ML033100141

Technical Specification Proposed Change No. 263 - Supplemental No. 3, Extended Power Uprate - Updated Information

ML033100141
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	10/28/2003
From:	Thayer J Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 03-98
Download: ML033100141 (69)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

~~~~~~~~~~~~~~~~~~~~~Vermont Yankee 322 Governor Hunt Rd.

PEntergy P.O. Box 157 Vernon, Vr 05354 Tel 802-257-771 1 October 28, 2003 BVY 03-98 U.S. Nuclear Regulatory Commission A1TN: Document Control Desk Washington, DC 20555

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specification Proposed Change No. 263 - Supplement No. 3 Extended Power Uprate - Updated Information By letter dated September 10, 20031 and initially supplemented by letter dated October 1, 20032, Vermont Yankee3 (VY) proposed to amend Facility Operating License, DPR-28, for the Vermont Yankee Nuclear Power Station (VYNPS) to increase the maximum authorized power level from 1593 megawatts thermal (MWt) to 1912 MWt. The letters dated September 10, 2003 and October 1, 2003, transmitted certain attachments which VY is updating herewith.

VY is providing three attachments to this letter: (1) an update to Attachment 3 of the September 10, 2003 submittal, which addresses VY's extended power uprate testing and modification plans; (2) an update to of the September 10, 2003 submittal, which provides the justification for an exception to large transient testing; and (3) an update to Attachment I of the October 1, 2003 submittal, which is a review matrix that cross-references the criteria of NRC review standard RS-00 14 for extended power uprates with the information in the VYNPS Constant Pressure Power Uprate Safety Analysis Report5 and the NRC-approved generic topical report for constant pressure power uprate6. Each of the attachments

' Vermont Yankee letter to U.S. Nuclear Regulatory Commission, "Extended Power Uprate," Proposed Change No. 263, BVY 03-80, September 10, 2003.

2 Vermont Yankee letter to U.S. Nuclear Regulatory Commission, "Extended Power Uprate - Technical Review Guidance," Proposed Change No. 263, Supplement No. 1, BVY 03-90, October 1, 2003.

3 Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. are the licensees of the Vermont Yankee Nuclear Power Station.

4 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, "Review Standard for Extended Power Uprates," RS-001 (Draft), December 2002.

5 GE Nuclear Energy, "Safety Analysis Report for Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate," NEDC-33090P, September 2003.

6 GE Nuclear Energy, "Constant Pressure Power Uprate," Licensing Topical Report NEDC-33004P-A (proprietary),

July 2003, and NEDO-33004-A (non-proprietary), July 2003.

1o)

BVY 03-98 / Page 2 retains its previous numerical designation to minimize confusion and completely replaces the earlier version; thus, no other aspects of the license amendment request are affected.

If you have any questions with this submittal, please contact Mr. Len Gucwa at (802) 2584225.

Sincerely, ka~yW~hyer Site Vice President STATE OF VERMONT

)

)ss WINDHAM COUNTY

)

Then personally appeared before me, Jay K. Thayer, who, being duly sworn, did state that he is Site Vice President of the Vermont Yankee Nuclear Power Station, that he is duly authorized to execute and file the foregoing document, and that the statements therein are true to the best of his knowledge and belief.

Silly A. Sarldstrum, Notary Public N

tOTARN My Commission Expires February Io00 Attachments (3)

PUBLIC cc:

(with attachments)

USNRC Region 1 Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS (two copies)

Vermont Department of Public Service

Docket No. 50-271 BVY 03-98 Attachment Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 263 Supplement No. 3 Extended Power Uprate - Updated Information Modifications and Testing

BVY 03-98 / Attachment 3 / Page I EXTENDED POWER UPRATE Modifications and Tests The following is a list of currently planned modifications necessary to support extended power uprate (EPU) for Vermont Yankee Nuclear Power Station (VYNPS). These modifications will be implemented during the next two refueling outages (i.e., the scheduled refueling outages beginning in the Spring of 2004 (RFO-24) and Fall 2005 (RFO-25). The following modifications constitute planned actions on the part of Vermont Yankee. Further evaluations may identify the need for additional modifications or, on the contrary, obviate the need for some modifications.

As such, this list is not a formal commitment to implement the modifications exactly as described or per the proposed schedule. Additionally, various minor modifications and adjustments to plant equipment, which may be necessary, are not listed.

BVY 03-98 / Attachment 3 / Page 2 ERMO NT YANKEE CPPU MODIFICATIONS/TESTING

,,,.,,;,:,;,.,,,...i, ;. :,...i,

,..; :;,..;~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~~ ~~~~~~~~~..........

Main Turbine Modifications include:

HP Turbine testing to include:

Replace HP Turbine steam Overspeed testing.

path:

Control Valve and Stop New control valve settings.

Valve testing.

Modify control valve As found and as left operating mechanism with performance test.

5% margin above CPPU conditions.

Modify turbine control and overspeed setpoint for CPPU conditions.

Replace 8h Stage diaphragms of the L.P. Turbine (Note:

This modification will be implemented in RFO-25).

Main Turbine Cross-Install higher capacity relief valves.

Relief valves to be bench tested around Relief Valve prior to installation.

(CARV) Discharge Piping Main Generator System Rewind/Upgrade the Main Generator Factory to perform applicable for CPPU conditions. Replace electrical testing of windings.

bushing current transformers. (Note:

Generator to be performance bushing current transformers to be monitored.

replaced in RFO-25).

Main Generator Replace Generator Hydrogen Coolers Performance monitoring.

Cooling Hydrogen with upgraded coolers.

System Isolation Phase Bus Install a new Isolation Phase Bus Performance monitoring.

Duct Cooling Duct Cooling System to remove Bus Duct heat under CPPU conditions.

HP Feedwater Heater

1. 1A, #1B, #2A, & #2B FW Heater Testing to include:

Replacement Replacement.

Pressure testing.

Visual Inspection.

Magnetic Particle testing.

Radiography.

In-service inspection.

Demonstration of thermal performance.

BVY 03-98 / Attachment 3 / Page 3 VERM:ONT YANKEE: CPPU MODIEICATIONSTESTIN I a ion,".

~

~

~

~

~

I

~;'Modificatio ecito etn

.~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~;:jj A:

Steam Dryer Any identified modifications needed Performance Monitoring for Steam to maintain steam dryer structural Dryer Cover Plates integrity integrity at CPPU conditions.

includes:

• Main Steam Line flow indication (check for unbalance).

RPV water level indication (check for unbalance).

Steam dome pressure (check for sudden drop).

Moisture carryover (unexpected step increase)

RHRSW System Modify RHRSW Pumps (Train A &

Testing of piping performed per B) Motor Bearing Oil Coolers to modification to include:

recuperate Service Water flow to the Visual inspection.

coolers.

Particle testing.

Ultrasonic flow testing.

In-Service inspection.

NSSS/BOP Instruments Upgrade specific NSSS/BOP Perform instrument resealing, Instruments for CPPU conditions.

calibration and functional testing.

NSSS/Torus Attached Upgrade particular NSSS and Torus As applicable, welds to be Piping attached piping supports.

examined visual, liquid penetration and magnetic penetration methods.

