ML14191A059

Application to Revise Technical Specifications to Adopt TSTF-523, Generic Letter 2008-01, Managing Gas Accumulation, Revision 2, Using the Consolidated Line Item Improvement Process

ML14191A059
Person / Time
Site:	Three Mile Island
Issue date:	07/10/2014
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMl-14-075, TSTF-523
Download: ML14191A059 (20)

Text

10 TMl-14-075 July 10, Nuclear Regulatory Commission ATTN: Document Control Desk Washington, 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed License No. DPR-50

Subject:

Application to Revise Technical Specifications to Adopt TSTF-523, "Generic Letter 2008-01, Managing Accumulation," Revision 2, using the Consolidated Line Item Improvement Process

References:

1. TSTF-523, "Generic Letter 2008-01, Managing Gas Accumulation,"

Revision 2, dated February 20, 2013

2. Notice of Availability of the "TSTF-523, 'Generic Letter 2008-01, Managing Gas Accumulation,' Using the Consolidated Line Item Improvement Process," dated January 15, 2014

3. Letter from K. R. Jury (Exelon Generation Company, LLC/AmerGen Energy Company, LLC) to U.S. NRC, "Nine-Month Response to Generic Letter 2008-01," dated October 14, 2008 Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (EGC) is submitting a request for an amendment to the Technical Specifications for Three Mile Island Nuclear Station, Unit 1 (TMI, Unit 1).

The proposed amendment would modify Technical Specification requirements to address Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," as described in TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation." TMI, Unit 1 committed to evaluate the resolution of the Technical Specification issues with respect to the elements contained in the TSTF, and submit a license amendment request, if deemed necessary based on the evaluation, within 180 days following NRC approval of the TSTF (Reference 3). This submittal satisfies the commitment.

Attachment 1 provides a description and assessment of the proposed change. Attachment 2 provides the existing Technical Specification pages and Technical Specification Bases pages marked up to show the proposed changes. Changes to the existing Technical Specification Bases, consistent with the technical and regulatory analyses, will be implemented under the Technical Specification Bases Control Program. They are provided in Attachment 2 for information only. Attachment 3 provides Table 1, which identifies the affected Standard Technical Specifications and provides the equivalent TMI, Unit 1 Technical Specifications.

U.S. Nuclear Regulatory July 10, 2

"'""'lit"'lrt~:..c- have and approved by TMI, Unit 1's Plant Operations

.l"\,....,m.,,. and approved by the Nuclear 00 Review Board in accordance with the the Quality l;J\.f\.ll;Ji;)~i;) approval of the proposed amendment by July 10, 2015. approved, the amendment shall be implemented within 1 days.

There are no regulatory commitments contained in this In accordance with 10 50.91, "Notice for public comment; State consultation," paragraph (b), is notifying the Commonwealth of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this letter, please contact Wendy E. Croft at (610) 765-5726.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 10th day of July 2014.

Respectfully, David P. Helker Manager - Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1. Description and Assessment

2. Proposed Technical Specification and Technical Specification Bases (For Information Only) Changes (Mark-Up)

3. Equivalent TMI, Unit 1 Technical Specification Nomenclature cc: USNRC Region I, Regional Administrator USNRC Project Manager, TMI, Unit 1 USNRC Senior Resident Inspector, TMI, Unit 1 Director, Bureau of Radiation Protection, PA Department of Environmental Resources Chairman, Board of County Commissioners, Dauphin County, PA Chairman, Board of Supervisors, Londonderry Township, PA R. R. Janati, Commonwealth of Pennsylvania 1.0 Surveillance Requirements to verify that the system locations susceptible to are sufficiently filled with water and to provide allowances which permit performance of the verification. changes are being made to address the concerns discussed in Generic Letter 2008- 01, "Managing Accumulation in Emergency Core Cooling, Decay Heat Removal, and 11 Containment Systems.

The proposed amendment is consistent with TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Accumulation."

