ML19263E730

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Safety Evaluation Supporting Amends 51,45 & 23 to Licenses DPR-33,DPR-52 & DPR-68,respectively
ML19263E730
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 05/11/1979
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML19263E725 List:
References
NUDOCS 7906250099
Download: ML19263E730 (10)
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Text

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) ' l WASHINGTON, D. C. 20555 Ogv j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 51 TO FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT NO. 45 TO FACILITY OPERATING LICENSE N0. DPR-52 AMENDMENT NO. 23 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS NOS. 1, 2 AND 3 DOCKET NOS. 50-259, 50-260 AND 50-296 1.0 Introduction By letter dated December 28, 1977 and supplemented by letter dated December 13, 1978, the Tennessee Valley Authority (the licensee or TVA) submitted proposed design modifications for the Low Pressure Coolant Injection (LPCI) systems of the Browns Ferry Nuclear Plant, Units Nos.1, 2 and 3 and a detailed safety analysis of the modification.

For Units Nos.1 and 2 the licensee's proposed modification consists of changing the power supply to the motor operators of certain LPCI system valves. The change involves the use of Class TE motor-generator sets as isolation devices between the swing bus of the 480 V reactor MOV boards that supply power to the valve operators and the divisional 480 V shutdown boards.

For Unit No. 3, the licensee's proposed modification consists of the following:

a.

Elimination of the Low Pressure Coolant Injection (LPCI) system's recirculation loop selection logic, revision of the logic and closure of the Residual Heat Removal (RHR) cross-tie valve and a recirculation equalizer valve; and b.

Changing the power supply to the recirculation pump discharge valves, LPCI injection valves, RHR pump minimum flow bypass valves, and RHR test isolation valves.

The change also modifies 2218 238 19o82s0 06

. independent valve a.c. power supplies to eliminate concerns on paralleling of divisional a.c. power supplies, and modifies d.c.

power supplies to 4kV shutdown board control.

Item a, above, of the proposed modification is similar to those modifi-cations previously approved by the staff and implemented at the Browns Ferry Nuclear Plant, Units 1 and 2 (BFNP-1 & 2) (Amendments Nos. 27 and 24 for Units Nos.1 and 2, respectively, dated August 20, 1976).

Item b, above, is in line with the modi fications proposed herein for Units Nos. I and 2.

2.0 Discussion 1.

Units Nos. 1 and 2 The proposed design modification to the LPCI systems of the Browns Ferry Nuclear Plant, Units 1 and 2, (BFNP 1 & 2) was submitted in accordance with a licensee agreement made when Amendment No. 27 to Facility License No. DPR-33 and Amendment No. 24 to Facility License No. DPR-52 were issued.

The amendments, in part, dealt with the elimination of the LPCI system's recirculation loop selection logic and closure of the RHR cross-tie valve. The proposed modification is designed to assure that the.480 V ac reactor M9V boards, with the associated auto-transfer feature, will be isolated from the redundant divisional power supplies.

In the existing LPCI system, redundant LPCI injection valves, recirculation pump discharge valves, RHR pump minimum flow valves and RHR test isolation valves are connected to separate 480 V reactor MOV boards and are supplied power from redundant power supplies, i.e.,

Diesel Generator A and Diesel Generator C for Unit 1 and Diesel Generator 8 and Diesel Generator D for Unit No. 2.

The design is such that upon loss of normal power supply to a reactor M0V board, an automatic transfer connects the MOV board to its redundant power supply.

This automatic transfer scheme compromises the independence between redundant power supplies.

Although analysis had shown that no single failure in the interlocks that effect automatic transfer between divisions would adversely affect redundant divisions of power supply, the licensee had committed to modify the power dis-tribution system to eliminate the concerns regarding paralleling of redundant power supplies.

2218 239

. The proposed modification changes the power supply to the reactor M0V boards that feed the motor operators of the LPCI injection valves, the recirculation pump discharge valve. and the RHR pump minimum flow bypass valves.

The change involves c e use of Class lE motor-generator (M-G) sets as isolation devices between the auto-transfer feature of the 480 V reactor MOV boards and the divisional 480 V shutdown boards.

The auto-transfer feature will be eliminated from all 480 V reactor M0V boards not protected by M-G sets.

As part of the modified LPCI system power supply the licensee will provide redundant M-G sets between the divisional 480 V shutdown boards and each 480 V reactor MOV board. The auto-transfer feature of each 480 V reactor MOV board will be retained.

For example, for BFNP-1, 480 V Reactor MOV Board 10 will be normally supplied by motor-generator set MG-lDN, connected to 480 V Shutdown Board 1A, with alternate supply to be available from M-G set MG-lDA connected to 483 V Reactor MOV Board 1B.

