ML20141D720

Summary of 970611 Meeting W/Parsons Power Group to Discuss Process to Be Used During Tier 2 Accident Analysis Review of ICAVP for Plant Unit 2.List of Attendees & Handout Encl

ML20141D720
Person / Time
Site:	Millstone
Issue date:	06/24/1997
From:	John Nakoski NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9706270316
Download: ML20141D720 (35)

Text

_

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Juns 24,1997

. j LICENSEE:

NORTHEAST NUCLEAR ENERGY COMPANY (NNECO)

FACILITY:

Millstone Nuclear Power Station Unit 2 i

SUMMARY

OF THE JUNE 11,1997, MEETING WITH PARSONS POWER GROUP TO DISCUSS THE PROCESS TO BE USED DURING THE TIER 2 ACCIDENT ANALYSIS REVIEW OF THE ICAVP FOR MILLSTONE UNIT 2 i

On June 11,1997, the Special Project Office (SPO) staff of thy Office of Nuclear Reactor j

Regulation (NRR) participated in a publicI9 observed meeting with Parsons Power Group (Parsons) representatives. The purpose of this meeting was to discuss the process to be used

.E by Parsons to implement the Tier 2 accident analysis review of the Independent Corrective

}

Action Verification Program (ICAVP) at Millstone Unit 2. During this meeting, Parsons used its preliminary review of the Main Steam Line Brea,k accident as described in Chapter 14 of Unit i

2's Final Safety Analysis Report (FSAR) to faailjtatQsysg s, qgarding its processes for the i

Tier 2 review. Also discussed during the meeting were1He de of the system review for the i

systems directly required to mitigate the apalyzed accident, the review required of supporting or i

interfacing systems, and the information (critical charadrisiics) for each of the systems that j

requires NRC review and approval prior to verification by Parsons.

I l provides a list of the attendees at the meeting. Enclosure 2 provides the handout used by Parsons as the outline for discussions during the meeting with the NRC. Enclosure 3 l

provides information used during the meeting to discuss and demonstrate the process Parsons proposed to use to conduct the Tier 2 review from its preliminary analysis of the Main Steam Line Break accident. Enclosure 4 provides examples of the information (critical characteristics) 3 i

that Parsons would typically propose to the NRC for review and approval. During the meeting, i

the NRC indicated that information provided in section 2.3.5 of Enclosure 4 was the type of I

information that would be necessary for the NRC to receive from Parsons to support its review

)-

and approval of the critical characteristics for each of the systems involved in mitigating the consequences of the analyzed accidents. Also, the NRC staffindicated that the processes l

proposed by Parsons for the ICAVP Tier 2 review at Unit 2 appeared reasonable, but will be i

subject to further review and approval with the U nit 2 ICAVP audit plan provided by Parsons.

l Msg Signed by l

John A. Nakoski, ICAVP Program Coordinator lCAVP Oversight Branch Special Projects Office

}

Office of Nuclear Reactor Regulation Dockets No. 50-336

Enclosures:

As Stated (4)

I cc w/att: See next page

{'

DISTRIBUTION:

j HARD COPY Docket File r SPO reading SPO-L reading i

PUBLIC WLanning, RI OGC (w/o encls 2, 3, & 4)

Elmbro JDurr, RI ACRS (w/o encls 2,3, & 4) j SReynolds RPerch DBeaulieu, RI 4

f E-MAIL (w/ encl 1 only) lllllll llfllllll l j

SCollins/FMiraglia PMcKee WDean (WMD) 2 RZimmerman DMcDonald Dross (e-mail to SAM)

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l LIST OF ATTENDEES l

June 11,1997 NAME ORGANIZATION POSITION l

Eugene Imbro NRC Deputy Director, ICAVP Oversight, SPO, NRR Steve Reynolds NRC Chief, ICAVP Oversight Branch, SPO, NRR John Nakoski NRC ICAVP Program Coordinator, SPO, NRR Peter Koltay NRC Unit 3 Team Leader, SPO, NRR Daniel L. Curry Parsons Power Project Director Eric A. Blocher Parsons Power Deputy Project Director John F. Hilbish Parsons Power Regulatory Review Group Manager Wayne L. Dobson Parsons Power Process Model & Operational Analysis Manager Randy Faust Parsons Power Accident Mitigation System Reviewer Rich Glaviano Parsons Power Accident Mitigation System Review Lead Mike Akins Parsons Power Accident Mitigation System Review Lead John loannidi Parsons Power System Review Manager Bruce Deist Parsons Power System Review Group Paul Shipper Parsons Power Accident Mitigation System Reviewer Abdul M. Ahmed Parsons Power Accident Mitigation System Reviewer Juan M. Cajigas Parsons Power Accident Mitigation System Reviewer Gordon Chen Parsons Power Accident Mitigation System Reviewer R. Wayne Choromanski Parsons Power Accident Mitigation System Reviewer William E. Meek Parsons Power Advisory Panel Member Ed House BWG, Inc.

