ML20215K741
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(jCMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT ~ 6201 S Street, P.O. Box 15830. Sacramento, CA 95813. (916) 452-3211 AN ELECTRIC SYSTEM SERVING THE HEART OF CALIFORNIA JEW 86-152 0Ctober 2 , 1986 Director of Nuclear Reactor Regulation Attention Frank J. Miraglia, Director PWR-B Division US Nuclear Regulatory Commission Washington DC 20555 Docket 50-312 Rancho Seco Nuclear Generating Station Unit Number 1 Proposed Technical Specification - Amendment 147 In accordance with 10CFR50.90, the Sacramento Municipal Utility District proposes to amend its Operating License DPR-54 for Rancho Seco Nuclear Generating Station, Unit 1.
Proposed Amendment No.147 consists of revisions and additions to Section 3.7, " Auxiliary Electrical Systems" and Section 4.6,
" Emergency Power System Periodic Testing" of the Rancho Seco Technical Specifications to permit the operation of two Transamerica Delaval (TDI) Diesel Generators to provide additional onsite emergency power as previously described in a November 26, 1980 District letter to R.W. Reid (NRR) . Additional sections are revised or added to address the impact on other systems when the TDI Diesel Generators are brought into service.
In addition, the District h'as performed a detailed review of existing Sections 3.7 and 4.6 and proposed a number of revisions that reduce the number of diesel fast starts during surveillances and implement some of the recommendations of Generic Letter (GL) 84-15, " Proposed Staff Action to Improve and Maintain Diesel Gener-ator Reliability." These revisions are designed to reduce unnec-essary wear on the diesel generators and ensure the maintenance of a reliable system.
These changes are discussed in detail in Enclosure 1 which is the Description of Proposed Changes, Associated Safety Analysis, and the "No Significant Hazards Evaluation". Enclosure 2 is the pro-posed technical specification.
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Frank J. Diraglia, Director , 1986 To assist in the review, the District is providing design information that supports this proposed technical specification amendment. The enclosures include the following information:
o Enclosure 3 - Summary of Diesel Generator Modification and Present Status of Emergency Power; o Enclosure 4 - Design Basis Report (DBR) for ECN A-3748, Rev. 5 - the installation of the TDI diesel generators, the associated building and supporting systems; o Enclosure 5 - A summary engineering report and the DBR for ECN A-3660, Rev. 5 - the expansion of the electrical distribution system including tie-in of the new diesel generators. (The DBR for ECN A-3660, Rev. 4, was submitted by a March 21, 1985 District Letter.);
o Enclosure 6 - A comparison of the TDI diesel system and electrical tie-in designs to the latest NRC requirements of the Standard Review Plan (SRP) in NUREG-0800; o Enclosure 7 - A Fire Hazards Analysis Report (FHAR) for the diesel generator building and o Enclosure 8 - A set of drawings supporting the modification.
- The District is evaluating the following additional changes to the i electrical distribution system, these changes will require a i
revision to this submittal.
l 1. Removal of Upper Dome Air Circulator f rom emergency power l (Proposed technical specification amendment 121, I
submitted 6/18/86)
- 2. The auxiliary feed water pumps and associated loads will be transferred from the existing diesel to the new TDI Diesel Generators. (see LER 86-17) l
- 3. Revision to inverter power supply systems to improve reliability.
These changes will not have a significant impact on the proposed
, technical specifications and the revisions will be submitted by l
November 14, 1986.
The District has decided to install the TDI diesel generators and implement changes associated with its electrical distribution system during the current extended outage. Therefore, NRC approval of proposed amendment is required prior to startup.
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Pursuant to 10CFR50.91(b) (1), the Radiological Health Branch of the California State Department of Health Services has been informed of this proposed amendment by mailed copy of this submittal.
Enclosed is a check for $150.00 as required by 10CFR170.21,
" Statement of Fees."
If you have any questions concerning this submittal, please contact Mr. Ron Colombo of ny staff at Rancho Seco Nuclear Generating Station.
e J. . Ward Subscribed and sworn to before me Deputy General Manager this 1 *Aday of Ottehe (,1986.
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Enclosure 1 Description of Proposed Changed, Associated Safety Analysis, and the "No Significant Hazards Evaluation."
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s 4 FACILITY CHANGE SAFETY ANALYSIS PROPOSED AMENDMENT 147 LOG NO. 760 Descrintion:
9 Proposed Amendment 147 consists of changes and additions to the Rancho Seco Technical Specifications to permit operation of two TDI diesel generators in the essential electrical system.
Reason 12I Chanae:
The safety analysis /50.59 review for ECN A-3660 Rev. 5 and ECN .
A-3748 Rev. 5 determined that a Tecnnical Specification change was required for these Mods.
Evaluation and Basis inz Safety Findings:
The following is a discussion of each Technical Specification change. The changes are divided into three groups or sections depending on the type of change. Each section~ lists the existing specification, the new specification, and a discussion of the change. Each section also includes a safety analysis and a no significant hazards consideration for the changes. Also, attached are copies of the proposed -amendment and the corresponding pages from the existing Technical Specifications.
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Section I - Changes that arm within the design basis inz Rancho Seco
- 1. Existing Specifications:
3.7.1.D. Emergency diesel generators A and.B are ,
operable and at least 35,000 gallons of fuel are in each storage tank.
Bases:
All of the normal power supply to plant auxiliary loads is provided through the two unit auxiliary transformers connected to the generator buses.
Emergency power for the nuclear service loads is obtained from two on-site diesel generators.
The 35,000 gallons of fuel stored in each storage tank permit operation of the two diesel generators for seven days.
Hgw Spacification: ;
3.7.1.F. Two separate and independent diesel generator trains (train A is both diesel generators A and A2, train B is both diesel generators B and B2) are operable each with:
- 1. Separate day tanks containing a minimum volume of 65 percent of tank capacity (265 gallons) of fuel for each A and B and 50 percent of tank capacity (135 gallons) of fuel for A2 and B2.
- 2. A separate fuel storage system containing a minimum usable volume of 35,000 gallons of fuel ror each A and B and 42,000 gallons of fuel for each A2 and B2.
- 3. A separate fuel transfer pump for each engine of a train.
Bases:
All of the normal power supply to plant auxiliary-loads can be provided through the two unit auxiliary transformers connected to the generator buses.
Emergency power for the nuclear service loads is obtained from two on-site diesel generator trains (train A is both diesel generators A and A2, train B is both diesel generators B and B2).
The 35,000 gallons of fuel stored in each storage tank permit operation of the diesel generators A and B for seven days. The 42,000 gallons of fuel stored in each 2
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storage tank permit operation of the diesel generators A2 and B2 for seven days.
Discussion:
This change adds the two new TDI diesel generators A2 -
and B2. Diesel generators A and A2 (one old and one new) comprise train A and diesel generators B and B2 (one old and one new) comprise train B. All diesel generator specificationa are written using this train approach to simplify the required limiting condition for
. operation for the operating staff. This is a conservative approach since there are only limited instances where a safety function cannot be provided by only one old diesel generator per train.