Flow Induced Vibration Install/remove FIV Instrumentation.

Collect FIV background data and (FIV)

FIV CPPU data and analyze data.

Reactor Recirculation Permits continued reactor power Modification testing to be (RR) System operation by Recirculation Pumps performed with breakers in "test speed running back to a preset position" and RR System not demand if reactor is operating at or operating.

greater than a predetermined power level and one Feed Pump trips.

Main Condenser Tube stake Main Condenser tubing to Perform tube leak testing per the reduce the effects of flow induced modification by Main Condenser vibration.

flood-up.

BVY 03-98 / Attachment 3 / Page 4 VERMONT YANKE~,CPPU;MODFCATONS/TESTING ModKficatio Descr.

T s ing Condensate Install a Condensate Demineralizer With filtered bypass strainer in Demineralizer Filtered Bypass Strainer to permit one service, monitor flows under demineralizer to be removed under various CPPU conditions.

CPPU conditions.

Feedwater System Protect Feed Pumps with two Normal testing to be performed sequential levels of low suction per modification testing, to be pressure trips at various time delays performed with breakers in "test to ensure only one pump trips at a position."

time.

Cooling Tower Replace fan blades with more Cooling Tower performance Fans/Motors efficient blades and drive motors with monitoring.

upgraded higher performance motors.

Core Spray & RHR Core Spray and RHR pump seals may Leak check at pump rated Pump Seal require replacement.

conditions.

Replacements (Contingency)

EQ Upgrades Re-route feed to SRV monitor to new Voltage check and meggar.

breaker.

BVY 03-98 / Attachment 3 / Page 5 AGGREGATE IMPACT OF CPPU MODIFICATIONS TO DYNAMIC PLANT

RESPONSE

The modifications listed on pages 2, 3, and 4 of this attachment were reviewed to ensure the aggregate impact of the modifications do not adversely impact the dynamic response of the plant to anticipated initiating events. This review has concluded that there is no adverse impact to the dynamic response of the plant to anticipated initiating events as a result of these plant modifications. A discussion of the modifications and their affect on integrated plant response is provided below.

All of the modifications listed in the "Vermont Yankee CPPU Modifications/Testing Table", with the exception of the Reactor Recirculation (RR) System, RHRSW System, Condensate Demineralizer, and Feedwater System, enhance and/or upgrade the existing plant components to allow for operation at CPPU conditions. With the exception of allowing operation at CPPU conditions, these modifications do not change the design functions of the equipment or the method of performing or controlling the function. Therefore, these modifications will not result in a significant change to the plant's dynamic response to anticipated initiating events.

The RR System modification provides an automatic runback of the recirculation pumps speed to a preset demand if the reactor is operating at or greater than a predetermined power level (-112%

CLTP) and any one condensate or reactor feed pump trips. The runback lowers reactor power to be within the capability of the feedwater system with only two reactor feed pumps running so that reactor water level can be restored and maintained within the normal operating range. (Note: trip of a condensate pump at these power levels will result in the trip of a feedwater pump due to low feedwater pump suction pressure). Under current plant design, only two of the three reactor feedwater pumps are required for operation at CLTP. If one pump trips, a standby pump starts.

If a standby pump is not available or does not start, operators would lower reactor power by reducing core flow. This modification turns a manual operator action, reducing core flow by lowering recirculation pumps speed, into an automatic action. The rate of the runback is within the current design of the control system. The plant dynamic response to a reactor feedwater pump trip is not significantly affected.

The RHRSW System modification reroutes the outlet of the RHRSW Pump Motor Bearing Oil Coolers so that the cooling water is directed back to the deep basin vice the current lineup which directs the outlet to a storm drain. This modification does not affect the cooling water flow rate to the motor bearing oil coolers and has no affect on any other system or component relied upon to mitigate anticipated initiating events. The plant dynamic response to any anticipated initiating events is not affected.

The Condensate Demineralizer modification installs a filtered bypass strainer to permit one demineralizer to be removed from service under CPPU conditions. Since feedwater system flow remains constant whether or not the filtered bypass strainer is in service, plant dynamic response to any anticipated initiating event is not affected.

The feedwater system modification provides two sequential levels of low suction pressure trips at various time delays to ensure only one reactor feed pump trips at a time. This modification

BVY 03-98 / Attachment 3 / Page 6 improves the reliability of the feedwater system. Therefore, it presents no adverse affect on plant dynamic response to any anticipated initiating events.

BVY 03-98 / Attachment 3 / Page 7

.CPPU POWER ASC io TS P LAN:

ST TEST DESCRIPTION PRIOR TO, TESTMESCRARTU PERCENT POWER CU.P (Allowance +4,13-3%).

(Al1lowance +0% 1%)

% 5 10 15

-20 25 30

.35 40 45 50 - 55 60 65 70 75 80 85 90 95 100 105 110 115 120 Main Turbine Overspeed testing &

backup overspeed testing.

x Main Turbine Demonstration of thermal performance Improvements and generator increase.

Main Turbine Perform CPPU performance monitoring of the X

X X

X X

X X

X Main Turbine.

Flow Induced Collect and analyze Vibration vibration data prior to power Increases.

X X

X X

X X

X Condensate Monitor the Demineralizer Condensate Filtered Demtneralizer Bypass Filtered Bypass X

X X

X X

Strainer flow Strainer and testing demineralizer flows Core Thermal Perform a Heat Limit Balance Calculation.

X X

X X

X X

Verification Cooling Tower Perform CPPU Modification performance x

x x

x x

x x

x monitoring.

Main Perform CPPU Generator performance x

x x

x x

x x

x Modification monitoring.

Hydrogen Perform CPPU Cooling performance Modification monitoring.

X X

X X

X X

X X

Iso Phase Bus Perform CPPU Duct performance Modification monitoring.

X X

X X

X X

X X

BVY 03-98 / Attachment 3 / Page 8 CPPU POWER ASCENSIONTEST PLAN":

TEST, TEST DESCRIPION PR R T;.

PERCENT POWER CLTP (Allowance,40%403%),

(Allowance 40% -%-)

STARTUP

% 5 10 15 20 25

.30 35 40' 45

'50 55 60

.65 70 75-80-85 90 95 100 105 110 115 120 HP Feedwater Perform Heater demonstration of Modification thermal performance.

X X

X X

X X

X X

Steam Dryer Monitor for steam Modification dryer ntegrity.

X X

X X

X X

X X

Steam Moisture carryover dryer/separator performance performance monitoring dryer X

X X

X:

X separator Integrity.

BOP Performance Monitoring monitoring of BOP X

x x

x x

x x

x Systems.

Radiation Perform radiation surveys.

surveys at various x

x x

x x

power levels.

IRM IRM/APRM overlap Performance will be done during the first controlled shutdown following APRM Calibration for EPU. If this Is not possible, perform during next startup.

APRM Calibrate each APRM Calibration channel to be consistent with the core thermal power, referenced to the x

LPU level, after the receipt of the CPPU SER from the NRC.

BVY 03-9 8 / Attachment 3 / Page 9 TEST`

-TESTDESCRIPTnON PI TOPRCENT POWER CLTP (AliowanceO'.-%

(llwac

+%1%

STARTUP

% 5