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MSLB Generic CDC Prehminary 4

Review By DBEv 3

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FSAR Chapter 14 DW M Define Each g y.

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Boundary Diagrams r

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d Review of Main Steam Line Break at HZP with Offsite Power Available The attached information was developed from review of the Main Steam Line Break (MSLB) Design Basis Event (DBEv) for the Millstone-2 Plant. This information was assembled to demonstrate the process that will be used to a

determine the critical system and component level characteristics resulting from the Tier-2 Accident Mitigation Systems Review. The information was extracted primarily from Chapter 14 of the Millstone-2 FSAR and from the Millstone-2 Operations Critical Drawings. The specific analyses supporting the MSLB DBEv, along with plant operating procedures for the involved systems, were not available at the time this information was prepared. The information presented herein may change as additional design and operating l

information is obtained from the supporting analyses and plant procedures.

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T l _ _ _ _ Analyses & Calcs Chapter 14 Develop Determine Extract System Critical i

CSF for Design Extract Design l

Characteristics

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Main Steam Line Break at HZP with Offsite Power Available i

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Event Description The Main Steam Line Break (MSLB) event at Millstone-2 'is a double-ended guillotine break inside containment between the steam generator and the flow restrictors.

Initial Conditions Plant is critical at Hot Zero Power (HZP) i Main Steam Isolation Valves (MSIV's) are open Steam generators are being fed by the AFW system using the motor driven pumps. The AFW control valves are in a fixed position to provide flow sufficient to remove Reactor Coolant Pump heat.

Offsite power is maintained throughout the event.

Once HPSI and one charging pump are assumed available.

The most reactive cont ol rod is assumed to be stuck in its fully withdrawn position.

E

MNPS-2 FSAR TABLE 14.1.5-6 STEAM LINE BREAK ANALYSIS

SUMMARY

l Maximum

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Initial Offsite Post Scram Maximum Power Power Retum to LHGR Level Available Power (MWt)

MDNBR (kW/ft) fkN HZP Yes 686 2.40

< 21.0 HZP No 294 1.18 16.5 HFP Yes 394 3.00 17.1 3

HFP No 147 4.60 5.7 1

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MNPS-2 FSAR TABLE 'a 4.1.5-7 STEAM LINE BREAK SEQUENCE OF EVEfES-HOT ZERO POWER-POWER AVAILABLE Time' Event O.

Reactor at hot zero power.

O. +

Double-ended guillotine break located between affected steam generator and the flow restrictors.

3.6 Main steam isolation valve closure signal generated by low steam generator pressure.

10.5 Main steam line isolation valves stop blowdown from intact steam generator 6.9 seconds after low steam generatoi pressure sign'al.

15.2 Safety injection signal generated by low primary coolant pressure, a

32.

Reactor becomes critical.

45.2 HPSI and charging pumps actuated.

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153.

Thermal power reaches maximum level at 25% of rated power.

i 153.

First boron has passed through core.

l 180.

Auxiliary feedwater initiated to affected steam generator.

600.

Auxiliary feedwater isolated manually.

600.+

Primary system temperature increase due to steam generator dryout and additional boron injection will terminate power excursion.

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' Time after break, seconds 14815 7.MP2 1 of 1 October 1994 l

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• 2.3 MAIN STEAM LINE BREAK (MSLB) ACCIDENT, RCS ANALYSIS.

2.3.1 ACCIDENT DESCRIPTION This event is initiated by a rupture in the main steam piping upstream of the MSIVs v/hich results in an j

uncontrolled steam release from the secondary system. The increase in energy removal through the secondary system results in severe ove cooling of the primary system. In the presence of a negative Moderator Temperature Coefficient (h iTC), this cooldown causes a decrease in the shutdown margin

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(following reactor trip) such that a retum to power might be possible following a steam line rupture assuming that the most reactive control rod is stuck in its fully withdrawn position. The MNPS-2 limiting MSLB from a safety standpoint is a Hot Zero Power (HZP) double-ended guillotine break inside containment between the steam generator and the flow restrictors. Regulatory requirements require that the plant be equipped with an emergency core cooling system (ECCS) diat refills the vessel in a timely manner to satisfy the requirements of 10CFR50 Appendix A GDC 27,28,31, and 35 as well as appropriate sections of NUREGs 0694,0718, and 0737. The MNPS-2 MSLB-RCS analysis is described in FSAR j

Section 14.1.5.1.