The requirements for the day tanks and the fuel transfer pumps for each diesel generator are new additions to the specification. These support systems are presently required to be operable by plant procedures. This change is consistent with the recommendations of Generic Letter 84-15.
- 2. Existing Specification:
3.7.1.K. The interconnections between 480 volt switchgear 3A and 3A2, and 3B and 3B2 are operable.
New Specification:
N/A Discussion:
This specification has been proposed to be deleted. This requirement was originally added to reflect the Cycle 7 inte!'.3 power supply configuration. Buses 3A2 and 3B2 will be powered by the new TDI diesel generators and will no longer require the interconnect.
- 3. Existing Specification:
3.7.1.E. Plant batteries are charged and in service.
3.7. Bases
The set of four 125 volt DC control panelboards (SOA, SOB, SOC, SOD) and the set of two 125 volt DC control panelboards (SOA2, SOB 2) are arranged so that loss of one bus will not preclude safe shutdown or operation of safety features systems.
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Hgw Specification:
3.7.1.G. Nuclear Service batteries BA, BB, BC, BD, BA2, BB2, BC2 and BD2, which supply vital 125 volt buses SOA, SOB. SOC, SOD, SOA2, SOB 2, SOC 2, and SOD 2 are charged and in service. >
3.7 Bases
The set of eight 125 volt DC control panelboards (SOA, SOB, SOC, SOD, SOA2 SOB 2, SOC 2, and SOD 2) are arranged so that loss of one bus will not preclude safe shutdown or operation of safety features systems.
Discussion:
Proposed Amendment 135 dated June 13, 1986 revises Specification 3.7.1.E shown to 3.7.1.G (above) without batteries BC2 and BD2 and buses SOC 2 and SOD 2. This change adds batteries BC2 and BD2 and buses SOC 2 and SOD 2. These batteries and buses are being added to support loads being added to the electrical distribution system. This change maintains consistency in the specifications by listing all nuclear service batteries and the vital buses they supply.
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- 4. Existing SpecificationI:
N/A Hgw Specification:
Table 3.14-1 FIRE DETECTION INSTRUMENTS FOR SAFETY SYSTEMS Minimum Operable Zone Instrument Location Heat Flame Smoke 105 Diesel Generator Building 0 3 6 106 Diesel Generator Building 0 3 6 3.14.3.1 m. Dic:cl Generator Building (Zone 105)
- n. Diesel Generator Building (Zone 106) 4
- - . . _ . - - . ._ .- - -.. .- - . .~. ._ -
3.14-2 INSIDE BUILDING FIRE HOSE STATIONS i
. .c ID No. Location III Diesel Generator Building Rose Stations FPHS-D-001 Train A Engine Room Elevation O' FPHS-D-002 Train A Control Room Elevation O' FPHS-D-003 Train A Mezzanine Elevation 18'6" FPHS-D-004 Train B Engine Room Elevation O' FPHS-D-005 Train B Control Room Elevation O' FFHS-D-006 Train B Mezzanine Elevation 18' 6" Discussion:
These changes add the fire detection instruments, the sprinkler systems, and the fire hose stations in the new diesel generator building. These are additions to the
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specifications that provide fire protection features for portions of redundant systems important to safe shutdown as described in the Rancho Seco Fire Hazard Analysis -
Report.
- 5. Existing Specification:
3.7.1.H. Three out of four inverters S1A, S1B, S10, and S1D, and both inverters S1A2 and S1B2 are operable for 120 volt AC vital bus power.
Bases-3.7:
l Sufficient redundancy is available with any three of the l - four 120 volt AC vital power-buses (S1A, S1B, S10, S1D) in service such that reactor safety is assured. Every reasonable effort will be made to maintain all safety instrumentation in operation. Following criticality, continued operation with inverters out-of-service as stated in Specification 3.7.1.H is governed by the individual LCOs for the components powered by the out-of-service inverter.
New specification:
3.7.1.I. Nuclear service inverters S1A, S1B, S1C, S1D, I
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S1A2, S1B2, S1C2, and S1D2, are operable for 120 volt AC vital bus power.
3.7.2.K. Nuclear service inverters as listed in 3.7.1.I shall be operable except should one inverter become inoperable for greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, o the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If more than one nuclear service inverter as listed in 3.7.1.I should become inoperable, within one hour take action to place the reactor in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Bases-3.7:
Eight nuclear service inverters (S1A, S1B, S1C, S1D, S1A2, S1B2, S1C2, and S1D2) are required to be operable to power the 120 volt AC vital buses. One nuclear service inverter is permitted to be inoperable for up to eight hours.
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Discussion:
This change identifies all of the nuclear service inverters required to be operable for criticality. It incorporates the' additional inverters added by the electrical distribution system upgrade. The existing specification allows:one inverter to be out of service indefinitely. The new specification requires that all the listed inverters be operable for startup and defines the amount of time inverters can be inoperable prior to shutting down the plant.
- 6. Existing Specification:
3.7.1.F. Two out of three battery chargers are operable for-125 volt DC buses "A" and "C", and "B"
. thru "D".
3.7.1.G. One out of two battery chargers are operable for each 125 VDC bus "A2" and "B2."
Egw Specification:
3.7.1 H. Each vital 125 volt DC bus SOA, SOB, SOC, SOD, SOA2, SOB 2, SOC 2, and SOD 2 shall have an operarable battery charger aligned to it.
Bases:
The 125 volt DC buses SOA and SOC, SOB and SOD, SOA2 and SOC 2, and SOB 2 and SOD 2, have a. standby battery charger.
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Loss of power from one battery charger per pair of redundant DC buses has no significant consequence since a standby battery charger is available.
Discussion:
This change adds the battery chargers for vital 125 volt DC buses SOC 2 and SOD 2 which are not presently in the specifications. The change is revised to only address the nuclear service battery chargers required for reactor criticality and does not address the installed spare battery chargers. The change in designation for vital buses (A,B,C,D,A2,B2 vs. SOA, SOB, SOC, SOD, SOA2, SOB 2, SOC 2, SOD 2) is addressed in proposed amendment 135 dated June 13,,1986.
- 7. Existing Specification:
N/A New Specification:
3.28 IDI Diesel Generator Control R99m Essential Ventilation System Applicability Applicability This specification applies to the operability of the TDI. Diesel Generator Control Room Essential Ventilation System.
Ob.iective To assure that this system will be able to perform its design function.
Specification
, 3.28.1 Both TDI Diesel Generator Control Room Essential Ventilation trains shall be operable at all times except as noted in 3.28.2 and 3.28.3.
3.28.2 With one TDI Diesel Generator Control Room Essential Ventilation train inoperable.
demonstrate the operability of the remaining train. Restore the inoperable train to operable status within 15 days or be in at least hot shutdown within six hours and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.28.3 With both TDI Diesel Generator Control Room Essential Ventilation trains inoperable, restore at least one inoperable train to 7
operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least hot shutdown within six hours and in cold.shutdowa within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With one inoperable train restored to operable status, follow 3.28.2.