~~~~~~~10

'15.

201 25

30.

35'1 40 4

0 5,'6 5

0 7

80 85 90, 95 100 105 110 115 120 Core Measure reactor and Performance system parameters, calculate core thermal power and core performance X

X X

X X

X parameters, evaluate data and project next power step's values.

Pressure Test and dynamically X

Regulator calibrate the pressure Tuning and regulator system prior Testing to start-up.

Pressure Average Main Steam Regulator-line flow versus Incremental pressure regulator Regulation output Data to be x

x x

x x

Data Gathering taken In less than 3%

Increments between 100% and 120%

CLTP.

Turbine First Validate the scram Stage bypass function for Pressure the TSV Closure and TCV Fast Closure -

Low Control Oil Pressure scram functions. Collect X

X X

X Data on the Turbine First Stage Pressure to Rated Thermal Power relationship over a band of 15%.

Pressure Pressure regulator Regulator control system Testing:

response to a Bypass Valves pressure setpoint x

change. This test to be performed on both pressure regulators.

BVY 03-98 / Attachment 3 / Page 10

- :;::CPPU POWER ASCENSION TE PAN :

TEST TEST D C

T

--PRIOR TO TEST

DESCRITON';;-

-START PERCENT POWER CLTP (Al oWance 0% 3%)

-(Alownce 40% -

1 ~START P--.;,I:-.-:--...... -,

,,5 10 15 20' 25 -30 35

-40 '45' 50 55 60 65 70 75

'8 0

85 90 95 100 105 110 115 1

Pressure Pressure regulator Regulator control system Testing: TCVs response to a pressure setpoint X

X X

X X

X change. This test to be performed on both pressure regulators.

Pressure Collect Data from Regulator Generator minimum Testing-load to maximum Incremental CPPU power In <3%

XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX XX Data Increments between Collection.

15% and 120%

CLTP.

Reactor Water Test performed In Level Setpolnt single element and In Changes three element control.

X X

X X

X Manual Manually raise and Feedwater lower feedwater flow Flow Step using manual control.

X X

X X

X Changes Maximum Pressure, flow and Feedwater controller data Runout Data collected on Collection feedwater system performance.

X X

X X

X Measured data is compared against expected values.

BVY 03-98 / Attachment 3 / Page II Comparison Of Iniial StTsting And 'Planed CPPU Te.stin

Test Descrptio AOO, Te.

Test Derived From VYNPS USR eto:Ok

.,Test -'. 'Derived FromUFSAR Section Org. S/U (System Planned For Evaluation/Justification'For Not Performing'

'Number' 1...

D-.:::..

3.5 :::'....}:,

... TestPhase: Tralnsent)

CPPU,:.:

.:(

O '

s3.1 d..

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i tn t.

(

o Chal eng an

( S u b - e ctio n s 3 5.2, 1 3.3 A n d 3.5.4 )

estA s E t

( e / o_

STP 1 Chemical and Radiochemical: Chemical and radiochemical tests were conducted to establish water conditions prior to initial operations and to maintain these throughout the test program.

Chemical and radiochemical checks were made at primary coolant, off-gas exhaust, waste, and auxiliary system sample locations. Base, or background, radioactivity levels were determined at this time for use in fuel assembly failure detection and long-range activity buildups studies.

Chemical and Radiochemical checks were made during heatup.

Chemical and Radiochemical Tests were continued.

Steam Separator-Dryer measurements of carryover and carryunder were made as a function of reactor water level and power level.

Open Vessel Testing Initial Heatup Power Tests Power Tests None Yes Notes: 1,2, And 3 None None None

BVY 03-98 / Attachment 3 / Page 12 s

,r s s

Comparison Of Initial Startu Testing And Planned CP PUTesting

! --i- --Test Description W.

--. :> AOO :--

a;

-est B

Test.-

Derived FromVYNPS UFSAR Section Org. S/U (Pln F

Ealuation/Jutstification For Not Performing Nu r

1 3.5 Test Phase Challege Plant..

Tes (Sub-SeconN 135.2'1353An 13.5.4)

Tes 'tAspect '

STP 2 Radiation Measurements were made for about a Open Vessel None Yes Note 4 year prior to nuclear operation to establish base Testing environmental monitoring levels.

Radiation Measurements were made periodically Initial None during heatup.

