Text

Transmittal of Technical Specifications Bases Revisions
	ML061040290
Person / Time
Site:	Robinson
Issue date:	04/06/2006
From:	Baucom C; Progress Energy Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RNP-RA/06-0033
	Download: ML061040290 (65)
	v • d • e

TS 5.5.14

¢ Progress Energy Serial: RNP-RA/06-0033 APR 06 2006 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/LICENSE NO. DPR-23 TRANSMITTAL OF TECHNICAL SPECIFICATIONS BASES REVISIONS Ladies and Gentlemen:

In accordance with Technical Specifications 5.5.14.d, Carolina Power and Light Company, also known as Progress Energy Carolinas, Inc., is transmitting revisions to the H. B. Robinson Steam Electri. Plant (HBRSEP), Unit No. 2, Technical Specifications Bases. The attachment to this letter provides Technical Specifications Bases pages for Revisions Number 26 through 29.

If you have any questions concerning this matter, please contact me at (843) 857-1253.

Sincerely, C. T. Baucom Supervisor - Licensing/Regulatory Programs CTB/cac Attachment c:

Dr. W. D. Travers, NRC, Region II NRC Resident Inspector, HBRSEP C. P. Patel, NRC, NRR Progress Energy Carolinas, Inc.

R 3binson Nuclear Plant 3381 West Entrance Road Hartsville, SC 29550

-1 (C) (

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/06-0033 64 Pages (including cover page)

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 TECHNICAL SPECIFICATIONS BASES PAGES FOR REVISIONS NUMBER 26 THROUGH 29

RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, the continued integrity of the RCS is ensured. According to 10 CFR 50 Proposed Appendix A (Ref. 1), GDC 9 "Reactor Coolant System Pressure Boundary" and GDC 34 "Reactor Coolant Pressure Boundary (RCPB) Rapid Propagation Failure Prevention," the reactor coolant pressure boundary design conditions are not to be exceeded during normal operations and transients. Also, in accordance with proposed GDC 33, "Reactor Coolant Pressure Boundary Capability," reactivity accidents, including rod ejection and inadvertent and sudden releases of energy to the coolant, do not result in damage to the RCPB.

The design pressure of the RCS is 2485 psig.

During normal operation and transients, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2).

To ensure system integrity, all RCS components were hydrostatically tested at 3110 psig, according to the ASME Code requirements prior to initial operation with no fuel in the core.

Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

Overpressurization of the RCS could result in a breach of the RCPB.

If such a breach occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100, "Reactor Site Criteria," or 10 CFR 50.67, "Accident Source Term." as applicable.

(continued)

HBRSEP Unit No. 2 B 2.0-6 Revision No. 26

RCS Pressure SL B 2.1.2 BASES (continued)

SAF:ETY LIMITS The maximum transient pressure allowed in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure.

The maximum transient pressure allowed in the RCS piping, valves, and fittings under USAS, Section B31.1 (Ref. 5) is 120% of design pressure.

The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is 2735 psig.

APPLICABILITY SL 2.1.2 applies in MODES 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODES due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head closure bolts are not fully tightened, making it unlikely that the RCS can be pressurized.

SAFETY LIMIT If the RCS pressure SL is violated when the reactor is in VIOLATIONS MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of the limits of 10 CFR 100, "Reactor Site Criteria,"

or 10 CFR 50.67, "Accident Source Term," as applicable.

The allowable Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of reducing power level to a MODE of operation where the potential for challenges to safety systems is minimized.

If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes. Exceeding the RCS pressure SL in MODE 3, 4, or 5 is more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes.

The action does not require reducing MODES, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.

(continued)

HBRSEP Unit No. 2 B 2.0-8 Revision No. 26

LCO Applicability B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.6 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

LCO

13.

0. 1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:

a.

Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and

b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.

There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable.

(Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.)

(continued)

HBRSEF' Unit No. 2 B 3.01 Revision No. 28 HBRSEP Unit No. 2 B 3.0-1 Revision No. 28

LCO Applicability B 3.0 BASES LCO ',.0.2 The second type of Required Action specifies the remedial (continued) measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Condition no longer exists. The individual LCO's ACTIONS specify the Required Actions where this is the case.

An example of this is in LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits."

The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally.

The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Alternatives that would not result in redundant equipment being inoperable should be used instead.

Doing so limits the time both subsystems/trains of a safety function are inoperable and limits the time other conditions exist wiich result in LCO 3.0.3 being entered.

Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable, and the ACTIONS Condition(s) are entered.

HBRSEP Unit No. 2 B 3.0-2 Revision No. 28

LCO Applicability B 3.0 BASES LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:

a. An associated Required Action and Completion Time is not met and no other Condition applies; or

b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit.

Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted: in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.

This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.

Upon entering LCO 3.0.3, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to prepare for an orderly shutdown before initiating a change in unit operation.

This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid.

The time limits specified to reach lower MODES of operation permit the shutdown to proceedin a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE.

This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.

A unit shutdown required in accordance with LCO 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs:

a. The LCO is now met.

(continued)

HBRSEP Unit No. 2 B 3.0-3 Revision No. 28

LCO Applicability B 3.0 BASES LCO 3.0.3

b. A Condition exists for which the Required Actions have (continued) now been performed.

c. ACTIONS exist that do not have expired Completion Tiimes.

These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.

The time limits of LCO 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the unit to be in MODE 5 when a shutdown is required during MODE 1 operation.

If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than alloded, however, the total allowable time to reach MODE 5, or other applicable MODE, is not reduced.

For example, if MODE 3 is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , then the time allowed for reaching MO)E 4 is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months , because the total time for reaching MO)E 4 is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.12, "Fuel Storage Pool Water Level." LCO 3.7.12 has an Applicability of "During movement of irradiated fuel assemblies in the fuel storage pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.12 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.12 of "Suspend movement of irradiated fuel assemblies in the fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3.

These exceptions are addressed in the individual Specifications.

HBRSEF Unit No. 2 B 3.0-4 Revision No. 28

LCO Applicability B 3.0 BASES LCO :3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in the Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

LCO 3.0.4.a allows entryinto a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

LCO3.0.4.b allows entryinto a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.

The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires risk impacts of maintenance activities to be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions.

(continued)

HBRSEP Unit No. 2 B 3.0-5 Revision No. 28

LCO Applicability B 3.0 BASES LCO 3.0.4 (continued)

These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability.

LCO 3.0.4.b may be used with single, or multiple systems and components unavailable.

NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.

The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified conditionin the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented.

The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above.

However, there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these system and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable.

LCO 3.0.4.c allows entryinto a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification.

(continued)

HBRSEF3 Unit No. 2 B 3.0-6 Revision No. 28

LCO Applicability B 3.0 BASES LCO 3.0.4 The risk assessments performed to justify the use of LCO 3.0.4.b (continued) usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., Containment Air Temperature, Containment Pressure, Moderator Temperature Coefficient), and may be applied to other Specifications based on NRC plant specific approval. LCO 3.0.4.c is applicable for LCO 3.4.16, RCS Specific Activity.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.

Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS.