2.3.2 DESIGN BASIS j

The MNPS-2 MSLB-RCS analysis is based on the following pnmary assumptions:

i j

a. Most reactive control rod stuck in its fully withdrawn position.

b. Rated power mode is bounding for all full power modes and Mode 2 is bounding for HZP.

c. Single failure criteria for offsite power case is loss of one HPSI pump.

d. Single failure criteria for LOOP case is loss of one diesel generator.

e. Safety injection actuation signal (SIAS) actuated by low pressurizer pressure.

l

f. Secondary isolation signal actuated by low steam pressure.

i j

2.3.3 SYSTEM INTERFACE 1

Tne following systems interface during the postulated MSLB-RCS recovery analysis:

a. Safety Injection System

b. Shutdown Cooling System

c. Emergency Power System

d. Auxiliary Feedwater System

e. Main Feedwater System (hot full power case) 2.3.4

SUMMARY

OF DESIGN INPUTS The following design inputs and assumptions have been identified during the FSAR review of the MSLB-RCS analyses. Additional inputs / assumptions and/or revisions will be developed upon review of the corresponding analysis calculation packages.

2.3.4.1 Control Rods Most reactive control rod to be stuck in its fully withdrawn position.

Reference:

FSAR Section 14.1.5.2 2.3.4.2 Power Mode - Full Power Rated power mode bounding for all HFP modes.

Reference:

FSAR Section 14.1.5.4

.- ~

s 4

4 2.3.4.3 Power Mode - Zero Power Mode 2 boundmg for all HZP modes. Four RCPs assumed to operate to maximize initialloop flow.

Reference:

FSAR Section 14.1.5.4 2.3.4.4 Single Failure - Offsite Power Available

~

2 One HPSI pump available.

l

Reference:

FSAR Section 14.1.5.4 and 14.1.5.5.1.3 5

2.3.4.5 Single Failure - LOOP Lose one DG with consequential loss of one HPSI pump and one charging pump.

Reference:

FSAR Section 14.1.5.5 and 14.1.5.5.1.3 2.3.4.6 Limiting Break Double-ended guillotine break mside contamment between SG and flow restrictors.

2 A(affected SG) = 6.31 ft and A(intact SG) = 2.35 ft

Reference:

FSAR Section 14.1.5.5.1.1, Table 14.1.5-3

(.

t 2.3.5 MSLB ANALYSIS -HZP WITH OFFSITE POWER AVAILABLE Reactivity Control System Reauirement - Insert control rods within 3.9 seconds of reaching reactor trip setpoint.

s Reactor Trio Delav-3.9 seconds 3.0 sec. insenion time plus 0.9 sec instrument delay. Setpoint on low steam pressure or low pressurizer pressure.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-4 Additionalinfo:

What evwt starts the 3.9 second clock?

Does the 3.0 seconds include the CRD breaker time?

Is the rod insertion requirement to 0% or some otherposition ?

Validation: Plant surveillance data for RPS testing, CRD breaker testing, and control rod drop times.

System Reauirement - Inject boron to reach the core by 153 seconds.

j Boron Iniection - Assumed from (1) HPSI pump en:1 (1) Charging pump, both taking suction from the RWST.

Reference:

FSAR Section 14.1.5.5.1.2, Table 14.1.5-3 Additionalinfo:

Versfy suction sourcefor charging pumps at the beginning ofthe event. Is it the RWST or the VCT7 Is the VCT a conservative source? If the VCT, then how long to switch over to the RWST?

Clarify the meaning of the comment in section 14.1.5.5.1.3 that describes " crediting charging" as not invalidating the conclusions of this analysis.

Validation: Charging system - System lineup and pump capacity from plant surveillance test. HPSI system -

System lineup and pump capacity from plant surveillance test.

RWST Boron Concentration - 1720 PPM.

Reference:

FSARTable 14.1.5 3 Validation: TS value is 1720 PPM per TS 3.1.2.8. No further validation required.

9 RCS Heat Removal i

System reauirement - close MSIV on intact main steam line within 10.5 seconds to limit cooldown from non-affected steam generator blowdown.

j Low Steam Line Pressure Trio Simm1 - FSAR indicates " analysis setpoint" of 500 psia, " uncertainty" of-22 psia for a "value" of 478 psia.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-4 Additionalinfo: Clarify setpoint and basisfrom the analysis.