Heatup Radiation Measurements of limited extent were Power Tests None made at 25% of rated power and thorough surveys were made at 50%, 75%, and 100%

power.

STP 3 Fuel Loading: Fuel Loading was performed Open Vessel None No The purpose of this test is to load fuel safely and according to detailed, step-by-step written Testing efficiently to the full core size. Current Technical procedures. Control curtains were in place Specifications and approved plant procedures and the test and operational neutron sources effectively govern the safe and efficient loading of were installed as required. Loading fuel. No new fuel types are introduced for CPPU.

proceeded to the full size core.

CPPU has no affect on this test, therefore this test is not required.

BVY 03-98 / Attachment 3 / Page 13 Comparison Of Initial Startup Testing And Planned

'CPPU Testing.'

Test Description

-AO

Test '

Test':-

DerivedFromVYNPSUFSAR Section Org. su.

(System edFr Evaluation/Justification For. Not Performing-Number

13.5 Test Phase U.

Sant; PPU Test (Sub-Sections 13.5.2,13.5.3 A

):

Te Aspect.

(Yeso)

STP 4 Shutdown Margin: Shutdown Margin was demonstrated periodically during fuel loading that the reactor was subcritical by more than a specified amount with the single highest worth control rod withdrawn. The magnitude of the margin was chosen with consideration for expected reactivity changes during the first operating cycle and for the accuracy of measurement. The test had three parts: (A) The analytical determination of the control rod having the greatest reactivity worth, (B) The calibration of an adjacent control rod, determined analytically, and (C) The demonstration of subcriticality with the highest worth rod fully withdrawn and the second at the position needed to insert the required margin. This demonstration was made for the fully loaded core and for selected smaller core loadings.

Open Vessel Testing None No The Shutdown Margin Requirement is not changed by CPPU. Shutdown Margin Testing is performed for each reload in accordance with approved plant procedures and Technical Specifications. This testing will be performed, as required by Technical Specifications, during the refueling outage that the CPPU core is loaded. Shutdown Margin Testing specifically for CPPU is not Required.

L I

I

BVY 03-98 / Attachment 3 / Page 14 Comparison Of Initial StartupTeting An d P laned CPPU Testing:

Test Description '

AOO

Test, Test' Derived From YNPS UFSAR Section.

Org. SU '

(System;.

Planned For Evaluatio Justification For Not Perorming Nber

.... ';.E.-.

13.5 Test Phase C

e P

Test Frinsienf)-

(Sub-Sections 13.5.2, 3.3 And 13.5.4)

'e Al A pec t' (YesN o)

STP 5 Control rod drive tests: Control rod drive Open Vessel None No CRD scram time testing will be performed in system tests were performed on all drives Testing accordance with Technical Specifications section prior to fuel loading to assure proper 4.3.C and approved plant procedures. The operability and to measure and adjust performance of the CRD hydraulic system is operating speeds. Functional testing of each independent of power level. There is no effect on drive was performed with dummy fuel just the performance of the CRD system with no prior to fuel loading. Functional testing was increase in reactor pressure. This is confirmed in again performed following the fuel loading section 2.5 of Attachment 4 to this LAR.

in each cell. Drive pipe friction and scram times were determined for all drives at zero reactor pressure.

Control rod drive system tests were made by Initial None measuring scram times on a selected number Heatup of drives at two intermediate pressures, scram times, and drive line friction tests on a representative set of drives at rated reactor pressure, and in-out driving times of selected rods during heatup.

BVY 03-98 / Attachment 3 / Page 15 Comparison Of nitial Startu Testi And Plann d CPPU Te sng-Test Description AOO,-

Test Test Derived From VYNPS UFSAR Section Org. S/U-(System Panned For -EvaluationJustfication For Not Performig Number i--13.

TestPse haclenge /Plant P;,,.

. ~~~~~~~~.

T ansent

CPPU, 1. ',

. Test

-(SubSections l3.5.2,13.5.3 And 13.5.4);

7 At'Aspe (Yeso)

STP 6 Control rod sequence: Control rod sequences Open Vessel None No This is an initial startup test requirement to achieve were evaluated to verify that the stated Testing initial criticality in a safe and efficient manner for criteria of safety, simplicity, and operating each of the two withdrawal sequences. Operation at requirements are met during routine cold CPPU increases the upper end of the power startup. The reactor was brought critical by operating domain. These changes in the higher end withdrawing control rods in a specified do not significantly or directly affect the manner of sequence and reactivity addition rates are operating or response of the reactor in the measured near critical. The pre-selected startup/low power range. Therefore, this test is not sequence was modified if necessary to meet required. Plant startups will be performed using criteria. A few nonstandard arrays were approved plant procedures.

utilized to check out the operation of the rod worth minimizer.

Control rod sequence to be used during the tial None heatup was checked periodically for Heatup satisfactory performance.

STP 7 Calibration of rods: Calibrations of rods Power Tests None No Operation at CPPU increases the upper end of the were performed to obtain reference power operating domain. These changes in the relationships between control rod motion higher end do not significantly or directly affect the and reactor power and steam flow in the manner of operating or response of the reactor in the specified control rod sequence.

startup/low power range. Therefore, this test is not required.

STP 8 Rod pattern exchange: Rod pattern Power Tests None No The methodology and approach taken to perform exchanges were demonstrated from one rod pattern exchange is not significantly affected by specified control rod sequence to the other at operation at CPPU conditions. Rod pattern the highest practical reactor power.

exchanges will be performed in accordance with approved plant procedures and within fuel warranty requirements. This test is not required.

BVY 03-98 / Attachment 3 / Page 16 Comparison Of Initial Statu T Anid Pland T

-PU

-Testing

.:Test Description AOO Test-.

--;Test

.D

.erived From VYNPS UFSAR Section Org SU

.(Synetem;

Plannd Fo; EvalutionJustifcato For ot Perform g Number 13.5 Test Cha enge/T Phase Transieii)

CP (Sub-Sect ns 13.5.2, 1353 A d 13.5.4)

.Teit'As...

