ML052940341

From kanterella
Jump to navigation Jump to search
NRC Inspection Report No. 05000390-05-013; Preliminary Greater than Green Finding; Watts Bar - Submittal of Regulatory Conference Meeting Package
ML052940341
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 10/18/2005
From: Pace P
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
IR-05-013
Download: ML052940341 (38)
v • d • e


Text

{{#Wiki_filter:Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 OCT 1I 2005 10 CFR 50.4 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen: In the Matter of ) Docket No. 50-390 Tennessee Valley authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - NRC INSPECTION REPORT NO. 05000390/2005013; PRELIMINARY GREATER THAN GREEN FINDING; WATTS BAR NUCLEAR POWER PLANT - SUBMITTAL OF REGULATORY CONFERENCE MEETING PACKAGE In accordance with NRC Letter dated September 7, 2005, TVA is providing TVA's meeting package one week prior to the NRC Region II Regulatory Conference scheduled for October 25, 2005. This package contains supplemental information related to the subject finding and will be discussed by TVA personnel during the meeting. The meeting package is provided in the enclosure. There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please call me at (423) 365-1824. Sincerely, < P. L. Pace Manager, Site Licensing and Industry Affairs Enclosure cc: See Page 2 e

vl U.S. Nuclear Regulatory Commission Page 2 OCT 18 2005 Enclosure cc (Enclosure) NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303

I Enclosure

l~~S St:::}::' ~A

IE 7,r17 A

Ao-.Sstm I'vI w. Z " "'" ' '"' 'i.i--. r; .';...4..'.:.. -:. !.v: .-i t' . S

,,4

,,I_r i-

Agenda _aM 0010

  • Introduction
  • Objective Of Presentation
  • Cold Overpressure Mitigation System Actuations
  • Differences Between The TVA And NRC Risk Analyses M. Skaggs M. Skaggs D. White F. Koontz
  • Regulatory Summary
  • Closing Remarks P. Pace M. Skaggs 2

Introduction and Objective of Presentation Mike Skaggs WBN Site Vice President 3

Introduction NRC Finding "... GO-6, Unit Shutdown from Hot Standby to Cold Shutdown, Section 5.5, Step [1] [e] states, "Slowly RAISE charging to fill Pressurizer at less than 30 gpm." SOI-74.01, Residual Heat Removal, Section 8.11, implemented a flush of the A train RHR heat exchanger bypass during shutdown cooling and contained a note which stated, "The effect on RCS heatup/cool down should be evaluated." Each procedure was not adequately implemented approaching and during solid plant operations on February 22, 2005." TVA agrees with the performance deficiency 4

Objective of Presentation Five TVA Focus Areas TVA has identified five key differences between the TVA and NRC analyses There were only 5 Power Operated Relief Valve (PORV) lifts that relieved pressure rather than the 7 shown in the NRC event tree More rigorous mathematical treatment of multiple initiating events (each successive PORV lift) is warranted RHR suction relief valve is more reliable than in NRC's evaluation Two additional RHR Discharge Relief Valves were available to relieve increasing RCS pressure Secondary plant cooling was available to prevent core damage

  • In addition - TVA will show our evaluation inputs meet the Manual Chapter 0609 guidance to be "reasonable and realistic" 5

Cold Overpressure Mitigation System (COMS) Actuations Dana White WBN Operations Manager 6

Simplified Reactor Coolant System Lineup To PRT 'CONS RHR Suction Relief 7

Event Description Initial Conditions

  • RCS temperature/pressure = 139°F at 288 psig
  • Pressurizer level - approaching solid plant operations
  • Charging/letdown = 177/154 gpm - Filling the pressurizer at 23 gpm
  • Charging Flow Control Valve out of service for modification
  • Cooling RCS using both trains of RHR
  • Two Reactor Coolant Pumps running
  • Secondary Plant Cooling available All four Steam Generators - 75% wide range level Four Steam Generator PORVs available Condensate Storage Tank Level at approximately 290,000 gallons Motor Driven Auxiliary Feedwater Pumps available 8

Event Description

  • COMS designed to protect the reactor vessel from brittle fracture during overpressure transients by limiting Reactor Coolant System (RCS) pressure during low temperature operations
  • Tech Spec requirements met:

1B-B Charging Pump was in service Other injection sources were isolated One PORV and the RHR Suction Relief Valve were the operable COMS relief valves 9

Event Description

  • After the Charging Flow Control Valve was made available, the operating crew decided to return to the normal charging alignment for better control during solid water operations
  • Charging Flow Control Valve functioned erratically, cycling charging flow and RCS pressure - returned to bypass valve
  • The PORV lifted 5 times to relieve pressure over a 1 minute 40 second period while crew removed the Charging Flow Control Valve from service.