(continued)

HBRSEP Unit No. 2 B 3.0-7 Revision No. 28

LCO Applicability B 3.0 BASES LCO 3.0.5 (continued)

The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of SRs to demonstrate:

a. The OPERABILITY of the equipment being returned to service; or

b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the allowed SRs.

This Specification does not provide time to perform any other preventive or corrective maintenance.

An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions and must be reopened to perform the SRs.

An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of an SR on another channel in the other trip system.

A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of an SR on another channel in the same trip system.

LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems that have an LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and RequiredActions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the unit is maintained in a safe condition are specified in the support system LCO's Required Actions.

These required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.

(continued)

HBRSEP Unit No. 2 B 3.0-8 Revision No. 28

LCO Applicability B 3.0 BASES LCO 3.0.6 When a support system is inoperable and there is an LCO specified (continued) for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability.

However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCOs' Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions.

However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action.

Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Specification 5.5.15, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine i f loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions.

The SFDP implements the requirements of LCO 3.0.6.

Cross train checks to identify a loss of safety function for those support systems that support multiple and redundant safety systems are required. The cross train check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained. If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

HBRSEP Unit No. 2 B 3.0-9 Revision No. 28

LCO Applicability B 3.0 BASES LCO :3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit.

These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions.

Test Exception LCO 3.1.8 allows specified Technical Specification (TS) requirements to be changed to permit performances of these special tests and operations. which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified continued)condition not directly associated with or required to be changed to perform the special test or operation will remain in effect.

The Applicability ofa Test Exception LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Test Exception LCOs is optional. A special operation may be performed either under the provisions of the appropriate Test Exception LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Test Exception LCO, the requirements of the Test Exception LCO shall be followed.

HBRSEFP Unit No. 2 B 3.0-10 Revision No. 28

l SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a. The systems or components are known to be inoperable, although still meeting the SRs; or

b. The requirements of the Surveillance(s) are known not to be met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a test exception are only applicable when the test exception is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR.

In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.

(continued)

HBRSEP Unit No. 2 B 3.0-11 Revision No. 28

SR Applicability B 3.0 BASES SR 3.0.1 Surveillances, including Surveillances invoked by Required (continued)

Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply.

Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2.

Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established.

In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performingits function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..." interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.

These exceptions are stated in the individual Specifications.

An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of "in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions."

(continued)

HBRSEP Unit No. 2 B 3.0-12 Revision No. 28

SR Applicability B 3.0 BASES SR 3.0.2 (continued)

The requirements of regulations take precedence over the TS.

The TS cannot in and of themselves extend a test interval specified in the regulations.

Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."

As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per... " basis. The 25%

extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time.

One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is less, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, (continued)

HBRSEP Unit No. 2 B 3.0-13 Revision No. 28

SR Applicability B 3.0 BASES SR 3.0.3 the time required to perform the Surveillance, the safety (continued) significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions or operational situations, is discovered not to have been performed when specified, SR 3.0.3 allows the full delay period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform the Surveillance.

SR 3.0.3 also provides a delay period for completion of Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence.

Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.

This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the (continued)

HBRSEF' Unit No. 2 B 3.0-14 Revision No. 28

SR Applicability B 3.0 BASES SR 3.0.4 Applicability for which these systems and components ensure (continued) safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with LCO 3.0.4. However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability. However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.

The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.

The precise requirements for performance of SRs are specified such that exceptions toSR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both.

(continued)

HBRSEFP Unit No. 2 B 3.0-15 Revision No. 28

SR Applicability B 3.0 BASES SR 3.0.4 (continued)

This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure requires entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met.

Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

HBRSEF' Unit No. 2 B 3.0-16 Revision No. 28

SDM 13 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure. The ejection of a rod also produces a time dependent redistribution of core power.

SDM satisfies Criterion 2 of the NRC Policy Statement.

Even though it is not directly observed from the control rDom, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

LCCI SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO.

For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed the limits of 10 CFR 100, "Reactor Site Criteria," (Ref. 4),

or 10 CFR 50.67, "Accident Source Term," (Ref. 7), as applicable.

For the boron dilution accident, if the LCO is violated, the minimum required time (Ref. 5) assumed for operator action to terminate dilution may no longer be applicable.

APF'LICABILITY In MODE 2 with keff < 1.0 and in MODES 3 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration."

In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6.

ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly.

A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components.

It is assumed that (continued)

HBRSEP Unit No. 2 B 3.1-4 Revision No. 26

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REOUUIREMENTS

c. RCS average temperature;

d. Fuel burnup based on previous critical boron concentration;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the verification.

REFERENCES

1.

UFSAR, Section 3.1.

2. UFSAR, Section 15.1.5.

3. UFSAR, Section 15.4.6.

4. 10 CFR 100.

5. UFSAR, Table 15.4.6-1.

6. UFSAR, Table 9.3.4-1.

7. 10 CFR 50.67.

HBRSEP Unit No. 2 B 3.1-6 Revision No. 26

RPS Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protection System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences (AOOs) and to assist the Engineered Safety Features (ESF)

Systems in mitigating accidents.

The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as specifying LCOs on other reactor system parameters and equipment performance.

The LSSS, defined in this specification as the Allowable Values, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (MBAs).

During AO0s, which are those events expected to occur one or more times during the unit life, the acceptable limits are:

1. The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling (DNB);

2.

Fuel centerline melt shall not occur; and

3. The RCS pressure SL of 2735 psig shall not be exceeded.

Operation within the SLs of Specification 2.0, "Safety Limits (SLs)," also maintains the above values and assures that offsite dose will be within the 10 CFR 50.67 or 10 CFR 100 criteria, as applicable, during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the unit life. The acceptable limit during accidents is that offsite dose shall be maintained within an acceptable fraction of 10 CFR 50.67 or 10 CFR 100 limits, as applicable. Different accident categories are allowed a (continued)

HBRSEP Unit No. 2 B 3.3-1 Revision No. 26

PAM Instrumentation B 3.3.3 BASES LCO

22.

PORV Position (Primary) (continued) from an emergency power source, to provide the direct (primary) means of valve position indication, from fully closed to fully open.

23.

PORV Block Valve Position (Primary)

Each PORV block valve operator and position qualified and powered to provide the direct position indication.

is equipped with a Limitorque indication which is seismically from an emergency power source, (primary) means of valve

24.

Safety Valve Position (Primary)

Each pressurizer safety valve is equipped with a single acoustical position indication system, which is seismically qualified and powered from an emergency power source, to provide the direct (primary) means of valve position indication. This system alarms in the control room to indicate an open safety valve.

APPLI(CABILITY The PAM instrumentation LCO is applicable in MODES 1, 2, and 3. These variables are related to the diagnosis and pre-planned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3.

In MODES 4, 5, and 6, unit conditions are such that the likelihood of an event that would require PAM instrumentation is low; therefore, the PAM instrumentation is not required to be OPERABLE in these MODES.

ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. The Completion Time(s) of the inoperable channel(s) of a Function are tracked separately for each Function starting from the time the Condition was entered for that Function.

(continued)

I HBRSEF' Unit No. 2 B 3.3-102 Revision No. 28

PAM Instrumentation B 3.3.3 BASES ACTIONS A.1 (continued)

Condition A applies when one or more Functions have one required channel that is inoperable.