2.0 2.1 Exelon Generation Company, LLC (EGC) has reviewed the model safety evaluation dated December 2013 as part of the Federal Register Notice of Availability. This review included a review of the NRC's evaluation, as well as the information provided in TSTF-523, Revision 2. As described in the subsequent paragraphs, has concluded that the justifications presented in the TSTF-523, Revision 2 proposal and the model safety evaluation prepared by the NRC are applicable to Three Mile Island Nuclear Station, Unit 1 (TMI, Unit 1) and justify this amendment for incorporation of the changes to the plant Technical Specifications (TSs).

The model safety evaluation discusses the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). TMI, Unit 1 is not licensed to the 10 CFR 50, Appendix A, GDC. TMI, Unit 1's Updated Final Safety Analysis Report (UFSAR),

Section 1.4, "Principal Architectural and Design Criteria," provides an assessment against the proposed Atomic Energy Commission GDC published in 1967. A review has determined that the plant-specific requirements are sufficiently similar to the Appendix A GDC as related to the proposed change. Therefore, the proposed change is applicable to TMI, Unit 1.

2.2 Optional Changes and Variations EGC is not proposing any significant variations or deviations from the TS changes described in TSTF- 523, Revision 2, or the applicable parts of the NRC's model safety evaluation dated December 23, 2013. EGC is noting the following minor variations from the TS changes described in TSTF-523, Revision 2.

• The TMI, Unit 1 TS utilize different formatting, numbering, and titles than the Standard Technical Specifications (STS) on which TSTF-523, Revision 2 is based. Table 1 in Attachment 3 identifies the affected STS number, title, and section (i.e., Surveillance Requirements, LCO Bases, etc.)

and provides the equivalent TMI, Unit 1 TS number, title, and section.

• EGC modified the proposed SR for the Emergency Core Cooling System (ECCS) to align with TMI, Unit 1 TS. The B&W STSs ECCS SRs encompass all of the ECCS systems. The TMI, Unit 1 TS has separate SRs for the three individuals ECCS systems (High Pressure Injection, Low Pressure Injection, and Core Flooding). To align with the TMI, Unit 1 TS the proposed ECCS SRs were modified to state the ECCS system name in place of the term ECCS. For example, the SRs state, " ... verify [Core Flooding] locations susceptible to gas accumulation are sufficiently filled with water."

Description and Assessment Page 2 of 4

• excluded the proposed SR Note on the ECCS that require verification that manual are in the correct position. The Note allowed the to not be met for system vent flow paths while performing the proposed accumulation TMI, Unit 1 does not have equivalent SRs that require verification that manual are in the correct position.

l"lor,o""'"t::l the Note and the corresponding text were excluded.

• modified the proposed SR Note for the Decay Heat Removal (OHR) System that states the SR does not have to be performed "until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 4." The Note was modified to state! "until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is less than or equal to 250 degrees F."

modified text corresponds with the Applicability statement in TMI, Unit 1 TS 4.9.2 and meets the intent of the Note as described in Revision 2, Section 2.1, "Revise or Add Surveillance Requirements."

has reviewed these changes and determined that they are administrative and do not affect the applicability of TSTF-523, Revision 2 to the TMI, Unit 1 TS.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination Exelon Generation Company, LLC (EGC) requests adoption of TSTF-523, Revision 2, "Generic Letter 2008-01, Managing Gas Accumulation," which is an approved change to the Standard Technical Specifications (STS), into the Three Mile Island Nuclear Station, Unit 1 Technical Specifications (TS).

The proposed change adds Surveillance Requirements to verify that the system locations susceptible to gas accumulation are sufficiently filled with water and to provide allowances which permit performance of the verification.

EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change adds Surveillance Requirements (SRs) that require verification that the Emergency Core Cooling System (ECCS), the Decay Heat Removal (OHR) System, and the Reactor Building Spray (RB Spray) System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. Gas accumulation in the subject systems is not an initiator of any accident previously evaluated. As a result, the probability of any accident previously evaluated is not significantly increased. The proposed SRs ensure that the subject systems continue to be capable of performing their assumed safety function and are not rendered inoperable due to gas accumulation. Thus, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. possibility of a new or different kind of accident from any adds that require verification that the the OHR, and the RB

,\/.::l!To*m are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change does not involve a nrn,rc::ir~I alteration of the plant no new or different type of equipment will be installed) or a