Be.ses The TDI Diesel Generator Control Room Essential Ventilation System is required to provide cooling whenever the temperature in the control room reaches 122 F to protect required electrical components.
Discussion:
This new specification is added to assure operability of this ventilation system. The ventilation system is required to provide cooling for the electrical couponents in the TDI Diesel Generator Control Room and not human Habitability considerations.
Safety Analysis far Items in Section I.i The addition of the TDI diesel generators and the modification of the electrical distribution system, results in a two train, two diesel generator per train limiting condition for operation power-system. The District has reviewed this modification in accordance with 10 CFR 50.59 and determined that it required a change in the technical specifications, but was not an unreviewed safety question.
A review of the Licensing documents for Rancho Seco show the following:
- 1. The Bases for Technical Specification 3.7, A.uxiliary Electrical Systems states "The auxiliary electrical power systems are arranged so that no single failure can inactivate enough safety features equipment to jeopardize plant safety."
- 2. USAR section 8.2.3.4 Single Failure Analysis 12I tha Nuclear' Service System, states "...a single fault within the system, with postulated loss of offsite power combined with a design basis accident, does not preclude the reactor protective system, safety features actuation system, and the safety features equipment from performing their safety function."
- 3. USAR section 14.1.2.8 Loss 21 Electric Power states' that "The unit is designed to withstand the effects of load rejection or loss of all ac power.
Emergency systems are described in Section 8.2.3."
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Based on this review, it was concluded that the design basis for the emergency power system is that a single failure of the system (including diesel generators) will i not preclude the reactor protection system and safety feature system from performing their intended safety function.
The loads on the new diesel generators were evaluated to determine the effects of loss of one diesel generator and associated power distribution system in the two diesel generator per train configuration. The systems interaction evaluation included a review of system components with regard to the r power supplies and whether failure of components within a system could escalate or aggravate an event or present confusing information to the control room operators (i.e., failure of one diesel generator in a train would not cause confusion or an unacceptable plant configuration). The evaluation found that the diesel generators and associated power distribution systems for each train were independent and that no unacceptable interactions exist.
The following four sets of electrical loads were identified as system interactions. The first three items are cases of redundant electrical components associated with common components, hence, systems interaction is essentially unavoidable. The last item, is a more " classic" case of systems interaction.
However, each of these cases were individually evaluated and no adverse system interactions were identified.
- 1. Containment and Containment Emergency Sump Level Inputs tn Computer. (LIT-20509 and -20510 and LSH-26112C and D1m These components provide electrically isolated SPDS and plant computer
- inputs only. Because they are associated with the containment which is common to redundant components, including components which are fed by either the A or B diesel generator (e.g., LPI system), these devices have been identified. However, there are alternate and redundant level devices which are available to the plant operators in the control room 1
and their failure would not provide any confusing information to the plant operators.
- 2. Reactor Coolant System High Point Vent Valves (HV-21522. -21528. -20579. -20533. -20534. -20580).
The vent valves are powered from GEB2 for pressurizer, GEA2 for A steam generator and GEB2 for B steam generator. Loss of the GEA2 power loses the capability to open "A" OTSG vent valves and requires that the redundant vent path ("B" OTSG) be used. On loss of GEB2 power a converse situation exists for the B train and, in addition,
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the pressurize venting function is lost. They are
. actuated manually based primarily on operator evaluation of incore temperatures from the Reactor Coolant System. Redundant incore thermocouples are powered from Class 1 GEA2 and GEB2 supplied batteries respectively and are fed.to SPDS. The OTSG 1evel and pressure are also key parameters which are used by the operator for venting. OTSG level and pressure readout in the control room depend on NNI-X and SPDS. The Class 2 NNI-X has a
- non redundant Class 1 " normal" power source fed by GEB power. It has Class ,2 " backup" power fed by GEA. Therefore, the venting function is redundant and there is no detrimental _effect on system operation from the four diesel arrangement.
- 3. Pressurizer Emergency Heaters (V-215L-13 through 21 and V-215M-10 through 21Hl. These heaters are associated with maintaining adequate pressure in the common RCS. The RCS pressure indication is the key variable used to manually control the heaters.
. The pressure instrumentation is supplied by A or B power and displayed on SPDS. Since both the heaters and the pressure instruments are redundant components controlling a commen RCS, loss of a power supply still" permits operation of the heaters and control of the RCS pressure without unusually confusing information.
- 4. Hydrogen Monitoring Analyzer Panels (H4HMA. H4HMB).
These panels are supplied by either the A2 and B2 diesel generators while the hydrogen sample pumps are supplied by either A or B unit, respectively.
However, based on failure analyses of each individual power supply to each hydrogen monitor system, the hydrogen monitors will go downscale to zero. This same downscale reading would result even if each' system was fed by a single power source. Hence, there is no confusing information presented to the control room operators.
Based on this review the District concludes that this modification does not change the single failure requirements in USAR Chapter 8 or Technical Specification 3.7 Bases because redundancy of trains and system independence is still maintained. It also does not reduce the margin provided in the accident analysis of USAR Chapter 14 because the additional < racity will support required safety related loads.
Basis for Ho Significant Hazards Determination for Items in Section 11 The proposed changes does not involve a significant hazards consideration because operation of Rancho Seco 10
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in accordance with this change would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated. This modification does not significantly alter the accident analysis in Chapters and 14 of the USAR. The modification of the electrical distribution system was designed to meet single failure criteria and withstand the effects of load rejection. The systems interaction evaluation concluded that a failure of one diesel generator and associated power distribution system would not introduce any unacceptable interactions or any failures in the remaining electrical distribution with its train. Therefore, this change does not increase the probability or consequences of an accident.
(2) create the possibility of a new or different kind of accident from any previously analyzed. The accident analysis in Chapter 14 of the USAR is not changed because the additional capacity ensures that the power distribution system will support required safety related loads. The systems interaction review shows that no new or different failures modes were created. This modification does not therefore create the possibility of a.new or different kind of accident.
(3) involve a significant reduction in a margin of safety. The design basis for the emergency power system is that a single failure of the system (including diesel generators) will not preclude the reactor protection system and safety features system from performing their safety function. The modification of the emergency power system does not change this basis. Based on the systems interaction review and the design basis documents this change will not reduce the margin of safety.
The modification to the emergency power system will provide redundant emergency power sources for the control room, TSC, and NSEB essential HVAC systems.
It will also provide additional capacity for future
- loads while ensuring that the existing emergency l' power system is not overloaded. This therefore increases the existing margin of safety.
The emergency power system at Rancho Seco is based r
on a dual train (A and B) concept. The existing diesel generators (A and B) primarily support ECCS loads. With the addition of the new diesel generators (A2 and B2) the dual train concept is maintained. The ECCS loads are still on A and B, with A2 and B2 supporting HVAC loads, loads required by NUREG 0737 modifications, and other f
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in accordar.ce with this change would not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated. This modification does not significantly alter the accident analysis in Chapters and 14 of
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the USAR. The modification of the electrical distribution system was designed to meet single failure criteria and withstand the effects of load rejection. The systems interaction evaluation concluded that a failure of one diesel generator and associated power distribution system would not introduce any unacceptable interactions or any failures in the remaining electrical distribution with its train. Therefore, this change does not increase the probability or consequences of an accident.