Required Action A.1 requires restoring the inoperable channel to OPERABLE status within 30 days.

The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval. Condition A is modified by a Note that excludes certain PAM Functions since each of these Functions has only one channel. Condition D provides appropriate Required Actions for PAM Functions that have only one channel with that channel inoperable.

B.1 Condition B applies when the Required Action and associated Completion Time for Condition A are not met. This Required Action specifies initiation of actions in Specification 5.6.6, which requires a written report to be submitted to the NRC immediately. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.

(continued)

HBRSEP Unit No. 2 B 3.3-103 Revision No. 28

Remote Shutdown System B 3.3.4 BASES (continued)

APPLICABILITY The Remote Shutdown System LCO is applicable in MODES 1, 2, and 3. This is required so that the unit can be placed and maintained in MODE 3 for an extended period of time from a location other than the control room.

This LCO is not applicable in MODE 4, 5, or 6. In these MODES, the unit is already subcritical and in a condition of reduced RCS energy. Under these conditions, considerable time is available to restore necessary instrument control functions if control room instruments or controls become unavailable.

ACTIOfNS A Note has been added to the ACTIONS to clarify the application of Completion Time rules. Separate Condition entry is allowed for each Function listed on Table B 3.3.4-1. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function are tracked separately for each Function starting from the time the Condition was entered for that Function.

A.

1 Condition A addresses the situation where one or more required Functions of the Remote Shutdown System are inoperable. This includes any Function listed in Table B 3.3.4-1, as well as the control and transfer switches.

I Condition A is applicable when any combination of the control parameters and equipment controlled by listed control parameters are inoperable such that the required number is not met. Example: Condition A shall be entered in the situation where three Service Water pumps have inoperable remote shutdown controls and the fourth pump is inoperable for other reasons.

In this case, the required number of functions of one (1) would not be met.

The Required Action is to restore the required Function to OPERABLE status within 30 days.

The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.

(continued)

HBRSEP Unit No. 2 B 3.3-111 Revision No. 28

Remote Shutdown System B 3.3.4 Table B 3.3.4-1 (page 1 of 1)

Remote Shutdown System Instrumentation and Controls FUNCTION/INSTRUMENT REQUIRED OR CONTROL PARAMETER NUMBER OF FUNCTIONS

1. Reactivity Control

a. Source Range Neutron Flux 1

b. Reactor Trip Breaker Position(a) 1 per trip breaker

c. Manual Reactor Trip(a) 1 per trip breaker

2. Re3ctor Coolant System (RCS) Pressure Control

a. Pressurizer Pressure 1

b. Pressurizer Heater Controls 1

3. De:ay Heat Removal via Steam Generators (SGs)

a.

RCS Hot Leg Temperature Wide Range Loop A 1

b.

RCS Cold Leg Temperature Wide Range Loop A 1

c.

Motor Driven AFW Pump Controls 1

d. SG Pressure 1 per SG

e. SG Level (Wide Range) 1 per SG

f. Condensate Storage Tank Level 1

4. RCS Inventory Control

a. Pressurizer Level 1

b. Charging Pump Controls 1

c. Refuel Water Storage Tank Level 1

5. Support Functions

a. Component Cooling Water Pump Controls 1

b. Service Water Pump Controls 1

I I

(a)

This function is local indication and manual trip feature at the breaker Breakers and Reactor Trip Bypass Breakers that are racked in.

and applies to Reactor Trip HBRSEF' Unit No. 2 B 3.3-114 Revision No. 28

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES APPLICABLE SAFETY ANALYSES The containment ventilation isolation radiation monitors ensure closing of the ventilation isolation valves.

They are the primary means for automatically isolating containment in the event of a fuel handling accident during shutdown.

Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 100 (Ref. 1) or 10 CFR 50.67 (Ref.3) limits, as applicable. Due to radioactive decay, containment is only required to isolate during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 56 hours6.481481e-4 days 0.0156 hours 9.259259e-5 weeks 2.1308e-5 months ).

I The containment ventilation isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement.

LCO The LCO requirements ensure that the instrumentation necessary to initiate Containment Ventilation Isolation, listed in Table 3.3.6-1, is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate containment ventilation isolation at any time by using either of two pushbuttons in the control room.

Either pushbutton actuates both trains. This action will cause actuation of Phase A and Containment Ventilation Isolation automatic containment isolation valves. Containment Ventilation Isolation can also be initiated by the manual Containment Spray buttons.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one push button and the interconnecting wiring to the actuation logic cabinet.

2. Automatic Actuation Logic and Actuation Relays The LCO requires two trains of Automatic Actuation Logic and Actuation Relays to be OPERABLE. The (continued)

HBRSEFP Unit No. 2 B 3.3-122 Revision No. 26

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES (continued)

SURVEILLANCE SR 3.3.6.6 (continued)

REQUIREMENTS every 18 months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.).

The test also includes trip devices that provide actuation signals directly to the relay logic, bypassing the analog process control equipment. The SR is modified by a Note that excludes verification of setpoints during the TADOT.

The Functions tested have no setpoints associated with them.

The Frequency is based on the known reliability of the Function and the redundancy available, and has been shown to be acceptable through operating experience.

SR 3.3.6.7 A CHANNEL CALIBRATION is performed every 18 months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor.

The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.

REFERENCES

1. 10 CFR 100.11.

2. NUREG-1366, "Improvements to Technical Specification Surveillance Requirements," December, 1992.

3. 10 CFR 50.67 HBRSEP Unit No. 2 B 3.3-127 Revision No. 26

RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes.

These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips.

This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

Figures 3.4.3-1 and 3.4.3-2 contain P/T limit curves for heatup, cooldown, inservice leak and hydrostatic (ISLH) testing, and data for the maximum rate of change of reactor coolant temperature.

The following limitations apply to these figures:

a. Over the temperature range from COLD SHUTDOWN to hot operating conditions, the heatup rate shall not exceed 600F/hr in any one hour period.

b. Allowable combinations of pressure and temperature for a specific cooldown rate are below and to the right of the limit lines for that rate as shown in Figure 3.4.3-2.

This rate shall not exceed 1000F/hr in any one hour period.

The limit lines for cooling rates between those shown in Figure 3.4.3-2 may be obtained by interpolation.

c. Primary system hydrostatic leak tests may be performed as necessary provided the test temperature limitation as noted on Figure 3.4.3-1 is not violated. The maxiffum hydrostatic test pressure should remain below 2485 psig.

Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

(continued)

HBRSEIP Unit No. 2 B 3.4-9 Revision No. 27

RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued)

The ability of the large steel pressure vessel that contains the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).

The vessel is the component most subject to brittle failure, and the LCO limits apply mainly to the vessel. The limits dc not apply to the pressurizer, which has different design characteristics and operating functions.

10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for specific material fracture toughness requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G (Ref. 2).

The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy ferritic pressure vessel steels, such as the ASTM A302 Grade B parent material of the HBRSEP Unit No. 2 reactor pressure vessel, are well documented in the literature. Generally, low alloy ferritic materials show an increase in hardness and other strength properties and a decrease in ductility and impact toughness under certain conditions of irradiation. Accompanying a decrease in impact strength is an increase in the temperature for the transition from brittle to ductile fracture.