(_

STP 9 SRM Performance: Source Range Monitor Open Vessel None No SRM instrumentation is not modified for CPPU and (SRM) Performance. Adequate performance Testing operational neutron sources will not be required.

of the SRMs was established from data Operation at CPPU increases the upper end of the taken with the operational neutron sources in power operating domain. These changes in the place. During initial reactor operations, the higher end do not significantly or directly affect the SRM subsystem was calibrated and its manner of operating or response of the reactor in the performance was compared with criteria on startup/low power range. Therefore, this test is not noise, signal-to-noise ratio, and response to required.

change in core reactivity.

SRM performance was determined by checking for proper overlap with the IRM Initial None subsystem.

Heatup

BVY 03-98 / Attachment 3 / Page 17

.Comparison Of Initial Startup Testing AdI nned CPPU Testing' Test Descripfion:,.-

AOO.

Test

.:. '.Test ' Derived From PS VFSAR Sec on.

Org.' S/U,

' ' ' Planned Fr EvaluatioJustification For Not Perf g

Number..

13.5 Tst Phase T,

a ewg CPPU-(Sub-Sections 13.5.2, 13.53 And 13.5A).

Test Aspect'

.(Yes/o).:.:.

STP 10 Intermediate Range Monitor (IRM)

Open Vessel None Yes Note 5 Calibration: The IRMs were initially Testing calibrated to give useful readings and to supply protection for this phase of the test program. This initial calibration was made by comparing to SRM readings in the overlap region.

IRM Calibration. The IRM subsystem was initial None recalibrated during heatup by making the Heatup IRM readings proportional to a known heat input to the reactor coolant from a non-nuclear heat source, such as the main recirculation pumps. The proportionality was determined by measuring the reactor coolant temperature rise produced by pump heating and by nuclear heating.

IRM Calibration. The final calibration of None the IRM system was made in the APRM-Power Tests IRM power overlap region subsequent to the calibration of the APRM system.

STP 11 LPRM calibrations, which included use of Power Tests None No LPRM calibration is performed at a frequency the traversing in-core probe (TIP) specified in the Technical Specifications using subsystem, were made at 50%, 75%, and approved plant procedures. The method and 100% of rated power. Each local power approach used to perform LPRM calibration is not range monitor was calibrated to read in affected by CPPU. This test is not required.

terms of local fuel rod surface heat flux.

BVY 03-98 / Attachment 3 / Page 18 Comparis on Of Initial Startup Testing n Planned CPPU Test.

Tes Derived,

-.:-iTest De' ription -

AOOM ;

'Tet' Test,: DerivedFrom'YNS FSAR Sction

Org.S S F(stem Planed Evalu tionJstfica onFor Not Peror ng Nmer 1-

f3.5

,>. ;Test Phse :

haegeflln:

Challenge Test 1 : - : '

s3.5.2, 13.5.3 And 13.5.4)

T Ansiet.

Yeo)

- C U..:::

T.n.:

S u b-S ec io sest Aspect'(./

o STP 12 APRM Calibration: APRM calibrations Power Tests None Yes APRMs will be re-calibrated to read 100% at CPPU were performed after making significant power level changes. Reactor heat balances formed the bases of the calibrations of these average power range monitors.

STP 13 Process Computer. As station process Prior To None No The plant process computer is maintained by variable signals became available to the Startup And approved plant procedures. Startup testingfor this computer, verification was made of these During system is not required.

signals and of the computerized systems Open Vessel performance calculations.

Testing Process computer functions were verified as Power Tests None sensed variables came into range during the ascension to and at rated power.

STP 14 RCIC: RCIC system was actuated when the Power Tests See Evaluation/

No RCIC Automatic Start From Cold Conditions reactor was shut down, but hot and Justification (Performed At = 25% Power) - CLTR Section 3.9 pressurized, to demonstrate full capacity For Not indicates that there is no effect on the RCIC System operation of the steam turbine driven pump.

Performing for a Constant Pressure Power Uprate. This is Test confirmed in Section 3.9 of Att. 4 to this LAR.

RCIC System testing, including automatic starts from cold conditions, is governed by Technical Specifications and approved plant procedures. As the CPPU does not have an effect on RCIC System, current surveillance testing remains valid for CPPU operations. Therefore, this testing is not required.

BVY 03-98 / Attachment 3 / Page 19 Coparison Of Initial Strtupesting AndPlanned CPPU Testing:- -

Test Description;...

AOO Test

-Test Derived From VYNPS UFSAR Secton Org.S/U (Sysem.

PlannedFo Ev Ion/Justific tio o N ot Pe for in Numer-:

1.:15 Test Phase Challenge /Pa nt::

PP.

Test -- -

system, including the steam turbine driven For Not the IIPCI System for a Constant Pressure Power pump.

Performing Uprate. This is confirmed in Section 4.2 of Att. 4 to Test this LAR. HPCI System testing, including automatic starts from cold conditions, is governed by Technical Specifications and approved plant procedures. As the CPPU does not have an effect on the HPCI System, current surveillance testing remains valid for CPPU operations. Therefore, this

_____________s___________

testing is not required.

STP 16 Reactor Vessel Temperature: Reactor vessel Power Tests None No This test obtains RPV temperatures during rapid temperatures were monitored during heatup heatup and cooldown to confirm thermal analysis and cooldown to check for proper operation models. CPPU does not affect the RPV and determine that specified temperature temperatures during rapid heatup or cooldown.

differences are not excessive.