(2) create the possibility of a new or different kind of accident from any previously analyzed. The accident analysis in Chapter 14 of the USAR is not changed because the additional capacity ensures that the power distribution-system will support required safety related loads. The systems interaction review shows that no new or different failures modes were created. This modification does not therefore create the possibility of a.new or different kind of accident.
(3) involve a significant reduction in a margin of safety. The design basis for the emergency power system is that a single failure of the system (including diesel generators) will not preclude the reactor protection system and safety features system from performing their safety function. The modification of the emergency power system does not change this basis. Based on the systems interaction review and the design basis documents this change will not reduce the margin of safety.
The modification to the emergency power system will provide redundant emergency power sources for the
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control room, TSC, and NSEB essential HVAC systems.
It will also provide additional capacity for future loads while ensuring that the existing emergency power system is not overloaded. This therefore increases the existing margin of safety.
The energency power system at Rancho Seco is based on a dual train (A and B) concept. The existing diesel generators (A and B) primarily support ECCS loads. With the addition of the new diesel generators (A2 and B2) the dual train concept is maintained. The ECCS loads are still on A and B, with A2 and B2 supporting HVAC loads, loads required by NUREG 0737 modifications, and other 11
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loads due to modifications to the plant. This i- change therefore does not involve a significant 1- reduction in the margin of safety.
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Section 11 : Administrative Chanaes:
- 1. Existing Specification:
3.7.1.C. One 6900 volt reactor coolant pump motor bus is energized.
New Specification:
N/A Discussion:
Specification 3.7.1.C has been proposed to be deleted since this specification is already addressed in Specification 3.1.1.1. This change is consistent with the overall format of the Technical Specification and eliminates a redundant requirement.
- 2. Existing Specification:
3.7.2.E. The essential nuclear service electrical buses, switchgear, load shedding, and ,
automatic diesel start systems shall be operable except as provided in C above and as required for surveillance testing.
Egw Specification:
N/A Discussion:
This specification has been proposed to be deleted. The requirements for the buses, switchgear, load shedding and automatic diesel start systems are addressed in Specifications 3.7.1.D, 3.7.1.E, 3.7.1.F, and 3.7.2.
This change is consistent with the overall format of the specifications. ,
- 3. EElating Specification:
3.7.5 The voltage protection system trip setting shall be as stated in Table 3.7-1.
3.7.6 Voltage Protectior. System Limiting Conditions A. Startup and operation are not permitted unless the minimum requirements and action statements of Table 3.7-2 are met.
B. In the event the number of protective channels falls below that listed in Table 13 1
3.7-2, the plant will be brought to a hot shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
New Specification:
3.7.3 The voltage protection system trip setting ,
shall be as stated in Table 3.7-1.
3.7.4 Voltage Protection System Limiting Conditions A. Startup and operation are not permitted unless the minimum requirements and action statements of Table 3.7-2 are met.
B. In the event the number of protective channels falls below the listed in Table 3.7-2, the plant will be brought to a hot shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Discussion:
This changes the numbering of these specifications only, i.e., 3.7.5 to 3.7.3 and 3.7.6 to 3.7.4. Existing Specification 3.7.3 is being incorporated into 3.7.2 and existing specification 3.7.4 is being changed to 3.1.1.7.
- 4. Existing Specification:
Table 3.7-1, Note 1 - The relay voltage values shown have been converted by the PT ratio (40:1) for review convenience.
New Specification:
Table 3.7-1, Note 1 - The relay voltage values shown have been converted to equivalent 4160 bus volts by the PT ratio for review convenience.
Discussion:
The note is revised to clarify the bases for the conversion. The ratio, 40:1, is not^ applicable to all busses. It is not necessary to identify the exact PT ratio, but to understand that the voltage values have been converted using the PT ratio. This does not change the set points.
- 5. Existing Specification:
4.6.6 The pressuriser shall be tested as follows:
A. The pressuriser water level shall be 14
determined to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B. The power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by using the '
Nuclear Service Bus to energize the heaters.
Han Specification:
4.5.2.2.C.The pressurizer shall be tested as follows:
- 1. The pressurizer water level shall be determined to be within its limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 2. The power supply for the nuclear service backed pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by using the Nuclear Service Bus to energize the heaters.
4.5. Bases
The operability of the nuclear service bus backed pressurizer heaters is demonstrated by energizing the heaters once per 18 months.
Discussion:
This change moves this specification from the Emergency Power System Periodic Testing section to the components tests section. It also adds clarification to identify the specific pressurizer heaters to be tested. This change enhances the format of the Technical Specifications.
l
- 6. Existing Specifications:
3.7.1.B. Two 220 KV lines are in service.
i 3.7.1.I. Both startup transformers, No. 1 and 2, are in service.
Egg Specification:
, 3.7.1.A. At least two 220 KV lines are in service.
4 3.7.1.C. Both startup transformers, No. 1 and No. 2, l are in service.
I 15
Discussion:
This change renumbers these specifications to be consistent with the other changes to this section. It also clarifies the requirement for two 220 KV lines to
! make it clear that more than two lines may be in service.
- 7. Existing Soecification:
3.7.1.J. The switchyard voltage is 215 KV or above.
k Egw Specification:
3.7.1.B. The switchyard voltage is 217 KV or above.
1 Discussion:
This change corrects an error in the voltage at which the reactor can be brought critical. The undervoltage 4
trip setpoint listed in Table 3.7-1 results in an equivalent 216.5 KV setpoint (3771V on the 4160V Bus).
Based on this, the reactor cannot be~ brought critical I without the switchyard voltage above 216.5 KV.
- 8. Existing Soecification-:
- 3.7.1.A. All nuclear service buses, nuclear service switchgear, and nuclear service load shedding systems are operable.
1 Egw Specification:
- 3.7.1.D. Nuclear services 4160V buses 4A, 4A2, 4B, and 4B2 are operable.
3.7.1.E. Nuclear services 480V buses 3A, 3A2, 3B, 3B2, 4
2A1, 2B1, 2A3, 2B3, 2A4, and 2B4 are operable.
i Discussion:
This is a clarification.to specifically identify which l nuclear services buses are-required to be operable. The nuclear service switchgear and load shedding systems are
, deleted from this specification. They are part of the
- nuclear service buses and without them being operable, i
the bus is not operable.
- 9. Existing Specification: -
4.31 Bases:
- 16
.g.___.. _
_=__ ._. , ,_
The system is automatically started when the temperature in the NSEB Switchgear Room exceeds 95 F, except itpon loss of offsite power; in which case, the system can be manually started by the operator.
Haw Specification ,
4.31 Bases:
The system is automatically started when the temperature in the NSEB Switchgear Room exceeds 95 F.
Discussion:
The modified electrical distribution system will now automatically start the system without operator action upon loss of offsite power.