A method for guarding against fast fracture in reactor pressure vessels is presented in Reference 2. The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature, RTNDT.

RTNDT is defined as the greater of:

1. The drop weight nil-ductility transition temperature (NDTT, per ASTM E-208), or

2. The temperature 607F less than the 50 ft-lb (and 35 mils lateral expansion) temperature as determined from Charpy specimens oriented in a direction normal to the major working direction of the material.

(continued)

I HBRSEP Unit No. 2 B 3.4-10 Revision No. 27

RCS P/T Limits B 3.4.3 BASES BACKGROUND The RTNDT of a given material is used to index that material (continued) to a reference stress intensity curve (Kic curve), which appears in References 2 and 6. The Kic curve is a lower bound of the static fracture toughness results obtained from several heats of pressure vessel steel.

When a given material is indexed to the Kic curve, allowable stress intensity factors can be obtained for this material as a function of temperature.

Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

The Certified Material Test Reports (CMTR) for the original steam generators provided records of Charpy V-notch tests performed at +10-F. Acceptable Charpy V-notch tests of +10OF indicate RTNDT is at or below this temperature. The steam generator lower assemblies were replaced in 1984 and the material test results indicate the highest RTNDT is 60OF or below.

The ASME Code recommends that hydrostatic tests be performed at a temperature not lower than RTNDT plus 60'F, thus the pressurizing temperature for the steam generator shell is established at 1200F to provide protection against nonductile failure at the test pressure. The value of RTNDT, and in turn the operating limits of nuclear power plants, can be adjusted to account for the effects of radiation on reactor vessel material properties. The radiation embrittlement or changes in mechanical properties of a given reactor pressure vessel still can be monitored by a surveillance program such as the HBRSEP Unit No. 2 Reactor Vessel Radiation Surveillance Program (Ref. 3), where a surveillance capsule is periodically removed from the reactor pressure vessel and the encapsulated specimens tested. These data are comparedto data from pertinent radiation effects studies and an increase in the Charpy V-notch 30 ft-lb temperature (A RTNDT) due to irradiation is added to the original A RTNDT to adjust the RTNDT for radiation embrittlement. This adjusted RTNDT (RTNDT initial + A RTNDT) is utilized to index the material to the Kxc curve and in turn to set operating limits which take into account the effects of irradiation on the reactor pressure vessel materials.

Allowable pressure - temperature relationships for various heatup and cooldown rates are calculated using methods (Ref. 4) derived from Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code and Code Case N-641. The approach specifies that the allowable total stress intensity factor, Ki, at any time during heatup or cooldown cannot be greater than that shown on the Kic curve in Appendix G for the metal temperature at that time. Furthermore, (continued)

HBRSEII Unit No. 2 B 3.4-11 Revision No.

27

RCS P/T Limits B 3.4.3 BASES BACKGROUND the approach applies an explicit safety factor of 2.0 on the (continued) stress intensity factor induced by pressure gradients.

Following the generation of pressure - temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced in the following fashion. First, a composite curve is constructed based on a point-by-point comparison of the steady state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to be the lesser of the two values taken from the curves under consideration. The composite curve is then adjusted to allow for possible errors in the pressure and temperature sensing instruments.

The use of the composite curve is mandatory in setting heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling analysis switches from the outside diameter (OD) to the inside diameter (ID) location: and the pressure limit must, at all times, be based on the most conservative case. The cooldown analysis proceeds in the same fashion as that for heatup, with the exception that the controlling location is always at the ID position. The thermal gradients induced during cooldown tend to produce tensile stresses at the ID location, and compressive stresses at the OD position. Thus, the ID flaw is clearly the worst case.

As in the case of heatup, allowable pressure - temperature relationships are generated for both steady state and finite cooldown rate situations. Composite limit curves are then constructed for each cooldown rate of interest. Adjustments are made to account for pressure and temperature instrumentation error.

The criticalitylimit curve includes the Reference 1 requirement that it be > 400F above the heatup curve or the cool down curve, and not less than the minimum permissible temperature for ISLH testing. However, the criticality curve is not operationally limiting; a more restrictive limit exists in LCO 3.4.2, "RCS Minimum Temperature for Criticality."

(continued)

HBRSE1' Unit No. 2 B 3.4-12 Revision No. 27

RCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the limits of Figures 3.4.3-1 and 3.4.3-2 is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes.

This Frequency is considered reasonable in view of the control room indication available to monitor RCS status.

Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing.

No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES

1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G, 1995 Edition with 1996 Addenda.

3. Yanichko, S. E., "Carolina Power & Light Company, H. B.

Robinson Unit No. 2 Reactor Vessel Radiation Surveillance Program," Westinghouse Nuclear Energy Systems, WCAP-7373, January 1970.

4. Laubham, T. J., et al, "Analysis of Capsule X from the Carolina Power and Light H. B. Robinson Unit No. 2 Reactor (continued)

HBRSEP' Unit No. 2 B 3.4-17 Revision No. 27

RCS P/T Limits B 3.4.3 BASES REFERENCES Vessel Surveillance Program," WCAP-15805, March 2002.

(continued)

5. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

6. ASME Code Case N-641, "Alternative Pressure - Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division 1," January 17, 2000.

[Includes Code Cases N-588 and N-640.]

I HBRSEF' Unit No. 2 B 3.4-18 Revision No. 27

Pressurizer PORVs B 3.4.11 BASES LCO An OPERABLE block valve may be either open and capable of (continued) being closed, or closed. Isolation of an OPERABLE PORV does not render that PORV or block valve inoperable provided *the relief function of either the block valve or the PORV remains available with manual action.

Satisfying the LCO helps minimize challenges to fission product barriers.

APPLICABILITY In MODES 1, 2, and 3, the PORV and its block valve are required to be OPERABLE to limit the potential for a small break LOCA through the flow path.

The most likely cause for a PORV small break LOCA is a result of a pressure increase transient that causes the PORV to open.

Imbalances in the energy output of the core and heat removal by the secondary system can cause the RCS pressure to increase to the PORV opening setpoint. The most rapid increases will occur at the higher operating power and pressure conditions of MODES 1 and 2. The PORVs are also an alternative measure for manual actuation to mitigate a steam generator tube rupture event.

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

Therefore, the LCO is applicable in MODES 1, 2, and 3. The LCO is not applicable in MODES 4, 5, and 6 with the reactor vessel head in place when both pressure and core energy are decreased and the pressure surges become much less significant.

LCO 3.4.12 addresses the PORV requirements in these MODES.

ACTIONS A Note has been added to clarify that all pressurizer PORVs are treated as separate entities, each with separate Completion Times (i.e., the Completion Time is on a component basis).

(continued)

HBRSEF' Unit No. 2 B 3.4-55 Revision No. 28

Pressurizer PORVs B 3.4.11 BASES ACTIONS A.1 (continued)

PORVs may be inoperable and capable of being manually cycled (e.g., excessive seat leakage). In this condition, either the PORVs must be restored or the flow path isolated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The associated block valve is required to be closed, but power must be maintained to the associated block valve, since removal of power would render the block valve inoperable.