Since thermal analysis models were confirmed during the initial startup test subsequent testing is

~~~~~~~~~~~~~~~~not required.

SIT 17 System Expansion: System expansion Initial None No Since CPPU does not include a reactor vessel checks were made during heatup to verifyT Heatup pressure increase, nor the corresponding primary freedom of motion of major equipment and coolant temperature increase, thermal expansion of piping.

drywvell piping is not affected by CPPU. This test is not required.

System expansion tests were continued on a Power Tests None limited basis as reactor power was increased.

heatupand____

_toconfirmthermalanalysis STP 18 Power Distribution: Axial power Per r Tests See Evaluation!

No There are no changes to the tip system as a result of distribution measurements were made with Justification the CPPU. This test is not required.

the tip system after significant changes in For Not power, control rod pattern, or flow rate. The Performing tip data were used for core performance Test evaluations and LPRM calibrations.

Power TestsNone

BVY 03-98 / Attachment 3 / Page 20 Com;parison Of Initil Staru I eti'gAnd Plinned CPPU Testing-Test Description AOO :-

Test,

-Test..

Derived From VYNPS UFSAR Section Org. SU'

'(System* Planned F Evaltion/Jstification For Not Performi N

ber..-

13 ;::, '

.5 st s

/ln

.CPP Niiinlier T~~~~~~~~~~~~~~~Trniet (Sub-Seti on135. 13.53 And 13.5.4),,

Tet pe

(e_--:-.-:

STP 19 Core Performance Evaluation: Core Initial None Yes This test will be performed for CPPU performance evaluations were made near or Heatup at rated temperature and pressure. This includes a reactor heat balance at rated temperature.

Core performance evaluations were made Power Tests None periodically to demonstrate that the core was operating within allowable limits on maximum local surface heat flux and minimum critical heat flux ratio. This test included reactor heat balance determinations.

STP 21 Flux Response To Rods: Flux response to Power Tests None No This initial plant startup test was performed at 17%

control rod movements were determined in and 52% CLTP power. Operation at CPPU both equilibrium and transient conditions.

increases the upper end of the power operating Steady-state noise was measured if possible.

domain. These changes in the higher end do not Power-void loop stability was verified from significantly or directly affect the manner of this data.

operating or response of the reactor at these lower power levels. Therefore, this test is not required.

STP 22 Pressure Regulator Reactor pressure Initial See Evaluation Yes Setpoint Step Change and Simulated Failure Testing control was instituted using the main turbine Heatup (Testing From 5% To 100% Power) - CLTR pressure regulator.

Section 5.2 indicates a CPPU effect on the reactor pressure control system due to the increased power level and steam flow. Section 10.4 of Att. 4 to this LAR requires testing to demonstrate acceptable CPPU performance.

Pressure regulator tests were made to Power Tests See Evaluation determine the response of the reactor and the turbine governor system. Regulator settings were optimized using data from this test.

BVY 03-98 / Attachment 3 / Page 21

'Comparisonf Of Initial Star-tup:Testing And Pla'nned:CPPU Testinig' Ta~tDescrption3/4 A O

Test, Test-Derive'd Firm VYNPS UFSAR Secin

r.

/f(ytelnn di or EvaluationlJustification ForNot Per-forming Number 13.5 Test Phase

~~~~~~~~~~Challenge;/Plant CPUTs (Sub-Sections 13.5.2,1353And 13.5.4).

Test Apc (YesNo STP 23 Feedwater System: Feedwater pump trip Power Tests See Evaluation Yes Setpoint And Flow Change Testing (Testing From tests were made to demonstrate reactor water 15% To 100% Power) - CLTR Section 5.2 indicates level and plant response to loss of part of a CPPU effect on the Feedwater Control System due feedwater supply. Reactor water level to the increased power level and feedwater flow.

changes were made to determine reactor Section 10.4 of Att. 4 to this LAR requires testing to response and to optimize level controller demonstrate acceptable CPPU performance.

settings.

Feedwater Pump Trip (Test Initial Conditions Range Between 75% And 100% Power) - CPPU LTR Section 9.1.3 indicates that loss of one feedwater pump has been included in CPPU transient analysis only for operational considerations (I.E., scram avoidance) and is not significantly affected by the CPPU. Therefore, this testing is not required. (Note 6 )

STP 24 Bypass Valves: Bypass valve measurements Power Tests See Evaluation Yes Bypass valve functional test (final test condition were performed by opening a turbine bypass range between 85% and 100% power) - Power level valve and recording the resulting reactor at which valve testing can be performed without transients. Final adjustments to the pressure causing a scram / isolation may increase due to regulators were made.

CPPU instrument-related changes (i.e., APRM re-calibration, MSL high steam flow isolation) as indicated in CLTR section 5.0. However, test power level is only a plant capacity consideration.

Valve testing can continue to be performed at pre-CPPU power level (MWTh). Therefore, this testing

___________is not required.

BVY 03-98 / Attachment 3 / Page 22 Comparison Of Initial StartupTesg And'Planned CPPU Testing

,.Test Description AOO Te'

,.Test,.

Derived From VNPS UFSARSection :' Org.

^ ;

(System Planned For Evaluation/iJustification For Not Prformig

'Numberg ' :. ';' '.:: - '. :. ' :'.

13.5 :..':-: ',::;',

Test Phase,,c a enge

, 'Challenge

/PlPPU Te s (Sub-Sectorns 1

Test As,'

(yeo):

Test STP 25 Main Steam Isolation Valves: Main Steam Initial See Evaluation/

No Single MSIV closure testing - Power level at which Isolation Valve functional tests were made at heatup Justification a single MSIV can be closed without a scram may rated pressure.

For Not increase due to CPPU instrument-related changes as Performing indicated in CLTR section 5.0. However, this Test power level is a plant capacity consideration. Single Main Steam Isolation Valve functional and Power tests MSIV closure can continue to be performed'at pre-operational tests were made as reactor power CPPU power level (MWTh). Therefore, this testing was increased.

is not required.

MSIV closure (test at 100% power) - This large transient test and others (i.e., generator load rejection, turbine trip) are evaluated for exemption from CPPU test program in Attachment 7 of this LAR.

STP 26 Relief Valves: This initial startup test was Power Tests See Evaluation/ No Technical Specifications and approved plant performed at -19.5% CLTP to verify proper Justification procedures govern the testing of the relief valves operation of the safety relief valves, verify For Not including manual opening of each relief valve once proper sealing, and determine their capacity.

Performing per cycle. Section 3.1 of Att. 4 to this LAR Note: No test description for Relief Valve Test documents the acceptable evaluation of the testing was identified in VYNPS UFSAR.

overpressure protection and no effect on valve Description is from GE Document No.

functionality by the CPPU. This test is not required.

22A2217 Startup Test Specification for Vermont Yankee.

STP 27 Turbine Trip: Turbine trip tests were Power Tests See Evaluation/

No This Large Transient Test and others (I.E.,

performed to determine speed and reactor Justification Generator Load Reject, MSIV Closure) are response.

For Not evaluated for exemption from CPPU test program in Performing of this LAR.