"'"'~"'"'.,1 '° in normal plant operation. In addition, the proposed change does not impose any new or different requirements that could initiate an accident. The proposed change not alter assumptions made in the safety analysis and is consistent with the safety analysis assumptions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change adds SRs that require verification that the ECCS, the OHR, and the RB Spray System are not rendered inoperable due to accumulated gas and to provide allowances which permit performance of the revised verification. The proposed change adds new requirements to manage gas accumulation in order to ensure the subject systems are capable of performing their assumed safety functions. The proposed SRs are more comprehensive than the current SRs and will ensure that the assumptions of the safety analysis are protected. The proposed change does not adversely affect any current plant safety margins or the reliability of the equipment assumed in the safety analysis. Therefore, there are no changes being made to any safety analysis assumptions, safety limits or limiting safety system settings that would adversely affect plant safety as a result of the proposed change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

EVALUATION ATTACHMENT 2 Proposed Technical Specification and Technical Specification Bases (For Information Only)

Changes (Mark-Up)

Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 Page 3-2 Page 3-23 Page 3-26d Page 4-41 Page 4-42 Page 4-43 Page 4-44 Page 4-52a

nn.:::.r~ 1* 1 nn with one idle RC pump in each loop has been imposed since the not been calculated in with the Final Acceptance 11 0 r mo.n1'c SPE~cm1ca11v for this mode reactor operation. A time period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is RC pump in each loop to effect repairs of the idle pump(s) and

"" ............................... combination of operating RC pumps. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for this this mode is expected to have considerable margin for the the likelihood of a LOCA within the 24-hour period is A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is by dilution with makeup water. Either pump will provide mixing which will caused by dilute coolant reaching the reactor. One 0

" " ..... \l~I pump will circulate the equivalent of the reactor coolant system volume in The removal system suction piping is designed for 300°F and 370 psig; thus, the system can remove decay heat when the reactor coolant system is below this temperature (References 1, 2, and 3).

Both steam generators must have tube integrity before heatup of the Reactor Coolant System to insure system integrity against leakage under normal and transient conditions. Only one steam generator is required for decay heat removal purposes. Refer to Section 3.1.6.3 for allowable primary-to-secondary leakage. Refer to Section 4.19 for Bases for Steam Generator tube integrity.

One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for a rod withdrawal or feedwater line break accidents (Reference 4). The pressurizer code safety valve lift set point shall be set at 2500 psig +/-1 % allowance for error. Surveillance requirements are specified in the lnservice Testing Program. Pressurizer code safety valve setpoint drift of up to 3% is acceptable in accordance with the assumptions of the TMl-1 safety analysis (Reference 5).

(1) UFSAR, Tables 9.5-1 and 9.5-2 (2) UFSAR, Sections 4.2.5.1 and 9.5 - "Decay Heat Removal" (3) UFSAR, Section 4.2.5.4 - "Secondary System" (4) UFSAR, Section 4.3.10.4 - "System Minimum Operational Components" (5) UFSAR, Section 4.3.7 - "Overpressure Protection" 3-2 Amendment No. 47 (12/22/78), 157, 222, 261, 200

3.3.2. 1 If the CFT boron concentration is outside of or if the TSP baskets contain amounts of TSP outside the limits in 3.3. 1 restore the to status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If C?\/l::l r1:.m is not restored to meet the 1 3.3.1 within 72 the

..a,,. ....+...... shall be in a HOT condition within six hours.

3.3.3 Both CFTs shall be OPERABLE at all times.

b. Both the motor op1::ratea valves associated with the CFTs shall be open at all times.

c. cooling fan and associated r't'\l"llll"'ll"I unit shall be ru::ir*rru++.::.rl to be out-of-3.3.4 1 n 1r 10r 1r'r1 maintenance on any of the components, the aut>nc:ate (redundant) component

\/Qf"ITIQl"I to be OPERABLE.

Spt::c1t1ca1t1on 3.3.1 assure that, before the reactor can be made critical, adequate 1i;;;a1Lu11::::i;::i are operable. Two safeguards makeup pumps, two decay heat removal pumps and two decay heat removal coolers {along with their respective cooling water systems components) are However, only one of each is necessary to supply emergency coolant to the

.___ reactor in the of a loss-of-coolant accident. Both CFTs are required because a single CFT has

• sufficient inventory to reflood the core for hot and cold line breaks (Reference 1) .