This permits operation of the plant until the next refueling outage (MODE 6) so that maintenance can be performed on the PORVs to eliminate the problem condition.

Quick access to the PORV for pressure control can be made when power remains on the closed block valve.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is based on plant operating experience that has shown that minor problems can be corrected or closure accomplished in this time period.

B.1, B.2, and B.3 If one PORV is inoperable and not capable of being manually cycled, it must be either restored, or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Times of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months are reasonable, based on challenges to the PORVs during this time period, and provide the operator adequate time to correct the situation. If the inoperable valve cannot be restored to OPERABLE status, it must be isolated within the specified time.

Because there is at least one PORV that remains OPERABLE, an additional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided to restore the inoperable PORV to OPERABLE status. If the PORV cannot be restored within this additional time, the plant must be brought to a MODE in which the LCO does not apply, as required by Condition D.

C.1 and C.2 If one block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the (continued)

HBRSEP Unit No. 2 B 3.4-56 Revision No. 28

LTOP System B 3.4.12 BASES (continued)

APPLICABILITY This LCO is applicable in MODE 4, MODE 5, and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T limits above 350'F. When the reactor vessel head is off, overpressurization cannot occur.

LCO 3.4.3 provides the operational P/T limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of the pressurizer safety valves that provide overpressure protection during MODES 1, 2, and 3.

Low temperature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time allows operator action to mitigate the event.

The Applicability is modified by a Note stating that accumulator isolation is only required when the accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves.

This Note permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable LTOP system. There is an increased risk associated with entering MODE 4 from MODE 5 with LTOP inoperable and the provisions of LCO 3.0.4.b, which allod entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.1 and B.1 With two or more SI pumps capable of injecting into the RCS, and all RCS cold leg temperatures 2 1750F and the requirements of LCO 3.4.12.b are not met (LCO 3.4.12.b requires the RCS to be depressurized and an RCS vent of

> 4.4 square inches established), or one or more SI pumps capable of injecting into the RCS with any cold leg temperature < 1750F and the requirements of LCO 3.4.12.b are not met, RCS overpressurization is possible.

(continued)

HBRSEPI Unit No. 2 B 3.4-69 Revision No. 28

LTOP System B 3.4.12 BASES ACTIONS A.1 and B.1 (continued)

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS from this condition.

C.1, D.1, and D.2 An improperly isolated accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

This is only required when the accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the P/T limit curves.

If isolation is needed and cannot be accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action D.1 and Required Action D.2 provide two options, either of which must be performed in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By increasing the RCS temperature to > 3500F, an accumulator pressure of 600 psig cannot exceed the LTOP limits if the accumulators are fully injected.

Depressurizing the accumulators below the LTOP limit also gives this protection.

The Completion Times are based on operating experience that these activities can be accomplished in these time periods and on engineering evaluations indicating that an event requiring LTOP is not likely in the allowed times.

E.1 In MODE 4, with one required PORV inoperable, the PORV must be restored to OPERABLE status within a Completion Time of 7 days.

Two PORVs are required to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time considers the facts that only one of the PORVs is required to mitigate an overpressure transient and that the likelihood of an active failure of the remaining valve path during this time period is very low.

F.1 The consequences of operational events that will overpressurize the RCS are more severe at lower temperature (Ref. 10).

Thus, with one of the two PORVs inoperable in MODE 5 or in MODE 6 with the head on, the Completion Time to restore two valves to OPERABLE status is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(continued)

HBRSEP' Unit No. 2 B 3.4-70 Revision No. 28

RCS operational Leakage B 3.4.13 BASES (continued)

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY ANALYSES do not address operational LEAKAGE. However, other operational LEAKAGE 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 0.3 gpm primary to secondary LEAKAGE as the initial condition.

Primary to secondary LEAKAGE is a 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.

For the SGTR, the activity released due to the 0.3 gpm primary to secondary LEAKAGE is relatively insignificant compared to the activity released via the ruptured tube.

The safety analysis for the SGTR accident assumes 0.3 gpm total primary to secondary LEAKAGE in all generators as an initial condition. After mixing in the secondary side, the activity is then released via the SG PORVs or safeties.

This release pathway continues until the SGs are isolated, which is relatively soon for the affected SG compared to the intact SGs. The dose consequences resulting from the SGTR accident are within the limits defined in 10 CFR 50.67.

The RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement.

LCCI 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.

(continued)

HBRSEP Unit No. 2 B 3.4-77 Revision No. 26

RCS operational Leakage B 3.4.13 BASES (continued)

LCO

b. Unidentified LEAKAGE

continued)

One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment atmosphere radiation monitoring systems, condensate measuring system, dewpoint monitoring equipment, 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.

C. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of identified 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.

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

Total primary to secondary LEAKAGE amounting to 0.3 gpm through all SGs produces acceptable offsite doses in the SGTR accident analysis. Violation of this LCO could exceed the offsite dose limits for this accident.

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 150 gallons per day limit on one SG produces acceptable dose consequences in the SGTR accident analysis.

Violation of this LCO could exceed the offsite dose limits for this accident. Primary to secondary LEAKAGE must be included in the total allowable limit for identified LEAKAGE.

(continued)

HBRSEP Unit No. 2 B 3.4-78 Revision No. 26

RCS Leakage Detection Instrumentation B 3.4.15 BASES (continued)

ACTIONS A.1 and A.2 With the required containment sump monitor inoperable, no other form of sampling can provide the equivalent information; however, the containment atmosphere radioactivity monitor will provide indications of changes in leakage. Together with the atmosphere monitor, the periodic surveillance for RCS water inventory balance, SR 3.4.13.1, must be performed at an increased frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to provide information that is adequate to detect leakage.

Restoration of the required sump monitor to OPERABLE status within a Completion Time of 30 days is required to regain the function after the monitor's failure. This time is acceptable, considering the Frequency and adequacy of the RCS water inventory balance required by Required Action A.1.

B.1.1, B.1.2, B.2.1, and B.2.2 With both gaseous and particulate containment atmosphere radioactivity monitoring instrumentation channels inoperable, alternative action is required.

Either grab samples of the containment atmosphere must be taken and analyzed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information.

With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor may be operated for up to 30 days to allow restoration of the required containment atmosphere radioactivity monitor.

Alternatively, continued operation is allowed if one fan cooler condensate flow rate monitor is OPERABLE, provided grab samples are taken every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect leakage. The 30 day Completion Time recognizes at least one other form of leakage detection is available.

(continued)

HBRSEP Unit No. 2 B 3.4-94 Revision No. 28

RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)

C.1 and C.2 With the required containment fan cooler condensate flow rate monitor inoperable, alternative action is again required.

Either SR 3.4.15.1 must be performed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information. Provided a CHANNEL CHECK is performed every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or a water inventory balance is performed every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , reactor operation may continue while awaiting restoration of a containment fan cooler condensate flow rate monitor to OPERABLE status.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval provides periodic information that is adequate to detect RCS LEAKAGE.

D.1 and D.2 With the required containment atmosphere radioactivity monitor and the required containment fan cooler condensate flow rate monitor inoperable, the only means of detecting leakage is the containment sump monitor. This Condition does not provide the required diverse means of leakage detection. The Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a reduced configuration for a lengthy time period.