Similarly, 480 V Reactor MOV Board lE will have MG-lEN as the normal power supply source connected to 480 V Shutdown Board 1B, and MG-LEA as the alternate source connected to 480 V Shutdown Board 1A, The arrangement of the 480 V reactor MOV boards and the M-G sets for BFNP-2 will be similar.

The M-G sets are designated as Class lE equipment and will be derigned to seismic Category I standards.

Each M-G set will be sized to accept the load requirements of the valve operators at any time during the initiating event. Tne sets will be designed to operate within design specifications when supplied by the diesel generators.

Each M-G set will act as an isolating device between the 480 V shutdown board and the reactor MOV board.

Overload protection of the motor and generator will be separately provided for each set.

Control for the M-G sets will be at the 483 V shutdown boards and loss of each M-G set output voltage will be annunciated in the main control room.

Although only one M-G set will normally supply power to each 480 V reactor MOV board, both M-G sets will run at all times to assure readiness of the alternate M-G set to accept load.

The auto-transfer feature of each reactor MOV board will be retained to assure power to the valve operators.

The auto-transfer scheme has already been analysed to ensure that a single failure in the circuit will not affect redundant divisions of power.

The insertion of isolation M-G sets between the Reactor M0V boards and the shutdown boards provides added assurance of independence between redundant divisions of power supply.

For the reactor M0V boards whic;h do not have the auto-transfer feature interlocks between the divisional supply breakers, use of redundant series feeder breakers at the 480 V shutdown boards assure that a single failure will not compromise divisional power supplies.

2218 240 The RHR test isolation valve operators were not included in the previous modification of the LPCI system. By the proposed modifica-tions, the RHR test isolation valves will be provided with redundant power supplies and closing control logics.

2.

Unit No. 3 The existing Low Pressure Coolant Injection (LPCI) mode of the RHR system at the Browns Ferry Nuclear Plant, Unit 3 (B'FNP-3) is the standard ~

BWR-4 configuration, using four pumps and a loop selection logic.

The LPCI injection valves are closed and the RHR cross-tie valve is open

~

during normal operations.

On receipt of an accident initiation signal follow:..g a recirculation line break in one loop, the loop selection valve in that loop is signaled to open, the recirculation pump discharge valve in that loop is signaled to close, and LPCI flow from all four pumps is directed to the unbroken loop.

The proposed modification to the LPCI system of BFP-3 was submitted by the licensee to inprove the Emergency Core Cooling System (ECCS). The proposed modification involves the following changes:

a) The recirculation loop selection logic will be eliminated, and the accident initiation signals will be rewired to direct both LPCI injection valves to open upon detection of accident conditions.

b)

Both recirculation loop discharge valves will te signaled to close when the reactor pressure decreases to an appropriate setting following detection of accident conditions.

c) The cross-tie valve between the two RHR system headers and a recirculation equalizer valve will be kept closed and an annunciator added to indicate any not-fully-closed conditicn.

The motive power to the valves will also be disconnected.

d)

The auto-transfer feature of the 480 V reactor MOV boards that supply motive power to the LPCI valves, will be isolated from redundant divisional power sources by motor-generator (M-G) sets.

e) A qualified battery that supplies de control p:wer to a 4kV shutdown board will be added, and the dc control power source to another 4kV shutdown board will be changed.

2218 241 3.0 Evaluation 1.

Units Nos. 1 and 2 The proposed modification has been designed to seismic Category I standards.

The modified system design has been reviewed against the standards and guides which were applicable to the original design to assure that the modified system design, equipment and installation meet or exceed the qualification of the unmodified system, including seismic qualification.

The licensee has committed to apply quality assurance and control to this modification in accordance with the requirements of 10 CFR 50, Appendix B.

The licensee has also submitted certain analyses which we have reviewed that were performed in accordance with the requirements of 10 CFR 50, Appendix K to consider the emergency core cooling performance with operation of the modified power systems.

Based on the analysis for a recirculation loop pipeline break, the limiting single failure (which resulted in the highest peak cladding temperature

[ PCT]), was seen to be the suction line break with the failure of the RHR injection valve in the unbroken loop.

For this condition, the resulting peak clad temperature (PCT) was below the allowable PCT limit.

(Table 1 shows the ECCS pump configuration for various postulated single failures).

We have also reviewed the licensee's analysis of the single failure which might influence the long-term suppression pool cooling mode of the modified RHR system.

For the worst case single failure, the suppression pool temperature was found to be within allowable limits.