The of the borated water storage tank (BWST) as part of the ECCS ensures that a sufficient supply of water is available for injection by the ECCS in the event of a LOCA (Reference 2).

The limits on BWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain at least one percent subcritical following a Loss-of-Coolant Accident (LOCA).

The contained water volume limit of 350, 000 gallons includes an allowance for water not usable because of tank discharge location and sump recirculation switchover setpoint. Redundant heaters maintain the borated water supply at a temperature greater than 40°F.

The Reactor Building emergency sump pH control system ensures a sump pH between 7.3 and 8.0 during the recirculation phase of a postulated LOCA. A minimum pH level of 7.3 is required to reduce the potential for chloride induced stress corrosion cracking of austenitic stainless steel and assure the retention of elemental iodine in the recirculating fluid. A maximum pH value of 8.0 minimizes the 3-23 Amendment No. 149, 157,165, 178, 227, 229, 263 :27S

Other plant show that two (2) MSSVs on either OTSG are more than sufficient to relieve reactor pump and stored energy when the reactor is below 5% full power operation but had been by 1 KJK for at one hour subsequent to power full power. 3.1.1 both OTSGs shall have tube integrity whenever reactor coolant temperature is above 200 degrees F. This assures that four (4) MSSVs are available for redundancy. During power operations at full power or above, if MSSVs are inoperable, the power level must be reduced, as stated in Specification 3.4.1 that the remaining MSSVs can prevent overpressure on a turbine trip.

The minimum amount of water in the CSTs required by Specification 3.4.1 1.c, provides at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of OHR with steam being discharged to the atmosphere. This provides adequate time to align alternate water sources for RCS cooldown. After cooling to 250 degrees F, the OHR System is used to achieve further cooling.

When the RCS temperature is below 250 degrees F, a single OHR String (Loop), or single OTSG with an EFW Pump and a flowpath capable of supporting natural circulation is sufficient to provide removal of decay heat at all times following the cooldown to 250 degrees F. The OHR String (Loop) redundancy required by Specification 3.4.2.1 is achieved with independent active components capable of maintaining the RCS subcooled. A single OHR flowpath with redundant active components is sufficient to meet the requirements of Specifications 3.4.2.1.a and 3.4.2.1.b. The requirement to maintain two operable means of OHR ensures that a single active failure does not result in a complete loss of OHR capability. The requirement to keep a OHR Loop in operation as necessary to maintain the RCS subcooled at the core outlet provides the to ensure that steam conditions which could inhibit core cooling do not occur.

With the Reactor Vessel head removed and 23 feet of water above the Reactor Vessel flange, a large heat sink is available for core cooling. In this condition, only one OHR Loop is required to be operable because the volume of water above the Reactor Vessel flange provides a large heat sink which would allow sufficient time to recover active OHR means.

Following extensive outages or major core off-loading, the decay heat generation being removed from the Reactor Vessel is so low that ambient losses are sufficient to maintain core cooling and no other means of heat removal is required. The system is passive and requires no redundant or diverse backup system. Decay heat generation is calculated in accordance with ANSI 5.1-1979 to determine when this situation exists (Reference 4).

REFERENCES (1) UFSAR, Table 6.1 ECCS "Single Failure Analysis" (2) UFSAR, Section 9.5 - "Decay Heat Removal System" (3) UFSAR, Section 10.6 - "Emergency Feedwater System" (4) TMI Unit 1 Calculation C-3320-85-001, "RCS Decay Heat Removal-Ambient Losses,"

Revision 0, February 28, 1985 3-26d Amendment No. ~. ,r:Fl-

COOLING SYSTEM Applies to periodic testing requirement for emergency core cooling To that the emergency core cooling systems are operable.