- 10. Existing Specification:
4.10 Bases:
During an SEAS and a-loss of offsite power, the "B" Train of essential HVAC equipment is sequenced to automatically start upon its actuation signals approximately 6 minutes after the diesel generator breaker closes. The-"A" Train of HVAC equipment can be manually energized and started as be$$re following automatic application of all essential loads.
Rew Specification:
4.10 Bases:
The filtering system is automatically started and the normal system isolated when the radiation level or when the chlorine level increases.
Discussion:
This section is deleted from the bases. The modified electrical distribution system will now automatically start the system.without operator action upon loss of offsite power.
Safety Analysis 12r Items in Section 111 The changes listed above are administrative changes.
They are made to improve the overall Technical Specification editorial consistency and format, clarify requirements and correct errors.
17
Basis for E2 Significant Hazards Determination:
The proposed change does not involve a significant hazards consideration because operation of Rancho Seco in accordance with this change would pot:
4
- 1) involve a significant increase in the probability or consequences of an accident previously evaluated. This enhanced clarity should decrease the potential for unacceptable consequences or accidents. These are editorial and administrative changes which do not increase the probability or consequences of an accident.
- 2) create the possibility of a new or different kind of accident from any previously evaluated. A new or different kind of accident will not be created due to these editorial and administrative changes.
These administrative changes do not create the possibility of a new or different kind of accident.
- 3) involve a significant reduction in a margin of safety. These editorial and administrative changes ensure that the- Technical Specifications address
- proper plant configuration following the
- installation of
- the new diesel generators and will preserve the margin of safety. Therefore, the administrative changes will not reduce the margin of safety.
18
- , . . . .=_; ..
_ . - . . _ _ = - . - - -
Section III - Increase SI Addition in Surveillance Reauirementa And Limi'tations
- 1. Existing Specification:
3.7.4 The pressurizer shall be OPERABLE with at ,
least 126 kw of pressurizer heaters. With the pressurizer inoperable due to inoperable emergency power supplies to the pressurizer heater either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Bases-3.7:
The requirement that 126 KW of pressurizer heaters and their associated controls being capable of being supplied with electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT SHUTDOWN.
~
Egg Specification:
3.1.1.7 The pressurizer shall be operable, except when the reactor is in cold shutdown, with 3 groups of heaters in two separate banks that are capable of- being powered by the diesel generator trains. With the heaters in one bank inoperable, either restore the bank to operable status within 15 days or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With the heaters in both banks inoperable, either restore one bank to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Bases-3.1:
There are 3 groups of heaters in bank 2, and 3 groups of I heaters in bank 3 that are capable of being powered by
- the diesel generator. Each set of the 3 groups of heaters has a nominal rating of 126 kw which provides assurance that these heaters can be energized during a l loss of offsite power condition to maintain natural
- circulation at HOT SHUTDOWN.
l Discussion:
Specification 3.7.4 is renumbered to 3.1.1.7 (3.7.4 and Bases from 3.7 deleted, 3.1.1.7 and Bases for 3.1 added). This change makes this item cons.istent with the l
19
overall format of the Technical Specifications. The revision to this specification:
- 1) More clearly identifies the requirements for i operability of pressurizer heaters by specifically identifying the number of heaters required. ,
- 2) Reduces the time permitted in a limiting condition for operation (LCO) to be consistent with Specifica. tion 3.7.2.B and 3.7.2.E.
This change more clearly reflects the evaluation made by-i the NRC in their May 1, 1980 letter which evaluated the Districts actions to satisfy NUREG 0578.
l -
i 2.
Existing Specification: ,
3.7.2.A. One 220 KV line shall be fully operational and i capable of carrying nuclear service and auxiliary power except as specified in D below.
Hgw Specification:
3.7.2.A. At least two 220 KV lines shall be in service except should all but one 220 KV line be
- removed from service the operability of the
~
remaining ~220 KV line shall be demonstrated by performing:_ surveillance requirement 4.6.1.A within i hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> '
thereafter and surveillance requirement 4.6.3.A.4 is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If at least two 220 KV lines are not in service within 72. hours, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
j Discussion:
l This change incorporates the recommendations of Generic
- Letter 84-15. It provides clear requirements for number -
of lines in service and defines the actions required if I this requirement is not met. The present Technical Specifications allow indefinite operation with only one l' 220 KV line in service. The new specification limits the time only one 220 KV line is in service.
f
- 3. Existing Specification:
f 3.7.2.C. Both diesel generators shall be operable except that from and after the date that one of the diesel generators is made or found to be inoperable for any reason, reactor 20
operation is permissible for the succeeding 15 days provided that during such 15 days the operable diesel generator shall be load tested daily and both startup transformers are available. If the diesel is not returned to service at the end of 15 days, the other .,
diesel will be started and run with at least minimum load continuously for an additional 15 days. If at the end of the second 15 days the diesel is not returned to service, the reactor shall be brought to the cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Hnw Specification:
3.7.2.B. Both diesel generator trains shall be operable except should one diesel generator train become inoperable the operability of at least two 220 KV lines circuits shall be demonstrated by performing surveillance requirement 4.6.1.A within i hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter and surveillance requirement 4.6.3.A.4 is performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 7 days thereafter.
If the diesel generator train is not restored to operable status within 15 days, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ands in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Q1;cussion:
This citange incorporates the recommendations of IE Information Notice 84-69, Supplement I, " Operation of Emergency Diesel Generators." It eliminates the 15 day period where the operable diesel is run continuously with minimum load. Running the diesel with minimum load may reduce its operability. The surveillance requirements are revised to reduce the number of starts and more clearly define the operability of offsite sources to be consistent with the recommendations of Generic Letter 84-15,
- 4. Existing Specification:
N/A ,
Enw Specification:
3.7.2.C. At least two 220 KV lines and both diesel generator trains shall be operable except should all but one 220 KV line and one diesel generator train both become inoperable the operability of the remaining 220 KV line 21
shall be demonstrated by performing surveillance requirement.4.6.1.A within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, and the diesel generator train shall be demonstrated to be operable by performing surveillance requirement 4.6.3.A.4 within 8 .
hours. If either the 220 KV line or the diesel generator train is not restored to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With the diesel generator train restored to operable status within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, follow 3.7.2.A. With the 220 KV line restored to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, follow 3.7.2.B.
Discussion:
The new specification is being added to clearly identify the requirements if only one 220 KV and one diesel generator train are operable. It also incorporates the recommendation in Generic Letter 84-15.
- 5. Existing Specification:
3.7.2.D. If the plant is separated from the system while carrying its own auxiliaries, or if all 220 KV lines are lost, continued reactor operation is permissible provided that one emergency diesel generator is started and run continuously until a transmission line is restored.
Egw Specification:
3.7.2.D. At least two 220 KV lines shall be in service except should all 220 KV lines become inoperable the operability of the two diesel generator trains shall be demonstrated by performing surveillance requirement 4.6.3.A.4 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This surveillance requirement does not have to be performed if the two diesel generator trains are already in service. If at least one 220 KV line is not returned to service in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. With one off-site circuit restored to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, follow 3.7.2.A.
