps by monitoring changes in sump radioactivity in the connected systems where the water level, in flow rate, or in the operating frequency of systems flows through the containment boundary and pumps. Sumps and tanks used to collect unidentified monitoring or airborne radioactivity where such systems leakage and air cooler condensate should be arc vented outside the containment boundary. Another instrumented to alarm for increases of from 0.5 to 1.0

mportant method of obtaining indications or gpm in the normal flow rates. This sensitivity would uncontrolled or undesirible intersystem flow would be provide an acceptable performance for detecting the use of awater Inventory balance designed to provide increases in unidentified liquid leakage by this method.

appropriate information such as abnormal water levels in tanks and abnormal watcr flow rates. An increase in humidity of the containment atmosphere would indicate release of water vapor to the Potential discharges from closed safety and relief containment. Dew point temperature measurements can valves are usually piped to tanks or water pools and be used to monitor humidi'y levels of the containment Considered part or identified leakage. Temperature atmosphere. A I" Increase in dew point is well within sensors In the discharec path or safety and relief valves the sensitivity range capability of available instruments.

or flow meters in the Icak-off lines would provide an Since thc humidity level is influenced by several factors.

acccptablc methjod or signaling small leakage from these a quantitative evaluation tf an .indicated leakage rate vilves. may be questionable and should be compared to observed increases in liquid flow from sumps and While the above-mentioned leakage detection condensate now from air coolers. Humidity level systems rclect tile present state of technology, it is monito:ing is considered most useful 'as an alarm or recognizeri that other detection methods may be indirect indicating device to alert the operator to a developed and used in order to obtain operating potential problem.

experience with them. Among such methods are sonic indicators and moisture sensitive tapes applied to RCPB Reactor coolant normally contains sources of component parts. Because or the potential importancC radiation which, when released to the containment. can of carly teak detection In the prevention of accidents. be detected by the monitoring systems. However.

continued improvements in leakage detection and reactor coolant radioactiviay should be low during initial locating techniques should be sought. reactor startup and for a few weeks thereafter until activated corrosion products have been formed and It is not necessary that all of the above-mcntioned fission products become available from failed fuel leakage detrection methods or systems be employed in a elemenis; during this period, radioactivity monitoring IA-42

. I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 32 Reference E.9 instrumients may bc of limited value in providing an Signal C-rrclation and Ca;iljration early warning of very smamll leaks in the RCP13.

Instrument sensitivilies Or IDI pCi/cc radioactivity for It is IrnrTssst1 to0he able to associate a si;enal or Dig 1airliculitc Illoniloritlr :nu of 10`6 pCi/vc indication of a dhiinge In tht normnal olerating radioactivity for sadiogas nuil itoring are practical fot conditions with a quantiltive leakage flow rate. Except these leakage detcctlion systems. .Radioactivity fr flow rate or level change measurements front tanks.

monitoring systems should be Included for every plant sumps. or pumps, signals from other leakage detection tespecially particulate activity monitoring) because of systems do not provide information readily con'e nible ilitii sensitivity and rmpid rcspons^ to Iczks from the to a conimon denomInator. Approximate rel2tionships RCI'I. converting these signals to units or water flow should be fdrmulated to assist the operator in interpreting signals.

Since upcrating conditions may influence some of the Air temnpcraturc and pressure monitoring methods conversion procedures, the procedures should be rYevised may also be used to infer 1VCPB leakage to the during such periods. To assure the continued reliability contlainmcnt. Containment temperature and pressure of the leakage detection systems, the equipment should Nuctvae slightly during plant operation, but a rise above comply with Paragraph 4.10 of IEEE Std. 279-1971.

the normally indicated range of values may Indicate "Criteria for Protection Systems for Nuclear Power RCPB leakage Into the containment. The accuracy and Generating Stations,"' for tests and calibration.

rclevance of lemperaturc and pressure measurements 's a function of containment frec volume and detector Seismic Qsatitication location. A1lrrrm signals from these instruments can be valuable in recognizing rapid and sizable energy releases Since nuclear power plants may be operating at the to the containment. time an earthquake occurs and may continue to operate after earthquakes, it is prudent to require the leakage While ihe concern about instrument sensitivity detection systems to function under the same applies to the lower range of service for which the conditions. If a seismic event comparable to a safe instruments are selected, the upper instrument range shutdown earthquake (SSE) occurs, it would be limits should be established to prevent e~xceeding the important for the operator to assess the condition within saturation limits of instruments. thus making them the containment quickly. The proper functioning of at useless as indicators of containment conditions. least one leakage detection. system would assist in evaluating the seriousness of the condition within the containment in the event leakage has developed in the Detaector Response 'lme RCPB. T11 airborne particulate radioactivity monitoring equipment has the desirable sensitivity to indicate RCPB The need to evalutIC the severity or an a1lrm or leakage, and it should be included for all plants.