Data to establish signal development time.

Validation: TS allowable is 2492.5 psia., TS Table 3.3-4 Need to validate with TS and surveillance requirements once analysis value can be determined. Validate signal development time using time response j

test data.

l MSIV Closure Delay - 6.9 seconds. Time from trip setpoint to full valve closure. Setpoint on low steam pressure.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-4 l

Validation: TRM value is s 6.9 sec., TRM Table 3.3-5. Further validation required?

System Reauirement - Limit AFW flow to affected steam generator for first 180 seconds of event.

AFW Flow - Initialized to match RCP heat. Then allowed to increase based on a fixed CV setting. Flow increased to pump runout flow,229.5 lbm/sec, at 180 sec.

l

Reference:

FSAR Section 14.1.5.5.1.4, Table 14.1.5-3 Additionalinfo:

Is valve position controlled by the steam generator level control system at HZP7 Ifso, does the control valve open to maintain steam generator level? Ifso, is this conservative?

Are both motor driven AFWpumps assumed to be operating as an initial condition?

Validation: For t<180 sec, calculate flow and compare with analysis value for " fixed CV setting." For t>l80 sec, compare pump flow at runout with analysis value.

System reauirement - Isolate AFW flow to affected steam generator at 600 seconds.

AFW flow control valve - Close valve from control room.

Reference:

FSAR Table 14.1.5-7 Validation: AFW surveillance test.

AFW Temocrature - Limiting AFW temperature assumed AFWTi = 32.1*F

Reference:

FSAR Table 14.1.5-3 Validation: None required.

0

t RCS Pressure & Inventory Control System reauirement - initiate HPSI flow to core within 30 seconds per assumed HPSI pump head curve.

d Low Pressurizer Pressure Trio Signal - FSAR indicates " analysis setpoint" of 1600 psia, " uncertainty" of -22 psia for a "value" of 1578 psia.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-4 Additionalinfo: Clarify setpoint and basisfrom the analysis.

Data to establish signal development time Validation: Need to validate with TS and surveillance requirements once analysis value can be determined.

TS allowable is 21592.5 psia., TS Table 3.3-4. Validate signal development time using time response test l

data.

HPSI Actuation Delay - 30 seconds. Time from trip setpoint to full pump speed. Setpoint on low pressurizer pressure.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-4 Validation: Need to validate with plant surveillance requirements. Note that TRM value is s 25 sec., TRM Table 3.3-5

)

HPSI Pumo Performance - Provided as Figure 14.1.5-1, HPSI flow vs. RCS backpressure curve.

I

Reference:

FS AR Section 14.1.5.5.1.2, Figure 14.1.5-1.

I Validation: Compare to plant ISI procedure acceptance criteria.

l System reauirement - initiate charging flow to core within 40 seconds per assumed pump capacity, i

Charnina Pumo Actuation Delav - 40 seconds. Time from trip setpoint to full pump speed. Setpoint on low pressurizer pressure.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-4, TRM Table 3.3-5 Validation: Compare to plant ISI procedure acceptance criteria.

Charzinz Pumo Performance - No data provided in FSAR.

Reference:

Additionalinfo: Chargingpumpflow used in analysis RWST Temocrature -

i

Reference:

I Additionalinfo:

What RWST temperature was usedfor this analysis?

Is it importantfrom an RCSPressure/ Inventory controlperspective?

Validation: Compare analysis value to TS requirement.

2.3.6 MSLB ANALYSIS - HZP WITH LOOP Additional rev w: diesel generator start and loading.

i 2.3.7 MSLB ANALYSIS - HOT FULL POWER WITH OFFSITE POWER AVAILABLE FW Temocrature - All FW heating ceases at time of the break. Limiting FW temperature assumed FWTi =

432.I'F

Reference:

FSAR Table 14.1.5.3 FW Flow - Prior to FW flow temunation, FW flow is a function of secondary system pressure. No FW flow vs. secondary system backpressure provided in FSAR.

Reference:

FSAR Section 14.1.5.5.1.4 FW Figw - FW termmated at 30 seconds after the reactor trip per closure of FW regulator valves.

Reference:

FSAR Section 14.1.5.5.1.4 Main FW Valve Closure Delav - 30 seconds. Time from trip setpoint to full valve closure. Setpoint on low steam pressure.

Reference:

FSAR Section 14.1.5.5.1.5, Table 14.1.5-1 TRM value is s 14 sec., TRM Table 3.3-5 2.3.8 MSLB ANALYSIS - HOT FULL POWERWITH LOOP Additional review: diesel generator start and loading.

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