(RHR suction relief valve setpoint of 450 psig was not reached) 10

Event Description PORV Actuations 2/22/05 14:09 thru 14:13 500 450 400 350 u" 300 $ 250 M s 200 150 100 50 0 1,,. ,UC.,.. \\z1 JfCE CP-7wir~~r~1repn --- -r I~ ? U O~L~~pili -~- -~- -i- - t. A . I  I 0,- Open 00 Gb Closed If If-s 7_va C___ _: fz t4____ .4 - XA*'*0C :t

_ tS S

.\\ 0

  • -- r 4+ ~

~d+ "I fr ~

~~~~~

~~ ;7t? P;0:S }XS AX:,20Tae !t:0\\f0XkaS~;A, V::.:0 0

0y fi;00:3T;-N IS i.

t:H t A :E \\ S i.  7777 W--- I-;-- tttt I ::-- 7 I------

- 7 77 

7---  t4 4 --- 11 tI II -n, ;I-t _ f _~,. _f t

_ _ _t I \\

Ab_ A 9t V S\\ E. _.RH 14; - i  -1. 4 -. i -  i44 -

  1. q##
H
::PpPA#;:;#5;:__I

-e ;+e

e'Vi:!

hi:Hi#Gid fee ee:eN+e:.:

Fi#.E#:t,+:#H:e';e i:ie+:.e:E: W#-e I,

II

  • t S t e

e t: e 0e j e H ;e

?

t tej e ^ 4 _:f:n > y t. e 2 tS :: t.Y. i: f i e F H-: h; e 7-e  I - I A) z tS

e S :

b  No 14:09:27 14A:1L06 TIme 11 r (-Z

Causes and Corrective Actions

  • Causes:

A lack of sensitivity to and failure to recognize the complexities associated with approach to solid water operations Contributing to this event was the unsuccessful attempt to fix the performance of the Charging Flow Control Valve

  • Corrective Actions:

Have revised procedures cautions and controls while approaching solid operations Pre-outage training on this event, COMS and planned system work Pursuing hardware improvements to Charging Flow Control Valve 12

Discussion of Differences Between the TVA and NRC Risk Analyses Frank Koontz WBN Engineering Specialist 13

Five Key Differences Between the TVA and NRC Analyses

1.

There were only 5 Power Operated Relief Valve (PORV) lifts that relieved pressure rather than the 7 shown in the NRC event tree

2.

More rigorous mathematical treatment of multiple initiating events (each successive PORV lift) is warranted

3.

RHR Suction Relief Valve failure to open is more reliable than in NRC's evaluation

4.

Two additional RHR Discharge Relief Valves were available to relieve increasing RCS pressure

5.

Secondary plant cooling was available to prevent core damage 15

Difference No. 1 5 PORV Lifts Rather Than 7 in NRC Event Tree

  • Only one in-service PORV lifted to relieve pressure
  • Have shown PORV relieved total of 5 times
  • Second PORV available but isolated and not credited for tech spec compliance

Conclusion:

Second PORV does not adversely impact risk analysis and 5 lifts should be limit of analysis 16

4 Difference No. 2 - More Rigorous Mathematical Treatment of Each Successive PORV Lift

  • Three PORV risk states possible Opens and closes successfully Fails to open Opens but fails to re-close
  • Not a straight 5 times multiplier for each lift
  • For example - If PORV opens but fails to close lst time, other four lifts never would have happened Second example - the 5th lift would only have happened if the first four lifts cycled successfully 17

Difference No. 2 - More Rigorous Mathematical Treatment of Each Successive PORV Lift

  • Added mathematical rigor developed by ABS consulting
  • [(l-p)*(l-q)]A n = opens/closes successfully p * {1[(1-p)*(1-q)]An} / [1-(l-p)*(l-q)] = fails to open
  • (1-p)*q * {l-[(.-p)*(1-q)]An} / [l.(bp)*(l-q)] = opens but fails to reclose
  • Where n is the number of open (and l

q reseat) challenges p is the probability PORV failure lq to open per challenge 1p q is the probability PORV failure l q to reseat per demand

Conclusion:

Straight multiplier for ll =3...etc successive lifts is overly conservative 18

Difference No. 3 Available RHR Suction Relief Valve AL

  • RHR suction valve a 3-inch Crosby model JB-35-TD-WR
  • Relief capacity of 900 gpm
  • Setpoint - 450 psig
  • WBN has not experienced a failure of this type of relief valve or similar Crosby model to relieve
