ML032250657

License Amendment Application to Modify Technical Specification 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems

ML032250657
Person / Time
Site:	Davis Besse
Issue date:	08/11/2003
From:	Meyers L FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2834
Download: ML032250657 (26)

Text

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FENOC 5501 North State Route 2 FhrstnryNuclar OeuugComany~a Oak Harbor, Ohio 43449 LewWs Mers 419-321.7599 Chief Operating Officer Fax: 419-321-7582 Docket Number 50-346 10 CFR 50.90 License Number NPF-3 Serial Number 2834 August 11, 2003 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Davis-Besse Nuclear Power Station License Amendment Application to Modify Technical Specification 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - Tavg 2 2800F (License Amendment Request No. 03-0004)

Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, an amendment is requested for the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS). The proposed amendment would modify Technical Specification (TS) 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - Tavg 2 2800F.

Limiting Condition for Operation (LCO) 3.5.2 requires two independent Emergency Core Cooling Systems (ECCS) Subsystems to be operable. Surveillance Requirement (SR) 4.5.2.f requires each ECCS Subsystem to be demonstrated operable by performing a vacuum leakage rate test of the watertight enclosure for Decay Heat Removal System valves DH-11 and DH-12 that assures the electric motor operators on valves DH-1 1 and DH-12 will not be flooded for at least seven (7) days following a Loss-of-Coolant Accident (LOCA). This SR provides assurance that a circulation flow path will be maintained to prevent boric acid concentration build-up and boric acid precipitation in the reactor vessel post-LOCA.

The proposed amendment would allow the relocation of SR 4.5.2.f to the DBNPS Updated Safety Analysis Report (USAR) Technical Requirements Manual (TRM). Future changes to the relocated specification, such as a planned extension of the current 18-month surveillance interval, will be evaluated under the requirements of Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR), and the NRC will be informed of these changes in accordance

V Docket Number 50-346 License Number NPF-3 Serial Number 2834 Page 2 with the requirements of 10 CFR 50.59(d)(2), as applicable, and the USAR update requirements of 10 CFR 50.71(e). to this letter contains the technical justification for these proposed changes and the proposed no significant hazards consideration determination. Approval of the proposed amendment is requested by February 6,2004. Once approved, the amendment shall be implemented within 120 days.

The proposed changes have been reviewed by the DBNPS Station Review Board and Company Nuclear Review Board.

Should you have any questions or require additional information, please contact Mr. Kevin L. Ostrowski, Manager - Regulatory Affairs, at (419) 321-8450.

Very truly yours,

-t:'A.O

MKI, Enclosures cc:

Regional Administrator, NRC Region III J. B. Hopkins, NRC/NRR Senior Project Manager D. J. Shipley, Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

C. S. Thomas, NRC Region III, DB-1 Senior Resident Inspector Utility Radiological Safety Board

Docket Number 50-346 License Number NPF-3 Serial Number 2834 Page 3 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 This submittal requests changes to the Davis-Besse Nuclear Power Station Unit Number 1, Facility Operating License Number NPF-3. The statements contained in this submittal, including its associated enclosures and attachments, are true and correct to the best of my knowledge and belief.

I declare under penalty of perjury that I am authorized by the FirstEnergy Nuclear Operating Company to make this request and the foregoing is true and correct.

Executed on:

vi tkl 03 Le0 W. Myt, 1

rating Officer

,J/

Docket Number 50-346 License Number NPF-3 Serial Number 2834 DAVIS-BESSE NUCLEAR POWER STATION EVALUATION FOR LICENSE AMENDMENT REQUEST NUMBER 03-0004 (21 pages follow)

LAR 03-0004 Page 1 DAVIS-BESSE NUCLEAR POWER STATION EVALUATION FOR LICENSE AMENDMENT REQUEST NUMBER 03-0004

Subject:

License Amendment Application to Modify Technical Specification 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - Tavg 2 280'F

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration (NSHC) 5.2 Applicable Regulatory Requirements/Criteria

7.0 REFERENCES

8.0 ATTACHMENTS

LAR 03-0004 Page 2

1.0 DESCRIPTION

This letter is a request to amend the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3.

The proposed change would revise Technical Specification (TS) 3/4.5.2, Emergency Core Cooling Systems - ECCS Subsystems - Tayg 2 2801F, to relocate Surveillance Requirement (SR) 4.5.2.f to the DBNPS Updated Safety Analysis Report (USAR) Technical Requirements Manual (TRM).