(continued)

HBRSEP Unit No. 2 B 3.4-95 Revision No. 28

RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)

E.1 and E.2 If a Required Action of Condition A, B, C, or D cannot be met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least 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 .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

F.1 With all required monitors inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE REQUIREMENTS SR 3.4.15.1 SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor.

The check gives reasonable confidence that the channel is operating properly.

The Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based on instrument reliability and is reasonable for detecting off normal conditions.

SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner.

The test verifies the alarm setpoint and relative accuracy of the instrument string.

The Frequency of 92 days considers instrument reliability, and operating experience has shown that it is proper for detecting degradation.

(continued)

HBRSEF' Unit No. 2 B 3.4-96 Revision No. 28

RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.15.3, SR 3.4.15.4, and SR 3.4.15.5 These SRs require the performance of a CHANNEL CALIBRATION for each of the required RCS leakage detectioninstrumentation channels.

The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of 18 months is a typical refueling cycle and considers channel reliability.

Again, operating experience has proven that this Frequency is acceptable.

REFERENCES

1. UFSAR, Section 3.1.

2. UFSAR, Section 5.2.

HBRSEP Unit No. 2 B 3.4-97 Revision No. 28

RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and gross specific activity in the reactor coolant. The allowable levels are intended to limit the offsite dose to less than the limits of 10 CFR 100 (Ref. 1) or 10 CFR 50.67 (Ref. 3), as applicable, for analyzed accidents.

APPLICABLE SAFETY ANALYSES The LCO limits on the specific activity of the reactor coolant ensure that the resulting offsite doses will not exceed the 10 CFR 100 or 10 CFR 50.67 dose limits following an analyzed accident. The limiting accident analysis used in establishing the specified activity limits is the SGTR.

Other accidents, such as the Main Steam Line Break accident also use the limits from this LCO in the dose analysis. The SGTR dose analysis (Ref. 2) assumes the specific activity of the reactor coolant at the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 0.3 gpm.

The analysis assumes the specific activity of the secondary coolant at its limit of 0.1 pCi/gm DOSE EQUIVALENT I-131 from LCO 3.7.15, "Secondary Specific Activity."

(continued)

HBRSEP Unit No. 2 B 3.4-98 Revision No. 26

RCS Specific Activity B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued)

The SGTR event is assumed to be caused by the instantaneous rupture of a steam generator tube which relieves to the faulted steam generator. The primary consequence of this event is the release of radioactivity from the reactor coolant. The analysis also assumes a concurrent loss of power, from which the loss of circulating water through the condenser eventually results in the loss of condenser vacuum. Valves in the condenser bypass lines would automatically close to protect the condenser, thereby causing steam relief directly to the atmosphere from the steam generator PORVs or safety valves. This direct relief of activity from the ruptured tube would continue until the faulted steam generator is isolated. Additional releases due to primary to secondary LEAKAGE would continue from the SG PORVs or safety valves on the intact SGs until they were isolated.

Since no fuel failures are assumed to occur from the event, the specific activity at the LCO limit, and the amount of coolant released would determine the radioactivity that was released to the atmosphere.

The safety analysis shows the radiological consequences of an SGTR accident are within the dose limits of 10 CFR50.67.

Operation with iodine specific activity levels greater than the LCO limit is permissible for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , if the activity level does not exceed 60 pCi/gm.

The permissible iodine level of 60 pCi/gm or less is acceptable because of the low probability of a SGTR accident occurring during the established 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time limit. The occurrence of an SGTR accident at 60 pCi/gm would increase the calculated site boundary dose levels, but still be within 10 CFR 50.67 dose limits.

Limits on shielding personnel RCS specific activity also ensure the radiation design of the plant remains acceptable for plant radiation protection.

RCS specific activity satisfies Criterion 2 of the NRC Policy Statement.

(continued)

HBRSEP Unit No. 2 B 3.4-99 Revision No. 26

RCS Specific Activity B 3.4.16 BASES (Continued)

LCO The specific iodine activity is limited to 0.25 pCi/gm DOSE EQUIVALENT I-131, and the gross specific activity in the reactor coolant is limited to the number of pCi/gm equal to 100 divided by E (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limits on DOSE EQUIVALENT I-131 and gross specific activity ensure the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose to an individual at the site boundary during the DBA will be less than the allowed dose.

The SGTR accident analysis (Ref. 2) shows that the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary dose levels are within acceptable limits.

Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to site boundary doses that exceed the 10 CFR 50.67 dose limits.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature 2 5000F, operation within the LCO limits for DOSE EQUIVALENT I-131 and gross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values.

For operation in MODE 3 with RCS average temperature

< 5000F, and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

ACTIONS A Note permits the use of the provisions of LCO 3.0.4.c.

This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.

(continued)

HBRSEP Unit No. 2 B 3.4-100 Revision No. 28

RCS Specific Activity B 3.4.16 BASES ACTIONS A.1 and A.2 (continued)

With the DOSE EQUIVALENT I-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the Dose Equivalent I-131 concentration is

< 60 jCi/gm.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample.

Sampling is done to continue to provide a trend.

The DOSE EQUIVALENT I-131 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit violation resulted from normal iodine spiking.

B.1 With the gross specific activity in excess of the allowed limit, the unit must be placed in a MODE in which the requirement does not apply.

The change within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 and RCS average temperature < 500OF lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.

C.1 If a Required Action and the associated Completion Tine of Condition A is not met or if the DOSE EQUIVALENT I-131 is

> 60 pCi/gm, the reactor must be brought to MODE 3 with RCS average temperature < 500OF within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.

(continued)

HBRSEP Unit No. 2 B 3.4-101 Revision No. 26

RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS SR 3.4.16.3 (continued) 15 minutes, excluding iodines.

The Frequency of 184 days recognizes E does not change rapidly.

This SR has been modified by a Note that indicates the E determination is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES

1.

10 CFR 100.11.

2.

UFSAR, Section 15.6.3.

3.

10 CFR 50.67 HBRSEP Unit No. 2 B 3.4-103 Revision No. 26

ECCS - Shutdown B 3.5.3 BASES (continued)

LCO (continued)

In MODE 4, an ECCS train consists of a safety injection subsystem and an RHR subsystem aligned either for shutdogn cooling or for ECCS mode. An ECCS train is OPERABLE when the train consists of piping, instruments and controls to ensure an OPERABLE flow path capable of taking suction from the RWST to the SI pumps and transferring suction to the containment sump. The RHR subsystem is OPERABLE when the pump meets its IST program requirements.

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the ECCS pumps and their respective supply headers to each of the three cold leg injection nozzles. Manual alignment of the RHR subsystem would be necessary. In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs. The hot leg injection paths of the SI System, including valves, are not subject to the requirements of this specification.

APPLI(:ABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 4 with RCS temperature below 3500F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low.

Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops--

MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level."

A Note prohibits the application of LCO 3.0.4.b to an inoperable ECCS high head subsystem when entering MODE 4. There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS high head subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after (continued)

HBRSEP Unit No. 2 B 3.5-22 Revision No. 28

ECCS -Shutdown B 3.5.3 BASES (continued)

APPLICABILITY performance of a risk assessment addressing inoperable systems (continued) and components, should not be applied in this circumstance.