  • Reviewed EPIX data - no failures to relieve were identified
  • Test results - Valves tested soon after the COMS event - Relieved within or below acceptable setpoint range 19 CON3

Difference No. 3 Available RHR Suction Relief Valve

  • NRC RHR relief valve failure probability 1E-3 No specific data in NRC SPAR model for small relief valves NRC selected "similar valve" - pressurizer code safety Pressurizer code safety designed to lift at 2500 psig and 600TF.

Pressurizer code safety challenged by adverse conditions

  • TVA RHIR relief valve failure probability 2.42 E-5 TVA used PLG-0500, "Database for Probabilistic Risk Assessment for Light Water Nuclear Power Plants" in IPE NRC staff evaluation of IPE documented in SER dated October 5, 1994 SER specifically recognizes PLG-0500, "Database for Probabilistic Risk Assessment for Light Water Nuclear Power Plants" PLG database reviewed by NRC - NUREG/CR5606 - concluded database was extensive and "state of the art"

Conclusion:

WBN value is Current Licensing Basis for this component 20

Difference No. 4 Two Available RHR Discharge Relief Valves WA

  • Two - 2 inch inlet Crosby Valves -

Model No. JB-35-TD-WR

  • Setpoint - 600 psig
  • WBN has not experienced a failure of this type of relief valve or similar Crosby model to relieve
  • Reviewed EPIX data - no failures to relieve were identified
  • Test results - Valves tested soon after

_the COMS event - met acceptance criteria 21 coCF

Difference No. 4 Two Available RHR Discharge Relief Valves

  • FSAR denotes greater than 20 gpm minimum flow capacity but installed valve capability is 400 gpm
  • At time of event, two trains of RHR were in service
  • Each train had operable Discharge Relief Valve
  • Each valve "sees" full RCS pressure and transients
  • Each valve capable of relieving Charging Pump discharge
  • NRC event trees did not credit this capability
  • TVA failure probability 2.42 E-5 from PLG-0500
  • Conclusion - WBN value is Current Licensing Basis for this component 22

Difference No. 5 Available Secondary Plant Cooling

  • In the event one of the COMS valves would fail to reseat, a loss of inventory that would impact RHR would be postulated
  • Operators would be directed to AOI-14 "Loss of RHR Shutdown Cooling"
  • AOI sends the operator to section 3.8 (reactor head on) if RHR cooling cannot be restored
  • With an RCP available Section 3.8 Step 2a directs use of steam dumps or Steam Generator PORV to restore cooling Steam Generators at 75% level during event Four Steam Generator PORVs were available Two motor-driven Auxiliary Feedwater Pumps were available Condensate Storage tank was full 23

Difference No. 5 Available Secondary Plant Cooling

  • NRC did not credit this capability May not normally be available in generic model for shutdown operation
  • TVA assumed operator failure probability associated with a moderate to high stress, but procedure driven activity
  • Operators had Just-in-Time Training on AOI-14 prior to the outage including use of Secondary Plant cooling
  • Tested successfully on WBN Simulator
  • AFW /SG PORV cooling was part of WBN's IPE submittal under Generic Letter 88-20 which was approved in SER dated October 5, 1994

Conclusion:

Use of Secondary Plant cooling is consistent with WBN Current Licensing Basis 24

Summary of Impact on NRC Event Tree 1 5 PORV Lifts Vice 7 In The NRC Event Tree 7 5 2 PORV - Rigorous treatment of successive lifts 4.2 E-2 2.01E-2 3 Available RHR Suction Relief Valve IE-3 2.42E-5 4 Two Available RHR Discharge Relief Valves N/A 1.21E-6 5 Available Secondary Plant Cooling N/A 5.0-E-4 25