2.0 PROPOSED CHANGE

Limiting Condition for Operation (LCO) 3.5.2 requires two independent Emergency Core Cooling Systems (ECCS) Subsystems to be operable. Surveillance Requirement (SR) 4.5.2.f requires each ECCS Subsystem to be demonstrated operable by performing a vacuum leakage rate test of the watertight enclosure for Decay Heat Removal System valves DH-1 1 and DH-12 that assures the electric motor operators on valves DH-1 1 and DH-12 will not be flooded for at least seven (7) days following a Loss-of-Coolant Accident (LOCA). This SR provides assurance that a circulation flow path will be maintained to prevent boric acid concentration build-up and boric acid precipitation in the reactor vessel post-LOCA.

The proposed amendment would allow the relocation of Technical Specification Surveillance Requirement 4.5.2.f to the DBNPS Updated Safety Analysis Report Technical Requirements Manual. Surveillance Requirement 4.5.2.f would be relocated to the USAR TRM at the time of the implementation of the approved license amendment. Future changes to the relocated specification, such as a planned extension of the current 18-month surveillance interval, will be evaluated under the requirements of Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR), and the NRC will be informed of these changes in accordance with the requirements of 10 CFR 50.59(d)(2), as applicable, and the USAR update requirements of 10 CFR 50.71(e).

There is a discussion of SR 4.5.2.f included in the Bases associated with TS 3/4.5.2. In conjunction with, but separate from this license amendment application, the Bases discussion will also be relocated to the USAR TRM. This change will be made under the provisions of the DBNPS TS Bases Control program. The affected TS Bases pages, as currently existing, are included in Attachment 3 for information.

3.0 BACKGROUND

The vacuum leakage rate test of the watertight enclosure for valves DH-1 1 and DH-12 (SR 4.5.2.f) is required to be performed: (1) At least once per 18 months, (2) After each opening of the watertight enclosure, and (3) After any maintenance on or modification to the watertight enclosure which could affect its integrity. As stated in the TS Bases, this SR ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem operability is maintained.

LAR 03-0004 Page 3 A watertight enclosure is required for valves DH-1 1 and DH-12 because these valves are located in an area which would be flooded following a LOCA, and the valves' electric motor operators, by themselves, are not qualified for submergence. The current TS Surveillance Requirement requires a vacuum leakage rate test of the watertight enclosure for valves DH-I 1 and DH-12 to assure their motor operators would remain unsubmerged for a period of up to 7 days following a LOCA. This ensures that the valves will remain capable of opening, providing a circulation flow path for reactor coolant. This circulation flow path will prevent boric acid concentration build-up and boric acid precipitation in the reactor vessel post-LOCA.

As described in USAR Section 6.3.3.1.2.1, "Boron Precipitation Control," two active means of ensuring the chemical additive concentration remains below its solubility limit throughout the post-accident cooling period are provided, a primary boron precipitation control (BPC) method and a backup method. The primary BPC method currently utilizes the auxiliary pressurizer spray (APS) flow path. For this method, High Pressure Injection (HPI) Pump 2 takes suction from the discharge of Decay Heat (DH)/Low Pressure Injection (LPI) Pump 2 to supply water to the APS line via a tie-line, providing dilution flow to the core via the pressurizer.

The backup BPC method utilizes one of the two operating DH/LPI pumps taking suction from the DH "drop line" via valves DH-1 1 and DH-12, and discharges a low (throttled) flow rate into the reactor vessel via the core flood nozzles. The flow through the drop line allows forward flow through the reactor vessel, so that any amount of flow of relatively low concentration water from the LPI train aligned to the containment emergency sump will enter and dilute the boric acid in the core. The backup BPC method would only be utilized if the primary method is unavailable and if both DHILPI pumps are functioning.

During the ongoing Thirteenth Refueling Outage (13RFO), a plant modification to provide an improved BPC method is planned for implementation. The plant modification would add a new cross-tie line and associated valving and instrumentation, allowing the discharge of either DH/LPI pump to backflow through the DH-1 and DH-12 drop line and into the reactor vessel.

This flow path would become the new primary BPC method. In addition, the current primary BPC method would become the backup BPC method, and the current backup BPC method would no longer be used. However, since the improved BPC methodology would still utilize a flow path through valves DH-1 and DH-12, the watertight enclosure remains an important design feature.

The watertight enclosure consists of a stainless steel-lined concrete cavity located at the lower elevation of the containment vessel. It is approximately nineteen feet-six inches long by seven feet wide by seven feet deep. During 13RFO, extensive modifications were made to the watertight enclosure to improve its sealing capability. Quarter-inch thick stainless steel plates were anchor-bolted to the four walls and the floor. For the top, quarter-inch thick stainless steel plates were mounted to the underside of existing structural steel supports, below the existing carbon steel checker plate. Expansion joints were incorporated into the design to accommodate differential expansion. Welded construction was used for joints and seams of the new stainless steel liner. In addition, the supporting structure for the top plates and access openings was

I.;

LAR 03-0004 Page 4 reinforced to improve the leaktightness of the seals for the removable manway and access covers.