ACTIONS A.1 With no ECCS RHR subsystem OPERABLE, the plant is not prepared to respond to a loss of cool ant accident or to continue a cool down using the RHR pumps and heat exchangers. The Completion Time of immediately to initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity.

Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop. If no RHR loop is OPERABLE for this function, reactor decay heat must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous.

With both RHR pumps and heat exchangers inoperable, it would be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status.

B.1 With no ECCS high head subsystem OPERABLE, due to the inoperability of the safety injection train or flow path from the RWST, the plant is not prepared to provide high pressure response to Design Basis Events requiring SI. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time to restore at least one ECCS high head subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required.

C.1 When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated. Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators.

(continued)

HBRSEP Unit No. 2 B 3.5-23 Revision No. 28

IVSW System 13 3.6.8 BASES BA(:KGROUND a period of time following the accident.

Headers "B",

"C",

(continued) and "D" are automatic headers that are pressurized through one or both of two redundant, fail open, air operated valves arranged in parallel. A loss of power will cause these valves to fail open.

System operation is initiated by a Phase A containment isolation signal which accompanies any Safety Injection (SI) signal.

APPLICABLE The Design Basis Accident (DBA) that results in a release of SAFETY ANALYSES radioactive material within containment is a loss of coolant accident (LOCA). The analyses for the LOCA assumes the isolation of containment is completed and leakage fron containment is at a rate equivalent to the design leakage rate.

As part of the containment boundary, containment isolation valves function to support the leak tightness of containment. By maintaining this barrier, offsite dose calculations will be less than the limits of 10 CFR 1)30 or 10 CFR 50.67, as applicable, during a DBA (Ref. 2).

The IVSW System actuates on a containment isolation signal and functions to assure the actual leakage is no greater than the design value.

IVSW assures the effectiveness of certain isolation valves to limit containment leakage by pressurizing the affected containment penetration flow paths at a pressure 2 1.1 times Pa.

IVSW is designed to maintain this seal for at least 30 days. A single failure analysis shows the failure of any active component will not prevent fulfilling the design function of the system. By meeting these requirements, IVSW is considered a qualified seal system in accordance with 10 CFR 50, Appendix J (Ref. 3).

The Isolation Valve Seal Water System satisfies Criterion 3 of the NRC Policy Statement.

LCO During the DBA, the IVSW System must function to seal the associated penetration flow paths. OPERABILITY of the IVSW System is based on the its ability to seal selected containment penetration flow paths, at elevated pressure for at least 30 days assuming a single active failure. This requires that the IVSW tank be maintained with an adequate volume of water at sufficient pressure to provide the motive (continued)

HBRSEP Unit No. 2 B 3.6-50 Revision No. 26

MSIVs 3 3.7.2 BASES APPLICABLE continuous uncontrolled steam release from more than SAFETY ANALYSES one steam generator.

(continued)

b. A break outside of containment and upstream from the MSIVs is not a containment pressurization concern.

The uncontrolled blowdown of more than one steam generator must be prevented to limit the potential for uncontrolled RCS cooldown and positive reactivity addition.

Closure of the MSIVs isolates the break and limits the blowdown to a single steam generator.

c. A break downstream of the MSIVs will be isolated by the closure of the MSIVs.

d. Following a steam generator tube rupture, closure of the MSIVs isolates the ruptured steam generator from the intact steam generators to minimize radiological releases.

e. The MSIVs are also utilized during other events such as a feedwater line break. This event is less limiting so far as MSIV OPERABILITY is concerned.

The MSIVs satisfy Criterion 3 of the NRC Policy Statement.

LCO This LCO requires that three MSIVs in the steam lines be OPERABLE. The MSIVs are considered OPERABLE when the isolation times are within limits, and they close on an isolation actuation signal.

This LCO provides assurance that the MSIVs will perform their design safety function to mitigate the consequences of accidents that could result in offsite exposures comparable to the limits of 10 CFR 100 (Ref. 4), 10 CFR 50.67 (Ref. 6),

or the NRC staff approved licensing basis.

(continued)

HBRSEP Unit No. 2 B 3.7-10 Revision No. 26

MSIVs B 3.7.2 BASES SURVEILLANCE SR 3.7.2.1 (continued)

RE(UUIREMENTS containment analyses with the exception of closure of the MSIVs for a MSLB at 100% RTP, in which case MSIV closure in 2 seconds is assumed for MSIVs which close in the forward flow direction. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The MSIVs should not be tested at power, since even a part stroke exercise increases the risk of a valve closure when the unit is generating power. As the MSIVs are not tested at power, they are exempt from the ASME Code,Section XI (Ref. 5), requirements during operation in MODE 1 or 2.

The Frequency is in accordance with the Inservice Testing Program. The specified Frequency for valve closure time is based on the refueling cycle.

Operating experience has shown that these components usually pass the Surveillance when performed at the specified Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

This test is conducted in MODE 3 with the unit at operating temperature and pressure, as discussed in Reference 5 exercising requirements. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR.

This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

REFERENCES

1. UFSAR, Section 10.3.

2. UFSAR, Section 6.2.

3. UFSAR, Section 15.1.5.

4. 10 CFR 100.11.

5. ASME, Boiler and Pressure Vessel Code,Section XI.

6. 10 CFR 50.67 HBRSEP Unit No. 2 B 3.7-13 Revision No. 26

AFW System B 3.7.4 BASES APPLICABILITY generator secondary inventory, lost as the unit cools to (continued)

MODE 4 conditions.

In MODE 4 the AFW System may be used for heat removal via the steam generators.

In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an ino erable AFW train. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an AFW train inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

When an AFW pump is found to be inoperable, its associated flow path is also intrinsically inoperable. The "swing" flow path is not made inoperable by the inoperability of a single motor driven AFW pump.

Likewise, when a flow path is found inoperable in a manner that prevents flow through an AFW pump, the affected AFW pump is also intrinsically inoperable.

A.1 If one AFW pump or one or two AFW flow path(s) are inoperable, action must be taken to restore them to OPERABLE status within 7 days.

The 7 day Completion Time is reasonable, based upon the following:

a. With any single AFW pump or one or two flow path(s) inoperable, redundant capability to inject flow into at least one steam generator exists.

b. With the AFW "swing" injection flow path inoperable concurrent with another motor driven flow path inoperable, redundant capability to inject flow into at least one steam generator exists.

Other combinations of inoperable AFW flow paths and pumps result in entry into either Condition B or Condition C.

(continued)

HBRSE1' Unit No. 2 B 3.7-25 Revision No. 28

AFW System B 3.7.4 BASES ACTIONS A.1 (continued)

The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of Conditions to be inoperable during any continuous failure to meet this LCO.

The 8 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The AND connector between 7 days and 8 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met.

(continued)

HBRSEP Unit No. 2 B 3.7-25a Revision No. 28

Secondary Specific Activity B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Reactor Coolant System (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives and, thus, indicates current conditions. During transients, iodine spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

I A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

The Steam Generator Tube Rupture (SGTR) and the Main Steam Line Break (MSLB) (Ref. 1) result in the release of activity contained in the secondary side.