"Adjusted" NRC Event Tree COLD OPS372 OPS RHR SUCTION RIR SUCTION I 0F2 RIIR RIR OP STOPS OVERPRESSURE PSIG RESEATS RV LIFTS RV CLOSES DISCHARGE DISCHARGE PUMPS OR RV OPEN RV CLOSES OPENS PORV COP OPS OPS-CLS RHR-S-RV RIR-S-CL OP-RECOV END-STATENANES I OK 2 RCS-BLOWDOWNRHR-OK 3 OK 5 3E6E4 4 BLOWDOWN4LOCA-(MI!3) OK 0.02 (4.27E-3) 4a BLOWVDOWN-LOCA-(4.92E.08) 1FiA 2.45E-5 5 OK 1.21E-6 6E-13 1.0 6 ISLOCA4UIR-SYSTEM-(4a) COP-Cold Overpressure 2005/07/18 Page 1 26

"Adjusted" NRC Event Tree Loss of RCS Injection Isolate 74-1 or Open Block Open PORV Charing Inventor before 74-2 Valve Available LOI FEED END-STATE-NAMES I OK 2 OK 0.5 OK 1E-2 1E-3 3 CD-{5E6) 3E-9 1)A 6E4 1E-2 OK 1E-3 4 CD-(1IES 7E-9* See Next Slide for Explanation 1E-4 CD51." 6E-WATTS BAR LOCA TREE - COP Scenario 4 2005/07/18 Page2 27

I "Adjusted" NRC Event Tree Expanded 1E-2 Path Loss Of RCS Injection Isolate 74-1 Open Block Open PORV RWST Refill SGEN Cool Inventory before or 74-2 Valve LOI FEEDE LO FED# ENDST@ATE-NAMES 9.94E-1 A OK 9.97E-1 A OK 9.99E-1 B OK lE+0 6E-3 lE-3 C CD -3.59E-9 9.99E-1 D OK 3E-3 I1E-3 E CD - 1.80E-9 1E-2 0.99 F OK ijjA 6E-4 3E4 1EA4 1E-02 C CD - 1.80E-9 WATTS BAR LOCA TREE - COP Scenario 4 2005107118 Page 2 Total 7E-09 28

I I "Adjusted" NRC Event Tree Loss of RCS Injection Isolate RIIR RHR RECOV RWST Inventory before open PORV Before RWST Makeup Before Depletes LO FEED ISOLATE RIRREC RWSTMU END-STATE-NAMES I OK 2 OK 1.0 IE-2 3 CDt4E)6E-5 4e-5 6E-13 1E-2 4 Cl+)6E-15 5E4 5 CDVE4)3E-16 WATTS BAR LOCA TREE - COP Scenario 6 2005/07/18 Page 2 NRC Total ICDF 2A 5 7E-8 29

I4 PSA Additional Conservatisms

  • Letdown Heat Exchanger Relief Valve not credited
  • Operator action per AOI-14 to stop charging pump not credited
  • Operator action per AOI-14 to open isolated PORV not credited
  • Potential for gas relief only on early PORV lifts not credited
  • Additional letdown flow based on increasing RCS pressure not credited
  • PORV is more reliable than in NRC evaluation 30

V Regulatory Summary

  • WBN PSA inputs - "reasonable and realistic"
  • WBN used values from event analysis or from Current Licensing Basis which were reviewed and approved by NRC
  • Five key differences with NRC evaluation
1. There were 5 PORV lifts that relieved pressure rather than the 7 in the NRC event tree
2. More rigorous mathematical treatment of each successive PORV lift is warranted
3. RHR suction relief valve more reliable than in NRC's evaluation
4. Two additional RHR discharge relief valves were available to relieve increasing RCS pressure
5. Secondary plant cooling was available to prevent core damage
  • Using "reasonable and realistic" values in NRC event trees provides an adjusted result of 7E-8
  • Conclusion - very low safety significance - Green 31

Closing Remarks 32

Ti NRC Event Tree COLD OPS372 OPS RESEATS RImR SUCTION RHR SUCTION OP STOPS OVERPRESSURE PSIG RV LIFTS RV CLOSES PUMPS OR COP OPS OPS-CLS RR-S-RV RIIRf-SCL OP-RECOV END-STATE-NAMES Difference No.I 1 0.1 2 lRCSBLOWOWNR OKl 3 OK l3E-2 4 BLOWDOWWNAXAHRIE. 6E-3 Difference No l 1E-3 1.0 C -Difference No.2 P 6 ISLOCA-RHR1YSTEM-Ed" COP-Cold Overpressure DifrneN.42005/07/18 Page I

) NRC Event Tree Loss of RCS Injection Isolate RIR RIMR RECOV RWST Inventory before open PORV Before RWST Makeup Depletes Before LOt FEED ISOLATE RIRREC RWSTMU END-STATE-NAMES I OK 0.51 1E-2 3 CD-{s IE-3 1E-2 lDifferenceNo.5l 4 CD-(I"E5) 1E-4 5 CD.(IE-7) WATTS BAR LOCA TREE - COP Scenario 4 2005/07/18 Page 2

0 4 NRC Event Tree Loss Of RCS Injection Isolate RIR RIIR RECOV RWST Inventory before open PORV Before RWST Makeup Before _ _ _ _ _ _ _ _ _ _ _ _ _ _D ep letes LOt FEED ISOLATE RHRREC RWSTMU END-STATE-NAMES I OK 2 OK 1.0 1E-2 3 CD.(4E-7) 4E-5 1E-2 4 CD-(4E-7) 5E4 5 CD-(2E4) WATTS BAR LOCA TREE - COP Scenario 6 2005/07/18 Page 3}}

Retrieved from "https://kanterella.com/w/index.php?title=ML052940341&oldid=5374638"