The DBNPS is currently operating on a 24-month fuel cycle, whereas SR 4.5.2.f is required to be performed at an 18-month frequency. Therefore, in the event that an outage of sufficient duration does not occur during an operating cycle, an early plant shutdown to perform the test may be necessary. Upon relocation of the SR to the USAR TRM, DBNPS personnel plan to perform an evaluation under the 10 CFR 50.59 process to determine if the current surveillance interval can be changed to a refueling interval frequency, based in large part on the modifications made to improve the sealing capability of the watertight enclosure. This would eliminate the need to perform the surveillance in mid-cycle outages, resulting in reduced radiation exposure to personnel.

The conversion of SR 4.5.2.f to a refueling interval frequency would be consistent with the original licensing basis. The SR is a non-standard, plant-specific requirement that was added to the DBNPS TSs, with an 18-month frequency, at the time the Operating License was issued in 1977. This requirement was added due to the design of the watertight enclosure, and was based upon performing testing at a refueling outage frequency rather than at a fixed absolute timespan.

Specifically, Section 6.3.3.5, "Submerged Valves," of Supplement I to the NRC's Operating License Safety Evaluation Report (NUREG-0136) refers to the NRC staff's requirement to perform "an acceptable leakage test of this enclosure at each refueling interval."

4.0 TECHNICAL ANALYSIS

The proposed amendment would relocate SR 4.5.2.f from ECCS TS 3/4.5.2 to the USAR TRM.

This SR provides assurance that at least one BPC method will be available to prevent boric acid concentration build-up and boric acid precipitation in the reactor vessel post-LOCA. The SR is a non-standard, plant-specific requirement that was added to the existing TS 3/4.5.2 at the time the Operating License was issued in 1977. A specific LCO for BPC methods was not added at that time. In addition, there is no LCO for BPC methods in the improved Standard Technical Specifications for Babcock and Wilcox-type plants (NUREG-1430).

The removal of the detail for performing this SR from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. Also, this change is acceptable because the detail in this SR is being relocated to the USAR TRM and will therefore continue to be adequately controlled. Any changes to the USAR TRM are made under the requirements of Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR), which ensures changes are properly evaluated. It is concluded that there is no adverse effect on plant safety as a result of this relocation.

LAR 03-0004 Page 5 5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Technical Specification Limiting Condition for Operation 3.5.2 requires two independent Emergency Core Cooling Systems (ECCS) Subsystems to be operable. Surveillance Requirement 4.5.2.f requires each ECCS Subsystem to be demonstrated operable by performing a vacuum leakage rate test of the watertight enclosure for Decay Heat Removal System valves DH-1 and DH-1 2 that assures the motor operators on valves DH-1 1 and DH-12 will not be flooded for at least seven (7) days following a Loss-of-Coolant Accident (LOCA). This surveillance requirement provides reasonable assurance that a circulation flow path will be maintained to prevent boric acid concentration build-up and boric acid precipitation in the reactor vessel post-LOCA. The proposed amendment would relocate Surveillance Requirement 4.5.2.f to the DBNPS Updated Safety Analysis Report (TSAR) Technical Requirements Manual (TRM). Future changes to the relocated specification will be evaluated under the requirements of Section 50.59 of Title 10 of the Code of Federal Regulations (10 CFR), and the NRC will be informed of these changes in accordance with the requirements of 10 CFR 50.59(d)(2), as applicable. and the USAR update requirements of 10 CFR 50.71(e).

An evaluation has been performed to determine 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.

Under the proposed change, initial conditions and assumptions remain as previously analyzed for accidents in the Davis-Besse Nuclear Power Station Updated Safety Analysis Report. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Under the proposed change, the manner in which the watertight enclosure is sealed and tested is not altered, and the operability requirements of the watertight enclosure for Decay Heat Removal System valves DH-1 1 and

LAR 03-0004 Page 6 DH-12 will continue to be adequately addressed by testing. No different accident initiators or failure mechanisms are introduced by the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

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

Response: No.

Since there are no new or significant changes to the initial conditions contributing to accident severity or consequences, there are no significant reductions in a margin of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, it is concluded that the proposed amendment 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.

