ML20136G287
| ML20136G287 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/15/1985 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Thompson H Office of Nuclear Reactor Regulation |
| References | |
| 2NRC-5-146, GL-85-12, TAC-62926, NUDOCS 8511220301 | |
| Download: ML20136G287 (5) | |
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'Af DuquesneUsht g=si6 af (412)923-1960 Nuclear Construction Division coef
- 2) 187-2629 Robmson Plaza, Buildmg 2 Sulle 210 Pittsburgh, PA 15205 November 15, 1985 United States Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
Mr. Hugh L. Thompson, Jr., Director Division of Licensing Of fice of Nuclear Reactor Regulation
SUBJECT:
Beaver Valley Power Station - Unit No. 2 Do cke t No. 5 0-412 Generic Letter 85-12 "Impleme ntat io n of TMI Action Item K.35,
' Automat ic Trip of Reactor Coolant Pumps,' Schedule"
REFERENCE:
2NRC-5-124, dated August 27, 1985 Gentlemen:
In accordance with the commitment made in Duquesne Light Company 's (DLC) submittal of August, 1985 (2N RC 124 ), the following info rmat ion fo r Beaver Valley Power Station Unit 2 (BVPS-2) is submi t ted for Generic Letter 85-12.
Items A.1, A.2, B.2, C.l, and C.2 of the Safety Evaluation Re po rt (SER),Section IV 1mplementation Section of Generic Letter 85-12.
Item A.1 of the implementation Section (IV) of the SER requests identification of the ins t rume ntat io n to be used to determine the RCP trip set po int, including the degree of redundancy of each parameter signal needed for the criterion chosen. In response to this item, the following information is submitted.
The reactor Coolant Pumps (RCP) are tripped if BOTH conditions listed below occur:
A) At least one Charging / High Head Safety Injection (HHSI) Pump running B) Reactor Coolant System (RCS)/ Steamline differential pressure is less than 145 paid The instrumentation used to determine when RCP trip is initiated is:
A) Charging /IIHSL pump breaker position indication (2CHS*P21A, P21B, and P21C) Control Room Bench Board-A (BB-A)
B) RCP breaker position (2RCS*P21A, P21B, and P21C) BB-A C) RCS pressure indicator (2RCS*P1402 and 403) Control Room Vertical Board-A (VB-A)
D) Main Steam pressure indic ato rs (2 MSS *P1474, P1475, P1476, P1484, P1485, P1486, P1494, P1495, and P1496) VB-C The instrumentation listed above is the minimum designated by Eme r-gency Operating Procedure backup documents to determine whether RCP 0011220301 851115 DH ADOCK 0500 2
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r United Sttteo Nuclerr Regulatory Commission Mr. Hugh L. Thorpcon, Director Generic Letter 85-12 Page 2 trip criteria are met.
In addition, the operators are trained to use all available, redundant instrumentation to dete rmine the existence of an RCP trip condition.
In regards to item A.2 of Section IV o f the SER, DLC submit s the following information for BVPS-2.
The wide range RCS pressure instrument channel uncertainty, under normal conditions, corresponds to 76 psi. This uncertainty considers all error components from the sensor through to the display on the PSMS.
The wide range pressure transmitters are located outside containment and as such are not subjected to adve rse containment conditions.
The re fo re, the adverse and normal containment instrument unce rt ain-ties are identical.
The instrument undertainty fo r the steamline pressure channels corresponds to 33 psi. This uncertainty includes all error components from the transmitters through to the indicator on the control board.
This uncertainty bounds that associated with the steamline pressure channels displayed on the PSMS.
The steamline pressure transmitters are also located out s ide of containment and are also not subjected to adverse containment cond i-tions.
The re fo re, the adve rse and normal containment instrument uncertainties are identical.
Combining the RCS wide range pressure and steamline pressure uncer-tainties using the square root of the sum of the squares method results in an overall uncertainty of 83 psi fo r both normal and adverse containment conditions.
The ef fects of fluid jets and pipe whip on instrument reliability has also been considered on the reliability of the instruments.
Stone &
Webster Engineering has verified that all the RCS wide-range pressure and Steamline pressure indications will be protected for all LOCA and Secondary side line breaks; t he re fo re, there is no impact on the instrument reliability.
Item B-2 of the SER requests identification of the components required to trip the RCPs. The following is DLC's response to this item:
The components required to trip the RCPs manually at BVPS-2 are:
1.
The control switch on the benchboard in the Main Contre i Room for each RCP ( 1-RCSAA, 1-RCS BA, and 1-RCSCA fo r RCPs e,
B, and C,
respectively).
2.
Proper operation of the 4160 volt circuit breaker fo r each RCP (breakers 2A6, 2B6, and 2C6).
These breakers are located in the normal switchgear area at elevation 760'-6" of the Service Building.
3.
The availability of control power (125 vde) to the RCP 4160 volt circuit breaker control circuits.
a
s United Stctea Nuclect R:gulatory Commission Mr. Hugh L. Thompcon, Director Generic Letter 85-12 Page 3 Control power is supplied from a de switchboard to the 4160 volt switchgear fo r each respective RCP.
RCPs A and B are fed from separate breakers of de switchboard 2-5, while RCP C is fed from a breaker on de switchboard 2-6.
The switchboard breakers feed two series connected molded case circuit breakers in each RCP 4kV breaker cubicle which energize the 4kv breaker control circuit.
The control breakers in the RCP 4ky breaker cubicle are Heinemann CD-3, 200 amp frame, magnetic breakers.