With the specified activity limit, the resultant offsite doses will be less than the limits of 10 CFR 50.67.

(continued)

HBRSEP Unit No. 2 B 3.7-77 Revision No. 26

Secondary Specific Activity B 3.7.15 BASES (continued)

APPLICABLE SAFETY ANALYSES The accident analyses of the SGTR and the MSLB, as discussed in Reference 1, assume the initial secondary coolant specific activity to be at the LCO concentration of 0.10 pCi/gm DOSE EQUIVALENT I-131.

This assumption is used in the analyses for determining the radiological consequences of the postulated accidents.

The accident analyses, based on this and other assumptions, shows that the radiological consequences of the accidents do not exceed the limits of 10 CFR 50.67 (Ref. 2).

With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the Main Stean Safety Valves (MSSVs) and steam generator power operated relief valves (PORVs). The Auxiliary Feedwater System supplies the necessary makeup to the steam generators.

Steaming via the unaffected steam generators continues until the reactor coolant temperature and pressure have decreased sufficiently for the Residual Heat Removal System to complete the cooldown.

Secondary specific activity limits satisfy Criterion 2 of the NRC Policy Statement.

LCC As indicated in the Applicable Safety Analyses, the specific activity of the secondary coolant is required to be

< 0.10 pCi/gm DOSE EQUIVALENT I-131 to limit the radiological consequences of a Design Basis Accident (DBA) to less than the required limit (Ref. 2).

(continued)

HBRSEP Unit No. 2 B 3.7-78 Revision No. 26

Secondary Specific Activity B 3.7.15 BASES SURVEILLANCE SR 3.7.15.1 (continued)

REOUUIREMENTS isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. The 31 day Frequency is based on the detection of increasing trends of the level of DOSE EQUIVALENT I-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.

REFERENCES

1. UFSAR, Chapter 15.

2. 10 CFR 50.67.

I HBRSEP Unit No. 2 B 3.7-80 Revision No. 26

AC Sources-Operating B 3.8.1 BASES LCO

d. High crankcase pressure (continued)

e. Start failure - governor shutdown Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY.

The AC sources in one train are separate and independent (to the extent possible) of the AC sources in the other train.

For the DGs, separation and independence are complete.

APPLICABILITY The AC sources are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AQOs or abnormal transients; and

b. Adequate core cooling is provided and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

The AC power requirements for MODES 5 and 6 are covered in LCO 3.8.2, "AC Sources-Shutdown."

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable DG. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable DG and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, sh:uld not be applied in this circumstance.

A.1 Required Action A.1, which only applies if the train cannot be powered from an offsite source, is intended to provide assurance that an event coincident with a single failure of the associated DG will not result in a complete loss of safety function of critical redundant required features. These features are powered from the redundant AC electrical poser train. This includes motor driven auxiliary feedwater pumps.

Single train systems, such as turbine driven auxiliary feedwater pumps, may not be included.

(continued)

HBRSEP Unit No. 2 B 3.8-5 Revision No. 28

AC Sources-Operating B 3.8.1 BASES ACTIONS A.1 (continued)

The Completion Time for inoperability of the offsite source is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The rationale for the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is that Regulatory Guide 1.93 (Ref. 9) allows a Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for two required offsite circuits inoperable when two offsite sources are incorporated into the design, based upon the assumption that two complete safety trains are OPERABLE. When no offsite sources are OPERABLE, this assumption is not the case, and a shorter Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is appropriate.

A.2 Operation may continue in Condition A for a period that should not exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With the offsite circuit inoperable, the reliability of the AC power system is degraded, with attendant potential for a challenge to the unit safety systems. In this Condition, however, the remaining OPERABLE DGs are adequate to supply electrical power to the onsite Distribution System.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.

The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of required AC power sources to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition A is entered while, for instance, a DG is inoperable and that DG is subsequently returned OPERABLE, the LCO may already have been not met for up to 7 days. This could lead to a total of 8 days, since initial failure to meet the LCO, to restore the offsite circuit. At this time, a DG could again become inoperable, the circuit restored OPERABLE, and an additional 7 days (for a total of 15 days) allowed prior to complete restoration of the LCO. The 8 day Completion Time provides a limit on the time allowed in a specified condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and B are entered concurrently. The "AND" connector between the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 8 day Completion Times means that both Completion Times apply simultaneously, and the more restrictive Completion Time must be met.

(continued)

HBRSEP Unit No. 2 B 3.8-6 Revision No. 28

DC Sources-Operating B 3.8.4 BASES SURVE:LLANCE SR 3.8.4.3 (continued)

REQUIREMENTS terminal connection.

The removal of visible corrosion is a preventive maintenance SR.

The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.3.

The 18 month frequency is based on engineering judgement taking into consideration the likelihood of a change in component or system status.

SR 3.8.4.4 This SR requires that each battery charger be capable of supplying 300 amps and 125 V for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . These current and voltage requirements are based on the design capacity of the chargers. The battery charger supply is based on normal DC loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state.

The minimum required amperes and duration ensures that these requirements can be satisfied.

The Surveillance Frequency is acceptable, given the other administrative controls existing to ensure adequate charger performance during these 18 month intervals.

In addition, this Frequency is intended to be consistent with expected fuel cycle lengths.

SR 3.8.4.5 A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length should correspond to the design duty cycle requirements.

This SR is modified by two Notes.

Note 1 allows the performance of a modified performance discharge test in lieu of a service test.

(continued)

HBRSEP Unit No. 2 B 3.8-43 Revision No. 29

DC Sources-Operaiting B 2.8.4 BASES SURVEILLANCE SR 3.8.4.5 (continued)

REQUIREMENTS The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems.

SR 3.8.4.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test.

The test is intended to determine overall battery degradation due to age and usage.

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.6; however, only the modified erformance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.5.

A modified discharge test is a test of the battery capacity.

and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test.

It may consist of just two rates: for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

(continued)

HBRSEP Unit No. 2 B 3.8-44 Revision No. 29

DC Sources-Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.6 (continued)

REQUIREMENTS The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 5). This reference recommends that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. An acceptance criterion of 80% of rated capacity is applicable to the "A" battery only. An acceptance criterion of 91% is applicable to the "B" battery since the battery's capacity is not as great.

The Surveillance Frequency for this test is normally 60 months. If the battery shows degradation, or if the battery has reached 85% of its expected life with capacity

< 100% of manufacturer's rating, the Surveillance Frequency is reduced to 12 months. Degradation is indicated, according to IEEE-450 (Ref. 5), when the battery capacity drops by more than 10% relative to its average on the previous performance tests or when it is > 10% below the manufacturer's rating. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity > 100% of the manufacturer's ratings. These Frequencies are consistent with the recommendations in IEEE-450 (Ref. 5).

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems.

REFERENCES

1. UFSAR Section 3.1.

2. UFSAR, Chapter 8.

3. UFSAR, Chapter 6.

4. UFSAR, Chapter 15.

5. IEEE-450-1995.

I HBRSEP Unit No. 2 B 3.8-45 Revision No. 29

RNP-RA/06-0033, Transmittal of Technical Specifications Bases Revisions

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