indication is importlnt to the operators, and the ability Components for the airbome particulate radioactivity to compare with indications from other systems is equipment should be qualified to function through the necessary. The system response time should therefore be SSE.

included in the functional requirements for leakage detection systems. Except for the limitations during the C. REGULATORY POSITION initial few weeks or plant operation as discussed previously. ail detector systemi should respond to a one The source of reactor coolant leakage should be gpmnor its equivalent, leakage Increase In one hour or identifiable to the extent practical. Reactor coolant less. Multiple instrument locations in monitored areas pressure boundary leakage detection and collection should be utilized if necessary to assure that the systems should be selected and designed to include the transport delay time of the leakage effluent from its following-source to the detector or instrument location will yield an acceptable overall response time. A useful technique 1. Leakage to the primary reactor containment from in ldentiryir'^ the general location of a leakage area is the identified sources should be coilected or otherwise placing of several sensors within the containment area isolated so that:

and observing diffcrences in response from the sensors. a. the flow rates -.re monitored separately from and this technique should be used to satisfy this unidentified leakage. and requirement orGeneral Design Criterion 30. b. the total flow rate can be established and monitored.

In analyzing the sensitivity of leak detection systems using airborne particulatc or gaseous 2. Lakage to the primary reactor containment from radioalctivity, a realistic primary coolant radioactivity unidentified sources should be collected and the flow concenttation assumption should be used. The expected I Copies may be obtained tsom the Instiuit of tkecuicat values wed irn the plant environmental report would be and Etectronics Engineers. United Enginecring Center. 345 East acceptable. 471hStreet. New York.N.Y. 10017.

I .45.4

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2004-178 Attachment 3 Page 33 Reference E.9 rate ts..tlitoreid Willi all accuri.t oF Eie Fplloti per eiipilyed fuournidentriflcd Ieakage swiould he adequatc tn ilitfitc (glitl) oft llcttcl. detect a leakap. rate. or its equivalent, of one gpns in less I!rall on1e Ihoulr.

3. Al least thiet: sepaic t tldectl<ill 111cinet l1s slnnttld lie rI5il'tyett at d twt, or t ieseiittoidis shiuldi be ( I) stlt,, N.. The Icakagc ICctition systems thou!'! he capable of level aind hlow tii'm'ilosifig anI (2) ait hoit patitircul atIe l.trstuuitiig. lite i rlfinctilhlr rotlllwing tilsrtic events thiat radioactivitv 11alsllioling. lThC tInild mchiod mayV lie do 11t't leluilte 1lilatut hI d11lowil. The airblornic pfirticula I seletel r,,;It, 111c roll'svilt.: radt oaclivily molliloring system should refrolif

a. ,,,ollitrt'irt ..r codldelsatle lorw ra.,fititt rfunctional wlhtn siubiccittl li the KMI..

Ih. ,.of olitotingi airliorite gasecoits aditihia.livi Vy.

7. . Indlicauols Dnd alairus fisr cach leak age ticicction e I Itiuidi ty. ICtettiCpeitrC *tir pre;sUrC tillotu oritig ief mstiel shtould: le provided itll Chesain cont rol rt)oom,.

thc Cutilltaintilicilt :1lt sIspliete shiotidlehCeeiissrl rCil aS 1'rocediures for converting various indicutions In) a alarllls .., inidirect itwic-it biti ,r ic;lekage t, Itli coninton leakapc equivalent should bc available tl tlie cori lailitleil . osctralors. Tlie calibrationr of ilte indicators should account for neetctat indepcendeit vuriablies.

4. P'rovision s slhititld he made la mit itar systcms

.S;. Tl;C Jeakagc detection sysiems shovhd be equipped contitcaC to thIe RC'll etesugits of intersrse:zt Ilakage.

Methods shouild Include radioactivity nionitoritng anid with provisions to readily permit testing for operability indicators to show abnormal waler levels or flow in the atid ctlibration during plant operation.

afrccted arc;a. t) The technical sp1cciflcutions should include the limiting etonditi.oo for. ideistifled atid ,idcni tified

5. The sensitivity and rcspolise timc oreacit leakage lIakaicc atid address the availability of various types of detection sysici in regulatory posilionu 3. above ititrutsicits to assure adcquatc covrcratc at all titifes.

1.454 i.

L-2004-178, Proposed Change to Emergency Plan Table 3-1: Classification of Emergencies

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Dear Kr. Plunkett:

SUBJECT:

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2.0 BACKGROUND

SUMMARY

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L-2004-178, Proposed Change to Emergency Plan Table 3-1: Classification of Emergencies

Text

Dear Kr. Plunkett:

SUBJECT:

Enclosure:

2.0 BACKGROUND

SUMMARY

SUMMARY

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