5.2 Applicable Regulatorv Reguirements/Criteria As stated in the DBNPS Updated Safety Analysis Report (USAR), Appendix 3D, "Conformance with the NRC General Design Criteria, Safety Guides, and Information Guides," the design of the Davis-Besse Nuclear Power Station meets the intent of 10 CFR 50, Appendix A, "General Design Criteria for Nuclear Power Plants," as published in the Federal Register on February 20, 1971, and as amended in the Federal Register on July 7, 1971.

Regarding General Design Criterion (GDC) 35, "Emergency Core Cooling,"

USAR Appendix 3D states:

A system to provide abundant emergency core cooling is provided. The system safety function is to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities are provided to assure that for onsite electric power system operation (assuming that offsite power is not available) and for offsite electric power system operation (assuming that onsite power is not available) the system safety function can be accomplished, assuming a single failure.

LAR 03-0004 Page 7 Abundant emergency core cooling is provided by the low-pressure injection (decay heat removal), high-pressure injection, and the core flooding systems. These three systems make up an Emergency Core Cooling System (ECCS) that maintains core cooling in the event of a loss-of-coolant accident (LOCA).

Redundancy of components, power supplies, and initiation logic and separation of functions are provided so that a single failure does not prevent the ECCS from fulfilling its function. The ECCS may be operated from either onsite or offsite power supplies.

The primary function of the Emergency Core Cooling System is to deliver cooling water to the reactor core in the event of a LOCA. The system provides protection for all potential break sizes in the reactor coolant system pressure boundary piping up to and including the double-ended rupture of the largest pipe. In addition, breaks in the High Pressure Injection line and the Core Flood Tank line are postulated.

The basic design criteria for loss-of-coolant accident evaluation are as follows:

a.

The calculated maximum fuel element cladding temperature shall not exceed 2200 'F.

b.

The calculated total oxidation of the fuel cladding shall not exceed 17% of the total cladding before oxidation.

c.

The amount of hydrogen generated from cladding metal-water reaction does not exceed 1% of the total amount of cladding in the reactor.

d.

The core geometry is maintained in a state that is amenable to cooling.

e.

The cladding temperature is reduced and maintained at an acceptably low value and decay heat is removed for extended periods of time.

For a rupture in the steam piping, the Emergency Core Cooling System adds shutdown reactivity, so that with minimum tripped rod worth and minimum ECCS operation, the reactor core does not return to criticality; thus, there is no core damage.

Reference:

USAR Chapter 6 and 10 CFR 50.46

LAR 03-0004 Page 8 Pursuant to 10 CFR Section 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," paragraph (a)(1)(ii),

the Davis-Besse Nuclear Power Station Emergency Core Cooling System (ECCS) is modeled in conformance with the required and acceptable features of 10 CFR 50, Appendix K, "ECCS Evaluation Models." Compliance with the 10 CFR 50.46 acceptance criteria is described in USAR Section 6.3, "Emergency Core Cooling System." One of the 10 CFR 50.46 acceptance criteria pertains to establishment of a mode of long-term core cooling. USAR Section 6.3.3.1.2.1, "Boron Precipitation Control," describes the methodology to ensure that the boric acid concentration in the reactor vessel will remain dilute throughout the post-accident cooling period, thereby ensuring that the long-term cooling acceptance criterion is satisfied.

Under the proposed license amendment application, Surveillance Requirement 4.5.2.f, which is related to the boron precipitation control methodology, is being relocated from the Technical Specifications to the USAR Technical Requirements Manual. Compliance with the intent of 10 CFR 50, Appendix A, GDC 35, and the requirements of 10 CFR 50.46 and 10 CFR 50, Appendix K are unaffected by the proposed relocation of the Surveillance Requirement.

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.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1.

DBNPS Operating License NPF-3, Appendix A Technical Specifications through Amendment No. 254.

LAR 03-0004 Page 9

2.

DBNPS Updated Safety Analysis Report through Revision No. 23.

3.

NUREG-0136 dated December 1976 (DBNPS Unit 1 Safety Evaluation Report),

including Supplement 1, dated April 1977.

4.

10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

5.

10 CFR 50.59, "Changes, Tests, and Experiments."

6.

NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants,"

Revision 2.

7.

Engineering Change Request (ECR) 03-0146-00, "Boron Precipitation Control Modification."

8.0 ATTACHMENTS

1.

Proposed Mark-Up of Technical Specification Pages

2.

Proposed Retyped Technical Specification Pages

3.