The de switchboard b reake rs supplying the RCP trip circuit are doub le mechanically-interlocked breakers.
One breaker in each pair is normally open and the other breaker is normally closed.
Both breakers are Gould-ITE molded case thermal-magnetic, type JL with a 400 amp frame.
The normally closed breaker is normally supplied powe r from a batter charger which is fed from a motor control center.
The motor control center is by the backed-up(one charge r fo r s t and-by diesel generator.
When the battery each switchboard) is not available, the station bat te ry (i.e.
BAT-2-5 or BA T-2-6 ) associated with the switchboard provides backup power to the nonnally closed breake r.
The normally open breaker is fed from the other de switchboard (e.g., normally open breaker on DC-SWBD2-5 is fed from DC-SWBD2-6).
The location of these 125vde control equipment (and associated equipment) is as follows:
Service Building (elev. 760'-6"):
125vde control battery 2-5 de switchgear 2-5 battery charger 2-5 de switchboard 2-5 battery breaker cubicle 2-5 battery charger 2-6 de switchboard 2-6 Turbine Building (elev. 752'-6"):
125vde control battery 2-6 de switchgear 2-6 battery breaker cubicle 2-6 Aux. Building (elev. 77 3'-6"):
motor control center 2-23 motor control center 2-26 The -ollowing info rma t ion is submit ted in response to items C.1 and C.2, respectively.
C.1:
RCP operation is of a concern during small break LOCA events. A c c i-dent analysis determinations are based on the assumption that offsite power is lost at the caset of the event.
This is not necessarily a conservative assumption during a small break LOCA. The peak cladding temperature limit of 2200*F may be exceeded if RCPs trip during a critical time period. The RCP trip criteria is based on tripping the RCP's prior to reaching this critical time.
If RCPs are not running during the event, RCS inventory will decrease to a level below the break elevation.
When this occurs, only steam is released out of the break.
With RCPs running, a s team / wate r
United Statec Nuclear Reguletory Core iss e-Mr. Hugh L. Thompton,' Director Generic Letter 85-12 Page 4 I
mixture is forced out of the break.
This results in a larger anount of mass loss from the RCS.
The fo rced circulation, however, also results in an increase in core cooling.
The critical time period is I
when the increased inventory loss results in a deeper core uncovery when the pumps trip, and the core has not been cooled enough to prevent exceeding the peak cladding temperature limit.
The RCP trip criteria ensures that the operator will trip the pumps
. prior to reaching the critical time period. The increase in RCS mass loss due to RCPs running does not occur until the break would have uncovered had pumps not been running. The break cannot uncover until the primary side of the SG U-tubes have drained.
The U-tubes cannot drain until they reach saturation.
The temperature of the fluid in the U-tubes will have reached saturation.
The RCPs must be tripped prior to the RCS reaching SG pressure.
The criteria fo r tripping the RCPs is RCS pressure 145 psi greater than the highest SG pressure.
This number is derived from summing i
primary to seconda ry del t a-P corresponding to the delt a-T required for heat transfer, he igh t dif ference of the U-tubes to the pressure instruments, steamline de l t a-P, and instrument inaccuracies.
If primary to secondary delta-P decreases to 145 psi and high head S1 flow is verified, the RCPs must be tripped.
During non-LOCA events it is desired to keep RCPs running to provide nonnal RCS pressure control and vessel head cooling.
If pumps are t ripped during a SGTR or other non-LOCA events, pressurizer PORVs would have to be utilized to reduce RCS pressure.
Due to past oper-ating experiences at TML and Ginna of PORVs sticking open, this is not the desired way to depressurize. The 145 psi primary to secondary delt a-P trip criteria provides protection during. small break LOCA events, while ensuring pumps are not tripped during a design basis SGTR event of a single tube and non-LOCA events. The non-LOCA events analyzed are a steam break and a feedline break.
The analysis shows that primary to secondary delta-P should not decrease below 350 psi for these events.
RCPs will not be tripped during a SGTR or non-LOCA event unless containment pressure rises to 10psig and CIB actuates.
The resulting loss of CCR to the RCPs would then require that they be tripped.
C.2:
The following E0Ps have RCP trip related operations:
A.
Optimal Recovery Procedures
- Reactor Trip or Safety injection
- Steam Generator Tube Rupture
- Loss of Reactor or Secondary Coolant
- Uncontrolled Depressurization of all Steam Generators
- Post LOCA Cooldown and Depressurization
- SI Termination
- Post-SGTR Cooldown Using Steam Dump B.
Critical Safety Function Status Tree
- Core Cooling I
n..
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Unitsd States Nuclear Ragulatory Commission Mr. Hugh L. Thorpzon, Director Ge neric. Le tte r ' 85-12 Page 5 C.
Function Restoration Procedures
- Response to inadequate Core Cooling
- Response to Loss of Secondary Heat Sink Response to Imminent Pressurized Thermal Shock Condition W stinghouse Owners The BVPS-2 E0Ps comply with Revision 1 of the e
Group guidelines for RCP trip criteria.
The above 'information completes BVPS-2 's. response to Generic Letter 85-12 addressing Section IV Implementation Section of the SER enclosed.
If there are any questions concerning this letter, please contact Mr. S. D. Hall of my staff at (412) 787-5141.
DUQUESNE L1 JI COMPANY h
By Jf arey Vi res ident SDH/wjs cc:
Mr. B. K. Singh, Project Manager Mr. G. Walton, NRC Resident Inspector INPO Records Center
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