Test

BVY 03-98 Attachment 3 / Page 23 Coparison Of Initial Startup Testing And Pla'nne-dCPPUTJ.'estig, Test Deription'O

~

Ts Test Derived Fromi VYNPS UFSAR Sectiont Org. Sju

'(Sytm Planned For 'Evaluation/Justification For Not Perfog orniing~~~~~~

e o

Number

~~~~13.5 Test Phase hfegeIin Test Sub-Sections'13.5.2,13.5.3 And 13.54 Tt set (Yes/No)

STP 28 Generator Trip: Generator trip tests were Power Tests See Evaluation/

No This Large Transient Test And Others (I.E., Turbine performed to determine speed and reactor Justification Trip, MSIV Closure) are evaluated for exemption response.

For Not from CPPU test program in Attachment 7 of this Performing LAR.

Test STP 29 Recirculation Flow Control: Flow Control Power Tests See Evaluation/

No Flow Change Testing - CUTR Section 3.6 indicates capabilities were determined at specified Justification a CPPU effect that increased voids in the core power levels.

For Not during normal uprate power operations requires a Performing slight increase in recirculation drive flow to achieve Test the same core flow. Section 3.6 of Att. 4 to this LAR documents that the plant-specific system evaluation of the reactor recirculation system performance at CPPU power determines that adequate core flow can be maintained without requiring any changes to the recirculation system and only a small increase in pump speed for the same core flow, the response to flow changes will be similar to that of original startup testing.

_ _ _ _ _T h ere fo re,_

T ereforeethisgtstin g ostnot qr uuire d

BVY 03-98 / Attachment 3 / Page 24 Com'parison Of InitialStartp Tes ng And n

CPP Testing' TestDescripion..

J AOO est

Test Derived'From VYNPS UFSAR Section Org. SU (System

'Planned For EvaluationlJtstification For NotPerforming

Number. -

13.5 T'est Phase Challenge I lant e s (Subection 135.2,13. 3 ndl3.

Tet Ast (s)

STP 30 Recirculation System: Recirculation pump trips and their effects on the jet pumps and the reactor were tested periodically during power increase.

Power Tests See Evaluation/

Justification For Not Performing Test No One Pump Trip (Final Test At 100% Power) -

CLTR Section 3.6 indicates a CPPU effect that increased voids in the core during normal uprate power operations requires a slight increase in recirculation drive flow to achieve the same core flow. Section 3.6 of Att. 4 to this LAR documents that the plant-specific system evaluation of the reactor recirculation system performance at CPPU power determines that adequate core flow can be maintained without requiring any changes to the recirculation system/pumps and only a small increase in their speed for the same core flow. The response to a one pump trip will be similar to that of original startup testing. Therefore, this testing is not required. Two Pump Trip (Final Test At 100%

Power) - Section 3.6 of Att. 4 to this LAR indicates a CPPU effect that increased voids in the core during normal uprate power operations requires a slight increase in recirculation drive flow to achieve the same core flow. Section 3.6 of Att. 4 to this LAR documents that the plant-specific system evaluation of the reactor recirculation system performance at CPPU power determines that adequate core flow can be maintained without requiring any changes to the recirculation system/pumps and only a small increase in their speed for the same core flow. The response to a trip of both pumps will be similar to that of original startup testing. Therefore, this testing is not required.

BVY 03-98 / Attachment 3 / Page 25 Compariso Of Initial Startup Testing n Plne PPU Testig' Test Description-,

AO.__

Tes

Test, Derived From VYNPS UFSAR'Section

'Org. SU (System

-P' ed For, Evaluation/Justification For Not Perforng Number':

13.5:

.Test Phase C~.haIenge

,,anut Transfent)

PUTs (Sub-Sections 13.5.2, 13.53 And 13.5.4)

TestApt; e;

)

STP 31 Loss of Turbine-Generator and Off-Site Power Tests See Evaluation/

No Loss Of Turbine Generator and Off-Site Power Power: Auxiliary power loss tests were Justification Initial Test Condition was -20% power - CLTR made to verify acceptable performance of For Not Section 6.1 indicates that, under emergency the reactor and the electric equipment and Performing operations/distribution conditions (emergency diesel auxiliary systems during the resulting Test generators), the AC power supply and distribution transients.

components are considered adequate and an evaluation assures an adequate AC power supply to safety related systems. Section 6.2 of Att. 4 to this LAR documents the acceptable evaluation of the AC power system. Technical Specifications and approved plant procedures govern the testing of the safety related AC distribution system, including loss of off-site power tests. Operation at CPPU increases the upper end of the power operating domain and does not significantly or directly affect the manner of operating or response of the plant in the startup/low power range. Therefore, this test is not required.

STP 32 Recirculation M-G Set Speed Control: Flow Power Tests None No This test determines the as built characteristics of control capabilities were determined at the recirculation control system including the drive specified power levels.

motor, fluid coupler, generator, drive pump and jet pumps. The CLTP recirculation pump speed range remains unchanged for operation at CPPU conditions. With the exception of adding a runback signal when a feedwater pump trips (Note 7) there are no modifications required to these components for CPPU. This test is not required since the recirculation system controls are unaffected by CPPU. (Note 8)

BVY 03-98 / Attachment 3 / Page 26 Comparison Of Initial Startp TestingA-d Plnn'ed CPPU T'sting Test-Descripti AOO:,

Test' Ts N Derived F VYNPS '.SAR Section:

Org'.".

'e Plnnd F 3,: Eva Iati'nustification For Not Pf orming

-Ndinbek, 13.5Test Phase ClegeT Ins nt P'

Test (Sub-Secon 1.5.2 135 And 13.5.4) '

-srect (Yeo.:...

STP X-5 (90)

Vibration Testing: Vibration measurements Open Vessel None No This test obtains vibration measurements on various at cold flow conditions were performed as Test reactor pressure vessel internals to demonstrate the necessary to determine the vibrational mechanical integrity of the system under conditions characteristics of reactor vessel internals of of flow induced vibration, and to check the validity Vermont Yankee design. The results of of the analytical vibration model. Analysis of the extensive vibration measurements made at reactor vessel internals at CPPU power level was other BWR installations were considered in performed to ensure that the design continues to selecting the components to be tested if it comply with the existing structural requirements.

should be required.

(Note 9)

Vibration measurements were performed as Power Tests None necessary.

BVY 03-98 / Attachment 3 / Page 27 Notes to Table

1. For CPPU Testing, added demonstration of proper steam separator-dryer operation.