1

a. At the frequency specified in the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and system high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable.

b. The test will be considered satisfactory if the valves (MU-V-14A/B

& 16A/B/C/D) have completed their travel and the make-up pumps are running as evidenced by system flow. Minimum acceptable injection flow must be greater than or equal to 431 gpm per HPI pump when pump discharge pressure is 600 psig or greater (the pressure between the pump and flow limiting device) and when the RCS pressure is equal to or less than 600 psig.

c Testing which requires HPI flow thru MU-V16A/B/C/D shall be conducted only under either of the following conditions:

1) Indicated RCS temperature shall be greaterthan 313°F.

2) Head of the Reactor Vessel shall be removed.

4.5.2.2 Low Pressure Injection

a. At the frequency specified in the Surveillance Frequency Control Program and following maintenance or modification that affects system flow characteristics, system pumps and high point vents shall be vented, and a system test shall be conducted to demonstrate that the system is operable. The auxiliaries required for low pressure injection are all included in the emergency loading sequence test specified in 4.5.1.

b. The test will be considered satisfactory if the decay heat pumps have been successfully started and the decay heat injection valves and the decay heat supply valves have completed their travel as evidenced by the control board component operating lights. Flow shall be verified to be equal to or greater than the flow assumed in the Safety Analysis for the single corresponding RCS pressure used in the test.

4-41 Amendment No. 19, 57, 68, 149, 203, 225, 234, 274, 276, '

c. operable, by observation l"\f"\£~r".:l1'11"\M or

a. in Frequency Control shall conducted to demonstrate proper operation system. Verification shall be made that the check and isolation valves in the core cooling flooding tank discharge operate properly.

b. The test will considered satisfactory if control board indication of core flooding tank level verifies that all valves have opened.

4.5.2.4

a. At the frequency specified in the Surveillance Frequency Control Program, the components required for emergency core cooling will be tested.

b. The test will be considered satisfactory if the pumps and fans have been successfully started and the valves have completed their travel as evidenced by the control board component operating lights, and a second means of verification, such as: the station computer, verification of pressure/flow, or control board indicating lights initiated by separate limit switch contacts.

Bases The emergency core cooling systems (Reference 1) are the principal reactor safety features in the event of a loss of coolant accident. The removal of heat from the core provided by these systems is designed to limit core damage.

The low pressure injection pumps are tested singularly for operability by opening the borated water storage tank outlet valves and the bypass valves in the borated water storage tank fill line. This allows water to be pumped from the borated water storage tank through each of the injection lines and back to the tank.

The minimum acceptable HPl/LPI flow assures proper flow and flow split between injection legs.

With the reactor shutdown, the valves in each core flooding line are checked for operability by reducing the reactor coolant system pressure until the indicated level in the,,core flood tanks verify that the check and isolation valves have opened.

Reference (1) UFSAR, Section 6.1 - "Emergency Core Cooling System" 4-42 Amendment No.57, 68, 149, 157, 167, 225, ~

ootemttal to r>C>'lf<JllF'>>"' voids and of entrained intrusion and accumulation necessary for proper of the ECCS pump and of noncondensible gas into the review is su1m1em1ente<l and to confirm the location and orientation of important components that can become or could otherwise cause gas to be or difficult to remove during system locations on plant and such as With and the ECCS is OPERABLE when it is sufficiently filled with water. criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for volume at the suction or of a pump), the Surveillance is not met. If it is determined by evaluation that the ECCS is not rendered inoperable by the accumulated gas (i.e., the is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

ECCS locations to gas accumulation are monitored and, if gas is found, the gas volume is compared to the criteria for the location. Susceptible locations in the same system flow path which are to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REACTOR BUILDING COOLING AND ISOLATION SYSTEM Applies to of the reactor building cooling and isolation systems.

verify that the reactor building cooling systems are operable.

a. Reactor Building Spray System

1. At the frequency specified in the Surveillance Frequency Control Program and simultaneously with the test of the emergency loading sequence, a Reactor Building 30 psi high pressure test signal will start the spray pump. Except for the spray pump suction valves, all engineered safeguards spray valves will be closed.

Water will be circulated from the borated water storage tank through the reactor building spray pumps and returned through the test line to the borated water storage tank.

The operation of the spray valves will be verified during the component test of the R. B. Cooling and Isolation System.

The test will be considered satisfactory if the spray pumps have been successfully started.