Technical Specification Bases Pages

LAR 03-0004 PROPOSED MARK-UP OF TECHNICAL SPECIFICATION PAGES (4 pages follow)

mmma=oOKBOM INOO BUSYSTBMS -XA.Z2=O 0

LM1NO CONDONFOROPERATON 352 Two Independent EOCS subsystems shall be OPERABLE with each subsystem comprised of:.

a. One OPERABLE high pressure Injection (HPI) pump,

b. One OPERABLE low pressure bdection OM) pump,

c. One OPERABLE decay heat cooler, and

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

APPLICABML

MODES 1, 2 and 3.

ACl0ON:

a. With one EHPI train Inoperable, restore the inoperable BPI train to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. wrth one LPI train or. Its associated decay heat cooler Inoperable, restore the inoperable equipment to OPERABLE status within 7 days or be In HOT SHUIDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. In the event thECCS ls actuated and lnjects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulat&d actuation cycles to date, URVEMLLANCE

_EQUIRME 45S2 Each OCS subsysm shal be d OPERABLE

a. Atleas onoe per3 dys by vefg wa hvav (mnuaL power operd orauto

) in theflowpathttIsnotocked seledro& wlesec di posMon, IS ns correctposIton.

[

ADDITIONAL CHANGES PPEVIOUSLY PROPOSED BY LETTER Serial No._ 2

.0 DateI SA/p4 THIS PAGE PIRIMMBED FOR INFORMATIN OfiY DAVIS-BMMSB UNIT I 314 5-3 AmendmentNo. 36,182, 253

£

t Revised by NRC Letter Dated

-June 6, 1995

b.

At least once each REFUELING INTERVAL, or prior to operation after ECCS pipinf has been drained by verifying that the ECm piping Is full of water by venting the ECCS pump casings and discharge piping high points.

C.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) Is present In the containment which could be transported to the containment emergency sump and cause restriction of the pump suction during LOCA conditions. This visual inspection shal be performed:

1.

For all accessible areas of the containment prior to establishing CONTAINHENT INTEGRITY, and

2.

For all areas of containment affected by an entry, at least once daily while work-is ongoing and again during the final exit after completion of work (containment closeout) when CONTAINMENT INTEGRITY is established.

d.

At least once each REFUELING INTERVAL by:

1.

Verifying that the interlocks:

a)

Close OH-ll and OH-12 and deenergize the pressurizer heaters, -if either 0H-1l or DH-12I s open and a simulated reactor coolant system.pressure which is 1=13

¢greater than the Allowable Value (<328ipsig) is IA6464 E

applied.

The interlock to close DHII and/or DH-12 Is C: :W r^

not required if the valve is closed and 480 V AC power is disconnected from its motor operators.

c(I2 Ii. sb)

Prevent the opening of OH-l1 and DH-IV ihen a g~s F"MAO simulated or actual reactor coolant system pressure which is greater than the Allowable Value (<328 psig) is applied.

<2¢g

2.

a)

A visual inspection of the containment emergency sump which verifies that the subsystem suction bnlets are not restricted by debris and that the sump components.

>a 9ii(trash racks screens, etc.) show no evidence of b)S9 structural distress or corrosion.

C b)

Verifying that on a Borated Water Storage Tank (BWST)

Low-Low Level Interlock-trip, with the motor operators

.^-ea

>bnr for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the B ST Outlet Valve HV-DH7A (HV-DHJB) automatically close in 675 seconds after thewoperator manually pushes the control switch to open the Containment Emergency Sump Valve HN-DH9A (HV-DH9B) which should be verified to open In a7B seconds.

3.

Deleted DAVIS-BESSE, UNIT 1 3/4 5-4 Amendment No. A.5,8404 XTin 48-.

49 G 1 4~

21B

-r EMERGENCY CORE COOLING SYSTEMS SUVEILLAN REMQUIREMENTS (Continued

4.

Verifying that a minimum of 290 cubic feet of trisodium phosphate dodecahydrate (MSP) is contained within the TSP storage baskets.

5.

Deleted

6.

Deleted

e.

At least once each REFUELING INTERVAL, by

1.

Verifying that each automatic valve in the flowpath actuates to its correct position on a safety injection test signal.

2.

Verifying that each BPI and LPI pump starts automatically upon receipt of a SFAS test signal.

fC By performing a vauum leakage rate test of the watertight enclosure

£1for vave DR 1 1 and DR 12 tht assures the motor eperatre _n valves DH ! 1 and DH 12 will not be floaded for-at least 7 day6 fellowing a LOGA:

At least ene pe pA months.

2.

After ach orpening of th watcr.

t enclosur-e.

ax A&ft mny maintenne en or-modification to thc wvatsrig enclosure which could affeet its integrity'.