Discussion:
The existing specification allows continuous operation with the loss all 220 KV lines provided one diesel generator is running continuously. The change 22
incorporates the recommendations of 9eneric Letter 84-15 and IE Information Notice 84-69 by rcquiring restoration of one 220 KV line within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and eliminating the requirement to continuously run a diesel generator.
- 6. Existing Specification: .,
3.7.3 If both diesel generators becone inoperable, the unit shall be placed in the cold shutdown condition.
New Specification:
3.7.2.E. Both diesel generator trains shall be operable except should both trains be inoperable the operability of at least two 220 KV lines shall be demonstrated by performing surveillance requirement 4.6.1.A within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. If one diesel generator train is not restored to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
With one diesel generator train restored to operable status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, follow 3.7.2.B.
Discussion:
This change identifies the specific actions required if both diesel generator trains are inoperable. This change is consistent with the recommendations in Generic Letter 84-15.
- 7. Existing Specification:
3.7.2.B. Both startup transformers shall be in service except that one will be sufficient if during the time one startup transformer is inoperable, the associated diesel generator is started and run continuously.
ERE Specification:
3.7.2.F. Both startup transformers shall be in service except should one startup transformer become inoperable the operability of at least two 220 KV lines shall be demonstrated by performing surveillance 4.6.1.A within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter and the diesel generator trains shall be demonstrated with 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If startup transformer no. 1 is not restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be in hot shutdown within 23
_, _ __ __ .._._7_. _
l the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 205 hours0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br />. If startup transformer no. 2 is not restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be r in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in.
cold shutdown within the following 205 hours0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br />. ,
Discussion:
This change provides operability requirements for the startup transformers that eliminates the requirement to continuously run a diesel generator. This is consistent with the recommendations in Generic Letter 84-15.
- 8. Existing Specification:
N/A New Specification:
3.7.2.G. If both startup transformers become inoperable, within one hour take action to place the reactor in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 205 hours0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br />.
~
Discussion:
The existing specifications do not address loss of both startup transformers. This change defines the actions required if both startup transformers are inoperable.
This change is consistent with the recommendations of Generic Letter 84-15. The 205 hour0.00237 days <br />0.0569 hours <br />3.38955e-4 weeks <br />7.80025e-5 months <br /> time is consistent with the Districts November 7, 1985 revised request for exemption to some requirements of 10 CFR 50, App. R.
- 9. Existing Specification:
3.7.2.G. Both sets of nuclear services buses 4A, 4A2, and 4B, 4B2 are operable except that one set of nuclear service buses (4A, 4A2 or 4B, 4B2) may be removed from service for_not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided that all equipment on the other set of nuclear service buses is operable.
ligw Specification:
3.7.2.H. Nuclear service buses as listed in 3.7.1.D and 3.7.1.E shall be operable except should one nuclear service bus become inoperable for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 24
.A
cold shutdoun within the following 30-hours.
If more than one nuclear service bus as listed in 3.7.1.D and 3.7.1.E should become inoperable, within one hour take action to g place the reactor in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the ,
following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Discussion:
This change more clearly defines the actions required if one nuclear service bus becomes inoperable. It also defines the actions required if more than one bus becomes inoperable. This change is consistent with the format and intent of the other changes to the specifications.
- 10. Existing Specification:
3.7.2.F. Nuclear service batteries are charged and in service except that one nuclear service battery may be removed from service for not
~
more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Egw Specification:
3.7.2.I. Nuclear service batteries as listed in 3.7.1.G shall be charged and in service except should one nuclear service battery become inoperable for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If more than one nuclear service battery as listed in 3.7.1.G should become inoperable, within one hour take action to place the reactor in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Discussion:
This change defines actions required if one or more nuclear service batteries are out of service. This change is consistent with the recommendations of Generic Letter 84-15.
- 11. Existing Specification:
N/A Egw Specification:
3.7.2.J. One nuclear service battery chargers as listed 25
in 3.7.1.H shall be operable except should one nuclear service battery lose its required battery charger for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If more than one nuclear ,
service battery loses its required battery -
charger as listed in 3.7.1.H, within one hour take action to place the reactor in hot
- shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Discussion:
This revision adds operability requirements for nuclear service battery changers. This revision incorporates the recommendations of Generic Letter 84-15.
- 12. Existing Specification:
3.7.2.H. If the switchyard voltage goes below 219KV, positive actions, within the District's procedures, will be implemented in an attempt to return the voltage to 219KV. If the switchyard: voltage goes below 217KV or remains below 219KV for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, one electrical division will be operated on its diesel generator independent of off-site power. The other electrical division will be operated on off-site power with its associated diesel generator on standby status. The switchyard voltage must be returned to 219KV within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Switchyard voltage above 219KV will allow unrestricted plant operation.
New Specification:
.i 3.7.2.L. Should the switchyard voltage drop below 219KV, positive actions, within the District's procedures, Will be implemented in an attempt to return the voltage to 219 KV. If the switchyard voltage goes below 217 KV or remains below 219 KV for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,'both diesel generator trains shall be demonstrated operable within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by performing surveillance requirement 4.6.3.A.4. Should the switchyard voltage not be restored above 219 KV within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
26
L
. . a Discussion:
The change eliminates the requirement to continuously run a diesel generator. The revised surveillance requirements are consistent with the recommendations of Generic Letter 84-15 and IE Information Notice 84-69, ,
Supplement I. ^ -
- 13. Existing Specification 1 N/A New Specification:
3.27 Nuclear Service Electrical _Euilding Emergency H2ating Ventilation and Air Conditioning System Aeolicability -
This specification 9pplies to the operability-of the Nuclear Ser' rice Electrical Building Emergency Hdating Ventilation ~and Air Conditioning System. 1 Objective To assure that this system will be able to perform its designed function.
Specification 3.27.1 Both Nuclear Service Electrical Building Emergency Heating Ventilation and Air Conditioning trains shall be operable at all '
times except as noted in 3.27.2.
3.27.2 With one Nuclear Service Electrical Building Emergency Heating Ventilation and Air Conditioning train inoperable, restore the train to operable status withi.' 7 days or be in at least hot standby within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in cold shutdown within following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. With both Nuclear Service Electrical Building Emergency Heating Ventilation and Air Conditioning trains inoperable, restore the trains to operable status within 3.5 days or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -
27
Bases The Nuclear Service Electrical Building (NSEB) Emergency Heating Ventilation and Air Conditioning system is required to provide cooling to protect required electrical components in the NSEB. . j Discussion: ;
This new section is added to provide a LCO for the surveillances required by Specification 4.31. It defines the operability requirements for the system to be consistent with Amendment 68 dated June 3, 1985 where the NRC stated that the system was required to be operable :
even though a LCO specification was not provided. Since the system is now backed by the diesel generators during loss of offsite power, the time to restore the system to ,
operable status if it is found to be inoperable has been increased from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 7 days for one train and 3.5 days for both trains. This approach is consistent with the requirements of specification 3.13.3 on the ;
Control Room /TSC Emergency Filtering System.