2. Compressed air will be introduced into the spray headers to verify each spray nozzle is unobstructed at the frequency specified in the Surveillance Frequency Control Program.

b. Reactor Building Cooling and Isolation Systems

1. At the frequency specified in the Surveillance Frequency Control Program, a system test shall be conducted to demonstrate proper operation of the system.

2. The test will be considered satisfactory if measured system flow is greater than accident design flow rate.

4-43 Amendment No. 167, 198, 212, 225, 274 ~8

a. At the frequency in the Surveillance Frequency Control Program, the components required for Building Cooling and Isolation will tested.

b. if the valves have expected as by the control board component operating lights, and a second means of verification, such as: the station computer, local verification, verification of pressure/flow, or control board component operating lights initiated by limit switch contacts.

The Building Cooling and Isolation and Reactor Building Spray System are designed to remove the heat in containment atmosphere to prevent the building pressure from exceeding the design pressure (References 1 and 2).

The delivery capability of one Reactor Building Spray Pump at a time can be tested by opening the valve in the line from the borated water storage tank, opening the corresponding valve in the test line, and starting the corresponding pump.

With the pumps shut down and the Borated Water Storage Tank outlet valve closed, the Reactor Building spray injection valves can each be opened and closed by the operator action. With the Reactor Building spray inlet valves closed, low pressure air can be blown through the test connections of the Reactor Building spray nozzl r t that the flow paths are open.

The equipm , instrumentation of the Reactor Building Cooling System are arranged so that they can be visually inspected. The cooling units and associated piping are located outside the secondary concrete shield.

Personnel can enter the Reactor Building during power operations to inspect and maintain this equipment.

The Reactor Building fans are normally operating periodically, constituting the test that these fans are operable.

Reference (1) UFSAR, Section 6.2 - "Reactor Building Spray System" (2) UFSAR, Section 6.3 - "Reactor Building Emergency Cooling System" 4-44 Amendment No. 68, 149, 157, 167, ~

cornpc:me~nts have the potential to r<"'""'r"" voids and .,...,,,.,,,rL>fet' of intrusion and accumulation is for proper trains and may also water hammer and pump With Spray OPERABLE when it is sufficiently filled with water. criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for volume at the suction or of a pump), the Surveillance is not met. If it is determined evaluation that the Reactor Building Spray System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

Reactor Building Spray locations susceptible to gas accumulation are monitored and, if gas is volume is compared to the acceptance criteria for the location. Susceptible locations in the same flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel For these locations alternative methods operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

1.6

"""'"e- 1 "'10 r.c~r1 s1at1s;ta<:to1rv if control board indication visual operated RCS Temperature than or equal to 1 At the frequency specified in the Surveillance Frequency Control Program, verify operability of the means for OHR required by Specification by observation of console status indication.

• These requirements supplement the requirements of Specifications 4.5.2.2 and 4.5.4.

The ASME Code specifies requirements and acceptance standards for the testing of nuclear safety related pumps. The EFW Pump test frequency specified by the ASME Code will be sufficient to verify that the turbine-driven and both motor-driven EFW Pumps are operable.

Compliance with the normal acceptance criteria assures that the EFW Pumps are operating as expected. The surveillance requirements ensure that the overall EFW System functional capability is maintained.

Deferral of the requirement to perform IST on the turbine-driven EFW Pump is necessary to assure sufficient OTSG pressure to perform the test using Main Steam.

Periodic verification of the operability of the required means for OHR ensures that sufficient OHR capability will be maintained.

4-52a Amendment No. 78, 119, 124, 172, 242, 266, ~

accumulation is based on a review and elevation review walk downs to validate the and to confirm the location and orientation of important components that can become or could otherwise cause to be or difficult to remove during restoration. locations on and such as

'""*""""h"fY conditions.