The inspection port on the watertight enclosure may be opened without requiring perforznnae of the vacuum leakage rate test, to perfom inspections. Afterm use, the inspection port must be verified as closed in its correct position. Provisions of TS 3.0.3 -

ire nnt cMnksofmie Munn tncre Mn~neoten-M-1 Xeted Err--

^-o Orx-

g.

By verifying the correct position of each mechanical position stop for valves DH-14A and DH-14B.

- 1.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of the opening ofthe valves to thei mechanical position stop or following completion of maintenance on the valve when the LPI system is required to be OPERABLE.

2.

At least once each REFUELING INTERVAL.

DAVIS-BESSE, UNIT I 3/4 5-5 Amendment No. 20,26,40,191,207,215,216,

A4 I

EMERGENCY CORE COOLIHG SYSTEms THIS PAGE PROVIDED COD I~mMMMAU1N lllY

- 6 SURVEtLLANCE RIQUIREflS (Continued)

I W1 li UIHIvIC t fun

h.

By perfoming a flow balance test, during shutdown. following completion of modifications to the HPI or LPI subsystemu that alter the subsystem flow characteristics and verifying the followfng flow rates:

HPI System - Single PuMp I

lUlU Injection Leg 1.1

, 375 gpm at 400 psfg*

Injection Leg 1-2

_ 375 gpm at 400 psfg*

Injection Leg 2-1 D 375 gpm at 400 psfg*

Infection Leg 2:2 7 375 gS t 400 psfg*

LPI System - Sinple Pwip Injection Leg 1 Injection Leg 2

.% 2650 gpm Z 2650 gpm at tO0 estg**

at 100 psig**

I ADDITIONAL CHANGES PREVIOUSLY PROPOSED BY LETTER Serial No. 2 9 4 9 Date 5k (o Reanctor coolant Pump discharge.

Reactor coolant vessel.

Pressure at Pressure at the NPI nozzle in the reactor coolant the core flood nozzle on the reactor IDAVIS-BESSE.

tJMIT I

314 S-.Z An-end~en.,

No.

20

LAR 03-0004 PROPOSED RETYPED TECHNICAL SPECIFICATION PAGES (1 page follows)

EMERGENCY CORE COOLING SYSTEMS SUEILANM REOI CEMN LCntinued):

4.

Verifying that a minimum of 290 cubic feet of trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

5.

Deleted

6.

Deleted C.

At least once each REFUELING INTERVAL, by I.

Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signal.

2.

Verifying that each HPI and LPI pump starts automatically upon receipt of a SFAS test signal.

£ Deleted

g.

By verifying the correct position of each mechanical position stop for valves DH-14A and DH-14B.

1.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of the opening of the valves to their mechanical position stop or following completion of maintenance on the valve when the LPI system is required to be OPERABLE.

2.

At least once each REFUELING INTERVAL.

DAVIS-BESSE, UNIT I 314 5-5 Amendment No. 20,26,40,191,207,215,216,

LAR 03-0004 TECHNICAL SPECIFICATION BASES PAGES (4 pages follow)

Note: The Bases pages are providedfor information only.

BASES--

TH6lIN PAGE PROVIDED 31451 WI ffDf

@F)lNTAMKS FOR 1NMRMADO 01f The OPEUBnTY of each core neoodf tank s that a aient volume of

-borated water wil be Immediately forced Into thcreactor vessel in th event the RCS pressure fanIs below I rc ofthe t lnbis Initial u of water io tie vesse! provdes tie hital cooling m

cudining large RCS pe ruptures.

The lits on volume, boron concentration ad pressure ensure that the assumptions used for core flooding tank injection in the safety analysis ae met.

The tank Power opeated Isolation valvesare considered to be opr

" in the.

context of MM Std. 279-1971, which requires that bypasses of a proteve finction be removed

-- aitomaticaiy whenever permissive conditions are not met in additon, as these tank isolation valves fall to meet single fallure criteria, removal of power to the valves is required.

The one hour limit for operation with acore flooding tank (CFI) inoperable for reasons other than boron concentratibn not within limits minizes the time the plant Is exposed to a possible LOCA event ouing with failure of a CIT. which may result in uaccptable peak claddi empe

. With boron concenonfor one CF not within limits the Condition must be corrected wi 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The72 hour limit was developed consid that thc effects of reduced boron cotic~nragon on core subcriticality during reflood are minor. Boiling of t ECS water in the core durin reflood concentrates the boron In the saturated liquid that remains in the core In addition, tie volume of the CFrs Is still available for Section.

Since the boron requirements arq based on the average boron concentration of the total volume of both CFss the consequences are less severe tan they would be If the contents of a CFT wer not avalable, tr ijection.