I
- 14. Existina Specification:
N/A . .
i NAM Specification:
4.8.1 Offsite Power Sources A. Each of the 220 KV lines required by 3.7.1.A shall be: Determined OPERABLE at least once per 7 days by verifying correct breaker alignments and indicated r power availability, and B. Each Start-up Transformer required in 3.7.1.C shall be demonstrated OPERABLE at least once per refueling shutdown by 3
transferring nuclear service buses as listed in 3.7.1.D from the normal startup ;
! transformer supply circuit to-the l alternate startup transformer supply
- l circuit.
Bases l The operability of the 220V lines, the nuclear service 4160V buses, and the nuclear service 480V buses are
}
i demonstrated by verifying correct breaker alignments and i indicated power availability. Surveillance 4.6.1.B can !
- j. only be performed during a refueling shutdown when both !
4 diesel generator trains are operable or the core is I
l l
28 i
. -. :LL - _:L T. ._ - _,L :.L__.,,;.- . r L % a . ' ~~ ~:L. , " __* . J_ '_.Li ,~ ^ '~ ~ . J.,. _
flooded to 37 feet to ensure the required decay heat removal capability is available.
Discussion:
This new specification adds surveillance requirements to ,
demonstrate the operability of the 220 KV lines and startup transformers required by Specifications 3.7.1.A and 3.7.1.C. This is a new surveillance requirement since the existing specifications do not require a i surveillance to demonstrate operability of the 220 KV lines or startup transformers.
- 15. Existing Specification:
N/A Haw Specification:
i 4.6.2 Nuclear service buses required by 3.7.1.D and i . 3.7.1.E shall be determined to be OPERABLE at least once per 7' days by verifying correct breaker alignment and indicated power availability.
i
) Discussion:
This adds surveillance requirements to demonstrate the ,
j operability of.the nuclear service 4160 V and 480V buses l
) required by Specifications 3.7.1.D and 3.7.1.E. This is l l consistent with the overall format of the specifications. ,
i I
- 16. Existing Specification:
(
4.6.1 At intervals not to exceed one month, a test of the diesel generators will be performed to
- verify proper operation of these emergency l power sources and associated equipment. This I test will be performed to assure that
A. Each diesel generator can be started from the control room.
B. Each diesel generator can be synchronized with its associated 4160 volt nuclear service bus.
! 4.6.5 Diesel generator fuel oil supply shall be
, tested as follows:-
l A. During the monthly diesel generator test, the diesel fuel oil transfer pumps shall l
29
_ = =: = = = = = = = = = _ = - , - - _ . .__:-==_._:.====. .
be monitored for operation.
B. Once a month, quantity of the diesel fuel
! oil shall be logged and checked against l minimum specifications.
The tests specified will be considered satisfactory if control room indication
- and/or visual examination demonstrates
' that all components have operated properly.
M Specification:
i 4.6.3 Each diesel ger.erator shall be demonstrated i OPERABLE by verifying that both diesel generators in the train are OPERABLE (A and A2, B and B2). Each diesel generator shall be l
i demonstrated OPERABLE.
i A. In accordance with the frequency l specified in Table 4.6-1 by:
e
- 1. Verifying the fuel level in the day tank,
- 2. Verifying the fuel level in the fuel storage tank, l 3.* Verifying the fuel transfer pump starts and transfers fuel from the l storage system to the day tank, 4.* Verifying the diesel starts from a manual signal and accelerates to a nominal 900 rpm for A and B and 450 rpm for A2 and B2. The generator voltage and frequency shall be 4160
. (1420) volts and 60 ( 1.2) Hz after the start signal. (Note 1) 5.* Verifying the generator is synchronized and operates with its full load of 2650 (1100) kw for A_
and B and 1650 (1100) kw for A2 and-B2 for at least 60 minutes, (Note 1).
6.* Verifying the diesel generator is aligned to provide emergency power to the nuclear services buses at the conclusion of the test.
- Surveillance shall not be performed when a diesel generator train is not operable in accordance with TS 30
3.7.2.
Note 1: All engine starts for the purpose of this surveillance testing may be preceded by an engine prelube period. Further, alls surveillance tests, with the exception of once ,
per 184 days and 10-year duration test, may be preceded by warmup procedures recommended by the manufacturer and may also include slow starting and gradual loading so that mechanical stress and wear on the diesel engine is minimized. The testing performed once every 184 days shall include fast starting (less than or equal to 10 seconds) and fast loading.
Table 4.6-1 DIESEL GENERATOR IEET SCHEDULE
. Number of Failures in Last-20 Valid Tests
- Test Frequency 1 At least once per 31 days 2 At least once per 7 days **
- Criteria for determining number of failures and number of valid test shall be in accordance with Regulatory
- Position C.2.e of Regulatory Guide 1.108, Revision 1, August 1977, where the number of tests and failures is i
determined on a per diesel. generator basis. For the purposes of this test schedule, only valid tests conducted after the license amendment issuance date shall be included in the computation of the "last 20 valid tests."
- This test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one or less.
Bases The limiting of the maximum load on the TDI diesel generators A2 and B2 to less than the qualified load of l
! 3300 kw provides assurance that the crankshafts will stay within the proven limits for high-cycle fatigue j cracks.
i a Diesel generators A2 and B2 uill be loaded during
! surveillance testing to 1650 KW which provides assurance that the qualified load of 3300 KW will not 4
be exceeded. The 1650 KW loading is one half the 31
qualified load and over twice maximum load of safety equipme'nt during the loss-of-offsite power.
Discussion:
This change incorporates recommendations from Generic , ,
Letter 84-15 and IE Information Notice 84-69, Supplement
- 1. These include the reduction of fast starts and test frequency based on failures. The change more clearly identifies the requirements for demonstrating the operability of the diesel generators.
- 17. Existing Specification:
N/A Mg.w Specification:
4.6.3.B By sampling at least once per 92 days the oil in each fuel oil storage tank and by sampling new fuel oil prior to addition to the storage tanks in accordance with ASTM-0270-1975 and verifying:
- 1. A water and sediment content of less than or equal to .05 volume percent when tested in accordance with ASTM-D975-77,
- 2. A kinematic viscosity at 40 C of greater than or equal to 1.9 but less than or equal to 4.1 when tested in accordance with ASTM-D975-77.
Discussion:
This requirement is presently being performed by plant surveillance procedures even though the items were not in the specifications. This is consistent with the requirements of Regulatory Guide 1.137, Fuel-Oil Systems For Standby Diesel Generators.
- 18. Existing goecification:
4.6.2 During each refueling interval, a test of the diesel generatcrs and emergency start circuits shall be performed to verify that these emergency power sources and associated equipment are operable by:
A. Simulating a loss of offsite power in conjunction with a safety features actuation signal, and:
32
- 1) Verifying de-energization of the nuclear services buses and operation of the load shedding circuitry.