The analysis and evaluation were done to assure that the changes do not introduce adverse effects to the overall plant. This investigation considered the effects on the capability of major affected equipmcnt (e.g., Diesel Generators, dc batteries, RHR pumps, and RHR system valves) and on the operating modes of the affected equipment (Diesel Generator Control, RHR Logic Panels and DC Control Power).

Based on our review of the proposed modifications to Units Nos. 1 and 2, we find that:

The proposed modification to the Low Pressure Coolant Injection a.

(LPCI) system will assure that the relevant sections of the onsite power system have sufficient independence between redundant power sources, and thus meets the requirements of General Design Criterion 17.

This has been accomplished by the provision of redundant Class lE motor-generator sets to act as isolation power supply devices between the 480 V shutdown boards and the 480 V reactor MOV boards.

2218 242

. b. The additional analyses performed in accordance with the requirements of 10 CFR 50, Appendix K, to consider ECCS performance with the modified power have confimed that for the design basis LOCA and assuming the worst single failure, the peak cladding temperature is below the allowable limit.

Tne proposed cha~nges do not introduce adverse effects to c.

t,e overall plant, Accordingly, we conclude that the proposed design modifications for Units Nos. 1 and 2 are acceptable.

2.

Unit No. 3 The proposed modifications to LPCI system for Unit No. 3 were described in the Discussion section. Our evaluation of each proposed modification is sumarized below.

a.

Elimination of Loop Selection Logic Elimination of the loop selection logic and rewiring of the logic circuitry will direct both LPCI injection valves to open, irrespective of the location of the break in the recirculation loop.

The start logic of the RHR pumps will be changed by the addition of redundant start comands to all pumps and the operating modes will be changed such that two pumps discharge to each injection header.

The wiring changes for elimination of tha

'op selection logic and the rewiring required are to the same s L.idards applied to the original design. All standards for engineered safety feature control equipment will also be maintained.

Additional relays and wiring will be added to assure single failure capability.

Orifices for additional flow resistance will be installed in the RHR pump discharge lines to limit the maximum pump flow when the RHR pumps discharge to the broken loop. The information obtained from tests that were conducted on similar pumps of BFNP-1 and BFNP-2 will be used to determine the additional resistance to be added on the discharge side of each BFNP-3 pump to ensure that the pumps' Net Positive Suction Head (NPSH) requirements are satisfied.

2218 243

. b.

Recirculation Loop Valves Recirculation pump discharge valve closure requires both a LOCA initiation signal and a decrease in reactor pressure to the permissive setting.

With valve closure initiation delayed until reactor pressure has decayed to less than 225 psig, the differential pressure across the closed valve will always be less than 200 psia, (i.e., within the capability of the valve).

The sensor and permissive circuitry will be designed to satisfy all requirements for engineered safety feature control systems.

RHR Cross - Tie Valve and Recirculation Loop Equalizer Valve c.

The RHR system cross-tie valve and a recirculation loop equalizer valve will be kept closed and motive power to the valve operators removed to prevent any inadvertent opening of these valves that could negate RHR system injection when needed.

d.

Motor-Generator Sets Qualified motor-generator (M-G) sets will be used as isolation devices on the 480 V reactor MOV boards with auto-transfer feature. These F0V boards supply motive power to those valves necessary for automatic operation of RHR injection (recir-culation pump discharge valves, LPCI injection valves, RHR pump minimum flow bypass valves and RHR test isolation valves) and will interface with the divisionalized 480 V shutdown boards through the M-G sets.

Each MOV board will have two sets and althou

~~

~

only one M-G set will normally supply power to the MOV board,ghboth~

M-G sets will run at all times to assure readiness of the alternate M-G set to accept load if required.

Each M-G set will be sized to accept valve loads at any time during the initiating event.

The M-G sets will be designed to operate within design specifications when supplied by the diesel generators.

Overload protection of the motor and oenerator will be separately provided for eut s6t Control for the M-G sets will be at the 480 V shutdown boards and loss of each M-G set output voltage will be annunciated in the main control room.

The auto-transfer feature of the reactor MOV boards will be retained to assure power tc the valve operators.

The auto-transfer scheme has already been analysed to assure that a single failure in the circuit will not affect redundant divisions of power.

For those reactor MOV boards which do not have the auto-tra 5 er scheme, interlocks between the divisional supply breakers and use of redundant feeder breakers at the 480 V shutdown boards assure that a single failure will flot compromise divisional power supplies.

2218 244

. e.

250 V D.C. Battery A separate qualified 25J V de battery will be added to provide control power to 4kV Shutdown Board 3EB.

The 250 V de control power source to 4kV Shutdown Board 3ED will be changed to a different station battery.