. lhe completion times to bring te pl~ant to a MODE in wichthe lig ondition for Opration (0) does not apI are reasonable based on peratng experience. Ue completion timesow plat contions Changed In an ordedy manner and without l

ng plant CFT bor connan mp within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> deran 80 gallon volume nese will d

yw ieteleakagc from the RCS caused a reduction In boron c e

on to below d@tei~ro dlimiKltismnotrveborif conceanIf e added ater v

osi theiborae d

StO WME M because the water containd line BW Ts winCFT bo coneaione s.

3f4Sf and 3M4S.3 BOCS SU KIS SM The opeabit of two Independent ECCS sbsystems wit RCS average temperature 280P ensures that sUfficient emergency core cooling capability will bea le In the event of a LOC assng the loss of onesub em hrough any igle failure consideion Each BCS bystem consistsof one Hgh Prssure Injection P)trin one Low Pressure LIection (P) train (Including the associated deca heat cooler), and the necessary piin valves, instrumentation and coptrols to provide the required flowpaths from the Borated Water Storage DAVIS-BESSAAWT1 B 314 S-1 AmeadmeatNo. 20,191, 253

4SDMSEMROENCY ORiE COOLNhG SYSTIMS (B1CS)

BASS(Continued)

Tank (BWSI) or die Containment Emergency Sump to the reactor vesseL E e subsystem operating In conjunction with the core flooding tanks Is capable of supplying sufIclent core cooling to maintain the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of tie largest RCS cold leg pipe downward.

With RCS average temperature 2 280T, the Lmiting Condition for Operation (LCO) equires the OPERABLITY of a number of independent trains, the inoperability of one component In a train does not necessarily render the ECCS incapable of performing its function. Neither does the Inoperability of two different components, each in a different train, necessarily result in a loss of function for the BCCS. The intent of this LCO is to maintain a combination of equipment such that 100% of the safety injection flow equivalent to 100% of a single subsystem remains avawiable. Ihis allows Increased flexibility in plant operations under circumstances when components in opposite subsystems are inoperable.

With one or more components inoperable such that 100% of the flow equivalenit to a single OPERABLE BCCS subsystem Is not available, the facility is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be immediately entered.

I -addition, each: BCS subsystem proides long term core cooling capability in the tecirculaon mode during the accident recovery period.

TillS PAGE PROVIDED

'FOR INFORDRON ON DAVM-BH= UMM I B 314 5-la.

Amendment No.20,191.253 1

EM(ERGENCY CORE M~OLWNG SYSMflS RASMS With the RCS temperature below 280 F. one OPERABLE ECCS subsystem Is acceptable witbout sinole failure consideration on the basis of the stable reactIvIty condition of the reactor and the limited core cooling requirements.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

The function of the trisodium phosphate dodecahydrate (TSP) contained in baskets located In the containment normal sump or on the 565' elevation of containment adjacent to the normal sump, is to neutralize the acidity of the post-LOCA borated water mixture during containment emergency sump recirculation.

The borated water storage tank (BWST) borated water has a nominal pH value of approximately S.

Raising the borated water mixture to a pH value of 7 will -ensure that chloride stress corrosion does not occur In austenitic stainless steels In-the event that.chloride -levels: Increase as a result of, contamination on the-surfaces; of the--reactor containment-building.

Also, a pH of 7 is assumed for the containment -emergency sump-for iodine retention and removal post-LOCA by the containment spray system..

The Surveillance-Requirement (SR) associated with TSP ensures that the minimum required volume of TSP is stored in the baskets.

The minimum required volume of TSP is the volume that will achieve a post-LOCA borated water mixture pH of t 7.0, conservatvely consideringthe maximum possible sump water volume and

- the -maxitum possible boron concentration./ -The'amount of TSP required is based on the mass of TSP needed to achieve the required pH.

However, a requitred volume is verified by the SR, rather -than the mass, since it, is not feasible to weigh the entire amount of TSP in containment..The minimum required volume Is based on the manufactured density of TSP (53 lblft).

Since TSP can have a tendency to agglomerate from high humidity in the containment, the density may increase and the volume decrease during normal plant operation, however,

.;.solob ty characteristics are not.expected to change.

Therefore, considering possible agglomeation and increase -in density, verifying the minimum volume of TSP 1n containment is-conservative with resoedt to ensuring the capability to achieve the minium required pH.

The vdnglum required volume of lSP to meet a I analyt1l requrements is 250 ftW.

The surveillance requirement of 290 ft' includes 40 ft of-spare TSP as margin. Total basket capacity is 326 ft5.