- 2) Verifying the diesel starts from ambient condition on the auto-start signal and energizes the nuclear ,
- services buses, and by verifying proper operation of the automatic load sequencing circuitry, including manual closing of the A Train intertie breakers from the Control room and manual energization of the A Train essential HVAC from the l
Control Room. The B Train intertie breakers are automatically closed i and the B Train Control Room essential HVAC System is j automatically energized. The diesel
- generators will be operated for at i least 5 minutes in this condition.
i B. Simulating a loss of offsite power and verifying that on interruption of the emergency power sources the loads are shed from the nuclear services buses in accordance with design requirements and that subsequent loading of the emergency power" sources is through the automatic load sequencing circuitry. The diesel generator will be operated for at least 5 l minutes in this-condition.
I C. Load testing the diesel generators for SFAS capacity.
4.6.3 Each diesel generator shall be given a thorough inspection at least biannually following the manufacturer's recommendations
!. for this class of standby service.
Bases:
They assure the 3B, 3B2 intertie breakers are automatically closed and the B Train Control Room essential HVAC System is automatically energized. The 3A-3A2 and 3B-3B2 interties are not required if the event is only a safety features actuation.
Egw Specification:
4.6.3.C. At each refueling shutdown, by:
- 1. Subjecting the diesel generator to an 1
inspection in accordance with procedures prepared in conjunction with the 33
~
_- .-.. - . . -..==========: - - - - . . = : =_ . : ::: . ._ w _
- l I
I maintsnance recommendations provided by i its manufacturer for this class of standby service.
- 2. Simulating a loss of off-site power in l' conjunction with a safety features ,
actuation signal, and a) Verifying de-energization of the I nuclear service buses and operation of the load shedding circuitry.
- b) Verifying the diesel starts on the 1 i auto-start signal (Note 1), !
j energizes the nuclear service buses, -
- verifying proper operation of the t automatic load sequencing circuitry, ;
and operates for greater than or equal'to 5 minutes in this condition. After energization, the ;
steady state voltage and frequency of the emergency buses shall be l maintained at 4160 (1420) volts and I 60 (11.2) Hz during this test.
l
- 3. Simulating a loss of off-site power and verifying that on interruption of the ;
i emergency power sources the loads are sh~ed.from the nuclear services buses in i accordance with design requirements and that subsequent loading of the emergency power sources is through the automatic load sequencing circuitry. The diesel generator will be operated for at least 5 minutes in this condition.
- 4. Verifying the diesel generator operates i for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of this test, the diesel generator shall be loaded to 2750 i ( 100) kw for A and B and 1650 (1100) kw
! for A2 and B2 and for the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of this test, the diesel generator i shall be loaded to 2650 ( 100) kw for A
,i and B and 1650 (1100) kw for A2 and E2. ,
Within 5 minutes after completing this 24-hour test, perform Surveillance ;
Requirement 4.6.2.C.2. l Note 1 All engine starts for the purpose of this surveillance testing may be preceded by an engine prelube period. Further, all surveillance tests, with the exception of once per 184 days and 10-year duration tests, may be preceded by warmup procedures recommended by the manufacturer and may also include slow 34 i
a:v. ,---mm., m.a w n ~-" w :: '- v~ x - -=.~-~m- -- - - -
c "- -~~--
starting and gradual loading so that mechanical stress and wear on the diesel engine is minimized. The testing performed once every 184 days shall include fast starting (less than or equal to 10 seconds) and fast loading. ,
l Discussion:
i This change expands and clearly identifies the
! requirements for load testing the diesel generators.
This change incorporates the recommendations in Generic Letter 84-15. The Cycle 7 testing requirements for the Control Room HVAC are also de-leted. With the addition of the TDI diesel generators the Control Room HVAC will be automatically loaded in accordance with the !
District's commitment to meet NUREG 0737, item III.D.3.4, Control Room Habitability.
- 19. Existing Es_ecification:
N/A ,
Hgw Specification:
- 4.6.3.D. At least once per 10 ' rears, by starting all 1 four diesel generators simultaneously and i verifying that they accelerate to a nominal
) 900 rpm for A and B and 450 rpm for A2 and B2 i within 10 seconds after the start signal. The i generator voltage and frequency shall be 4160 (1420) volts and 60 (11.2) Hz within 10.0 seconds after the start signal.
} 4.6.3.E. In lieu of performing 4.6.3.C.2 at least once per 10 years simulating a loss of off-site I power in conjunction with a simulated SFAS,
! and i 1. Verifying de-energization of the nuclear service buses and load shedding from the nuclear service buses.
- 2. Verifying the diesel starts on the auto-start signal, energises the nuclear service buses with permanently connected loads, loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency buses i shall be maintained at 4160 (1420) volts and 60 (11.2) Hz during this test.
35
. - _ - . n. - = - _ = - - _ = - =_=_ = _ _______-
- 3. Verifying for the A and B diesel generators that all automatic diesel generator trips, except engine overspeed, ground fault and generator differential, and verifying for the diesel generators A2 and B2 that all automatic diesel ,
generator trips, except engine overspeed, low lube oil pressure and generator differential, are automatically bypassed with an SFAS.
Bases:
At least once per 10 years a diesel generator test will ,
be performed simulating a loss of offsite power in conjunction with a simulated SFAS and loading of actual loads to the maximum extent possible without damaging plant systems (i.e. use of recirculation flow or manual valving out of a system to protect plant components).
Discussion:
These specifications are added to demonstrate that all four diesel generator systems will start simultaneously and that they will accept actual emergency loads. This incorporates the recommendations of Generic Letter 84-
- 15. ,
Safety Evaluation 12r Items in Section III:
These are changes to the operability and surveillance requirements. In most cases the new requirements are more restrictive. In some cases the requirements are less restrictive. These are cases where the existing surveillance may actually degrade the operability of the system. The requirements for cold. fast starts of diesel generators are an example of this. The changes incorporate recommendations from Generic Letter 84-15
! " Proposed Staff Actions to Improve and Maintain Diesel i Generator Reliability", IE Information Notice 84-69, Supplement 1, " Operation of Emergency Diesel Generators,"
and Regulatory Guide 1.137, " Fuel-Oil Systems for Standby Diesel Generators."
l Basis f2r En Significant Hazards Determination:
The proposed changes do not involve a significant j hazards consideration because operation of Rancho Seco in accordance with this change would not:
, 1) involve a significant increase in the probability or consequences of an accident previously evaluated. The changes are to the. operability and surveillance requirements. They add requirements for new systems, improve requirements for existing 36 w .. . -. . .
_ _ _ . . . . . = . _ .
e systems, and incorporate NRC recommendations. The changes demonstrate the operability of required systems to ensure safe operation of the plant.
Therefore, these changes do not increase the probability or consequences of an accident.
- 2) create the possibility of a new or different kind of accident from any previously analyzed. The changes demonstrate the operability and surveillance of critical plant systens. The changes to the operability and surveillance requirements do not create the possibility of a new or different kind of accident.
- 3) involve a significant reduction in a margin of safety. The changes improve operability and surveillance requirements and therefore preserve the margin of safety. They will not reduce the margin of safety.
9 e M e
37
. . - . _ ._ ~_ ..
_ .-= .- _ ,