These changes will assure that control power for all four 4kV shutdown board breakers are fed from different sources to meet the single failure criterion.

The design of tne proposed battery will be to original standards for similar batteries on BFNP-1 and BFNP-2.

Ventilation and fire protection systems will be provided for the proposed battery as required.

The proposed modifications for Unit No. 3 have been desig.'ed to seismic Category I standards. We have reviewed the modified system design against the standards and guides which were applicable to the original design and have assured that the modified system design, equipment and installation meet or exceed the qualification of the unmodified system, including seismic qualification.

The licensee has committed to apply quality assurance and control to these modifications in accordance with the requirements of 10 CFR 50, Appendix 8.

The licensee has also submitted certain analyses which we han reviewed that were performed in accordance with the requirements of 10 CFR 50, Appendix K to consider the emergency core cooling performance with operation of the modified power system.

Based on the analysis for a recirculation loop pipeline break, the limiting single failure (which resulted in the highest peak cladding temperature [ PCT]), was seen to be the suction line break with the postulated failure of the RHR injection valve in the unbroken loop.

For this condition, the resulting peak clad temperature (PCT) was below the allowable PCT limit.

(Table 2 shows the ECCS pump configuration for various postulated single failures).

The single failure which might influence the long-term suppression pool cooling mode of the modified RHR system has also been analyzed.

For the worst case single failure, the suppression pool temperature was found to be within allowable limits. The analyris and evaluation were done to assure that the changes do not introduce adverse effects to the overall plant.

This investigation considered the effects on the capability of major affected equipment (e.g., Diesel Generators, dc batteries, RHR pumps, and RHR system valves) and on the operating modes of the affected equipment (Diesel Generator Control, RHR Logic Panels and DC Control Power).

2218 245

. Based on our review of the proposed modifications to the LPCI system for Unit No. 3 we find that:

The elimination of the LPCI system loop selection logic and rewiring a.

of the ECCS initiation signals will assure that the modified circuits meet the single failure criteria.

This has been accomplished by the use of redundant and separate relays and wiring in the RHR logic panels.

b.

For the recirculation loop discharge valve, the valve closure initiation is delayed until reactor pressure has cecayed to less thar. 225 psig.

This ensures that the differential pressure across tne cicsed valve will always be less than 200 psic.

Further, the sensor anc cermissive circuitries are designed to satisfy all re-cuirements for engineered safety feature control systems.

The proposed revisions assure that the RHR cross-tie valve and a c.

recirculation loop equalizer valve will remain closed during normal plant operations and accident conditions.

This has been done by disconnecting valve motive power with the valves closed and by providing an annunciator to indicate when a valve is not fully closed.

d.

The provision of redundant Class IE moto -generator sets which function as isolation devices between the 480 V shutdown boards and the new 480 V reactor MOV boards will assure that the relevant sections of the onsite power systems have sufficient independence between redundant power sources.

The addition of a qualified battery to supply control power to a e.

4kV shutdown board and tne change of de control power source of another 4kV shutdown board will assure that the modified LPCI system will meet the postulated single failure of a de power source.

f.

The proposed changes do not introduce adverse effects to the overall plant.

g.

The modifications brings the BFNP-3 Emergency Core Cooling System (ECCS) in line with the ECCS of BFNP-1 and BFNP-2.

Accordingly, we conclude that the proposed design modifications for Unit No. 3 are acceptable.

2218 245 4.0 Environmental Considerations We have detennined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this detennination, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR Section Sl.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.

5.0 Conclusion We have concluded that:

(1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the comon defense and security or to the health and cafety of the public.

Date1: May 11, 1979 2218 247

TABLE 1 ECCS PUMP CONFIGURATION Pumps Available**

Suction Side Break 4 Core Spray, 2 RHR in each Loop No Failures Opposite Unit Spurious Accident 2 Core Spray, 1 RHR in each Loop Signal 2 RRR in one Loop RHR Injection Valve Failure

  • 480 V Reactor MOV Board

4 Core Spray, 2 RHR in one Loop No Failures RHR Injection Valve Failure

  • Limiting Single Failure
    • In Unbroken Loop

TABLE 2 ECCS PUMP CONFIGURATION IN MODIFIED SYSTEM Suction Side Break Pumps Available**

No Failures 4 Core Spray, 2 LPCI in each Loop LPCI Injection Valve Failure

  • 480 V Reactor MOV Board Failure

No Failures 4 Core Spray, 2 LPCI in one Loop LPCI Injection Valve Failure

  • 4 Core. Spray 480 V Reactor MOV Board Failure
  • Limiting Single Failure
  • *In Unbroken Loop
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