Decay Heat Yemoval System valves O1-11and 0Ii2are located in an area that would be flooded following a LOCA. -These valves are located--in a-watertight enclosure to ensure -their operability up.to seven days following a-L0CA Surveillance Requirements are-provided to verify the acceptWble leak tightness of this enclosure.

An inspection port is located on this waterti ht enclosure; which Is typically used for performing inspections inside the enclosure.

During the vacum leaka e rate test, the inspection port Is in a closed-position and subject to the test. This inspection port say be subsequently.opened foruse -in vewing inside -the enclosure.

Opening this

-inspection portitill-not require-performance of the-vacuum leakage. ratqtest because of the desgn of the -closure fitting, which will preclude leakage under LOCA Wcontions when properly installed. Proper installation includes independent verificatAon.

DAVIS-BESSt.

AmnmetNoNnIp Hi j

AGRri ennY FOR INFORMATON ONL

.4.

0fflMMCYC0=Co0-VN(f THIS PAGE PROVIDED Surveillance requirements for throttle valve position stops and flow balance testing provide assuranc that proper 1CCS flows will be maintained In the event of a LOCA. MaInteance of proper flow resIstance and pressure drop In the piping system to each Irjection point Is necessary to: (1) prevent total pump flow from exceeding rmnout conditions when the system is in its minimum resistance configuratio n (2) provide the proper flow split between Injection points in accordance with the assumptions used in the BCCS-LOCA analyses, and (3) provide an acceptable level of total BOCS flow to all injection points equal to or above that assumed in the BOCS-LOCA analyses.

Containment1 mergency Sump Recirculation Valves DH-9A and DH-9B are d-&erzed during MODES s 2, 3 and 4 to precude postulated inadvertent open31g of the valves in the event of a Control Room fire, which could result in draining the Borated Water Storage Tank to the Cont Et mergency Sump and the loss of this water source for normal plant shutdown.

Re-energization of DH-9A and DH-9B Is permitted on an intermittent basis during MODES 1, 2, 3 and 4 under administrative controls. Station procedures identify the precautions which must be tak When re-enerlizlng these valves under such controls.

Borated Water Storage Tank(BWSI) outlet isolation vlves DH-7A and DH-7B are de-energized during MODES 1, 2.3, and 4 to preclude posted inadvertent closure of the valves j#te event of a flre, which could result in a loss of the availability of the BWST. Re-Tekgizatizon of valves DH-7A and DH-7R is permitted on an Intermittent basis during MODES t1, 3, and 4 under administrative controls. Station procedures Identify the irecautions which must be taken when re-nergizing these valves under such controls.

The Decay Heat Isolation Valve and Pmssurizer Heater Interlock setpolnt lsbased an preventing oav ressurization of the Decay Weat Removal System normal suction line piping. The value stated is the RCS pressure at the sensing instrumet% tap. It has been a4justed to reflect the elevation difference between the senso? location and the pipe of concern.

N 4,SABORATERATMORAGETAN The OPERABI3 MTY of the bopated water storage tank (BWS1) as part of the BOCS ensr that a sufficient suly of borated water Is available for Injection by the BOCS In the event of a LOCA. The limits on the BWST minimum volume (500,100 gallons of borated waer, conservatively rounded up from the calculated value of 500,051 gallons) and boron concentration ensure that:

I) sufficient water Is available within containment to permit cr on cooling flow to the core following manual switchover to the recirculation mode, and DAVIS-BESSE, UNIT I B 314 5-2a Amendment No. 1919 207,215,218, 241

Docket Number 50-346 License Number NPF-3 Serial Number 2834 COMMITMENT LIST THE FOLLOWING LIST IDENTIFIES THOSE ACTIONS COMMITTED TO BY THE DAVIS-BESSE NUCLEAR POWER STATION (DBNPS) IN THIS DOCUMENT. ANY OTHER ACTIONS DISCUSSED IN THE SUBMITTAL REPRESENT INTENDED OR PLANNED ACTIONS BY THE DBNPS. THEY ARE DESCRIBED ONLY FOR INFORMATION AND ARE NOT REGULATORY COMMITMENTS. PLEASE NOTIFY THE MANAGER - REGULATORY AFFAIRS (419-321-8450) AT THE DBNPS OF ANY QUESTIONS REGARDING THIS DOCUMENT OR ANY ASSOCIATED REGULATORY COMMITMENTS.

COMMITMENTS DUE DATE Relocate Surveillance Requirement (SR) 4.5.2.f to the Upon implementation of the DBNPS Updated Safety Analysis Report (USAR) approved license Technical Requirements Manual (TRM).

amendment.