Accer>tat1ce criteria are established for the volume of accumulated gas at susceptible locations. If accumulated is discovered that exceeds the criteria for the susceptible location (or the volume of accumulated at one or more locations exceeds an criteria for gas volume at the suction or of a pump), the Surveillance is not met. If it is determined by sutJseqrn~nt evaluation that the OHR is not rendered inoperable by the accumulated gas (i.e., the 0111*~ 1 " 1 ""r"r'" filled with water), the Surveillance may be declared met. Accumulated gas should within the criteria limits.

locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is to the criteria for the location. Susceptible locations in the same flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

SR 4.9.2.2 is modified by a Note that states the SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after RCS temperature is less than or equal to 250 degrees F. In a rapid shutdown, there may be insufficient time to verify all susceptible locations prior to RCS temperature reaching less than or equal to 250 degrees F.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

ATTACHMENT 3 Equivalent TMI, Unit 1 Technical Specification Nomenclature Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50

Attachment 3 Equivalent TMI, Unit 1 TS Nomenclature Page 1of1 STS TMI, Unit 1 TS Number nue Part Number Title Part Decay H l R mov I (OHR) C pab1hty-346 RC oops- 00 4 SR34 3 49 SR 4 9.2 2 PenocUc Test.mg 3 46 RCS Looos - MOOE 4 LCO Bases 3 1 Reactor Coolant System Bases 49 D yH t Remov I (OHR) Capability-346 RCS Loops - MOD 4 SR ases Bases Periodic Testma Decay Heat Removal (OHR) Capability-347 RCS Loops - MOO 5, oops Filled SR 3 4 7 4 4 SR 4 9 2.2 Penod1c Test1nQ 3.4 7 RCS Loops

• MOOE 5 Looos Fiiied LCO Bases 3.1 Reactor Coolant System Bases Decay H at Remov I (OHR) Capability-347 RCS Loops - MOO 5, Loops ill SRB s 49 Bases Penod1c Testing Decay H at Removal (OHR) Capability-3.4 8 RCS Loops - MOOE 5, Loops Not Filled SR 3 4.8 3 49 SR 4.9 2.2 Penochc Testing 3 48 RCS Loops - MODE 5. Loops Not illed LCO Bases 3. 1 Reactor Coolant System Bases Decay Heat Removal (OHR) Capability-3.48 RCS Loops MOOE 5, Loops Not Filled SR Bases 4.9 Bases Penodic Testina SR 4.5 2.1 .d 352 ECCS - Operating SR 3 5 2 3 452 Emergency Core Cooling System SR 4 5 2 2.c SR 4.5.2.3.c Emergency Core Cooling, Reactor 3.5.2 ECCS - Operating LCO Bases 33 Building Emergency Cooling and Reactor Bases Building Sorav Svstems 3.5 2 ECCS - OoerabnQ SR Bases 4 5.2 Emergencv Core Cooling System Bases Emergency Core Cooling, Reactor 3.5.3 ECCS - Shutda.vn LCO Bases 33 Building Emergency Coohng and Reactor Bases Bulldina $Oldv Svstems Reactor Building Cooling and Isolation 3.6 6 Containment Spray and Cooling Systems SR 3.6 .6 4 4 5.3 SR 4 5.3.1.a.3 Svstem Emergency Core Coollng, Reactor 3.6.6 Containment Spray and Cooling Systems LCO Bases 3.3 Building Emergency Cool ng and Reactor Bases Buildlna Scrav Svstems Reactor Building Cooling and Isolation 3.6 6 Containment Spray and Coohng Systems SR Bases 453 Bases Svstem OHR and Coolant Circulation - High Decay Heat Removal (OHR) Capab lity-3.9.4 SR 3.94.2 49 SR4 9.2.2 Water Level Penodic Tesbna OHR and Coolant Circulallon - High 3.9.4 LCO Bases 34 Decay Heat Removal (OHR) Capability Bases Water Level OHR and Coolant Circulation - High Decay Heat Removal (OHR) Capability-3.9.4 SR Bases 4.9 Bases Water Level Penodic Testing OHR and Coolant Circula ion- Low Water Decay Heat Removal (OHR) Capability-3 9.5 SR 3 9.5.3 49 SR 4.9 2.2 Level Penodic Testina OHR and Coolant Circulation - La.v Water 3.9.5 LCO Bases 3.4 Decay Heat Removal (OHR) Capability Bases Level OHR and Coolant Circulation- Low Water Decay Heat Removal (OHR) Capability-3.9 5 SR Bases 49 Bases Level Periodic Testina