2. Startup test included objective to determine that the sampling equipment, procedures and analytical techniques are adequate to supply the data required to demonstrate that the coolant chemistry meets water quality specifications and process requirements. This objective is not applicable to CPPU and is not required.

3. Startup test included objective to evaluate the performance of the fuel, operation of the demineralizers and filters, condenser integrity, operation of the off gas system and calibration of certain process instruments. The current Vermont Yankee chemistry and plant performance monitoring programs gather information on plant equipment and system performance. This information is evaluated in order to maintain equipment, system and plant performance within process requirements, chemistry/radiochemistry specifications -and guidelines and fuel warrantee.

This testing is not required for CPPU implementation.

4. Startup test included objective to determine the background radiation levels in the plant environs prior to operation for base data on activity buildup. This initial startup requirement is not applicable to CPPU and is not required.

5. The IRM overlap with the SRMs is not affected by CPPU. The APRMs will be re-referenced to read 100% at CPPU conditions, therefore, the IRM performance test will be performed to reestablish the IRM to APRM overlap.

6. Feedwater System startup testing included a feedwater pump trip test. For this test one of two operating feedwater pumps was tripped and the standby feedwater pump was allowed to automatically start. At CPPU conditions all three feedwater pumps will be required; there will be no standby pump available. This test is not required for CPPU.

7. Testing associated with the recirculation pump runback on reactor feed pump trip is discussed above in the modification testing section of this attachment.

8. The recirculation system will have to overcome a slight increase in two-phase flow resistance due to an increase in the core average void fraction. The system will accommodate the expected insignificant increase at CPPU condition when operating at maximum core flow.

9. Results of this analysis are in section 3.4.2 of NEDC-33090P (Attachment 4 of the LAR).

Docket No. 50-271 BVY 03-98 Attachment Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 263 Supplement No. 3 Extended Power Uprate - Updated Information Justification for Exception to Large Transient Testing

BVY 03-98 / Attachment 7 / Page 1 JUSTIFICATION FOR EXCEPTION TO LARGE TRANSIENT TESTING

Background

The basis for the Constant Pressure Power Uprate (CPPU) request was prepared following the guidelines contained in the NRC approved, General Electric (GE) Company Licensing Topical Report for Constant Pressure Power Uprate (CLTR) Safety Analysis: NEDC-33004P-A Rev. 4, July 2003.

The NRC staff did not accept GEs proposal for the generic elimination of large transient testing (i.e., Main Steam Isolation Valve (MSIV) closure and turbine generator load rejection) presented in NEDC-33004P Rev. 3. Therefore, on a plant specific basis, Vermont Yankee Nuclear Power Station (VYNPS) is taking exception to performing the large transient tests; MSIV closure, turbine trip, and generator load rejection.

The CPPU methodology, maintaining a constant pressure, simplifies the analyses and plant changes required to achieve uprated conditions.

Although no plants have implemented an Extended Power Uprate (EPU) using the CLTR, thirteen plants have implemented EPUs without increasing reactor pressure.

Hatch Units I and 2 (105% to 113% of Original Licensed Thermal Power (OLTP))

Monticello (106% OLTP)

Muehleberg (i.e., KKM) (105% to 116% OLTP)

Leibstadt (i.e., KKL) (105% to 117% OLTP)

Duane Arnold (105% to 120% OLTP)

Brunswick Units I and 2 (105% to 120% OLTP)

Quad Cities Units I and 2 (100% to 117% OLTP)

Dresden Units 2 and 3 (100% to 117% OLTP)

Clinton (100% to 120%)

Data collected from testing responses to unplanned transients for Hatch Units I and 2 and KKL plants has shown that plant response has consistently been within expected parameters.

Entergy believes that additional MSIV closure, turbine trip, and generator load rejection tests are not necessary. If performed, these tests would not confirm any new or significant aspect of performance that is not routinely demonstrated by component level testing.

This is further supported by industry experience which has demonstrated plant performance, as predicted, under EPU conditions. VYNPS has experienced generator load rejections from 100% current licensed thermal power (see VYNPS Licensee Event Reports (LER)91-005, 91-009, and 91-014). No significant anomalies were seen in the plant's response to these events. Further testing is not necessary to demonstrate safe operation of the plant at CPPU conditions. A Scram from high power level results in an unnecessary and undesirable transient cycle on the primary system. In addition, the risk posed by intentionally initiating a MSIV closure transient, a turbine trip, or a generator load rejection, although small, should not be incurred unnecessarily.

VYNPS Response to Unplanned Transients:

VYNPS experienced an unplanned Generator Load Rejection from 100% power on 04/23/91.

The event included a loss of off site power.

A reactor scram occurred as a result of a turbine/generator trip on generator load rejection due to the receipt of a 345 KV breaker failure signal. This was reported to the NRC in LER 91-009, dated 05/23/91. No significant anomalies

BVY 03-98 / Attachment 7/ Page 2 were seen in the plant's response to this event.

VYNPS also experienced the following unplanned generator load rejection events:

On 3113/91 with reactor power at 100% a reactor scram occurred as a result of turbine/generator trip on generator load rejection due to a 345KV Switchyard Tie Line Differential Fault. This event was reported to the NRC in LER 91-005, dated 4/12/91.

On 6/15/91 during normal operation with reactor power at 100% a reactor scram occurred due to a Turbine Control Valve Fast Closure on Generator Load Rejection resulting from a loss of the 345KV North Switchyard bus. This event was reported to the NRC in LER 91-014, dated 7/15/91.

No significant anomalies were seen in the plant's response to these events. Transient experience at high powers and for a wide range of power levels at operating BWR plants has shown a close correlation of the plant transient data to the predicated response.

Based on the similarity of plants, past transient testing, past analyses, and the evaluation of test results, the effects of the CPPU RTP level can be analytically determined on a plant specific basis.

The transient analysis performed for the VYNPS CPPU demonstrates that all safety criteria are met and that this uprate does not cause any previous non-limiting events to become limiting. No safety related systems were significantly modified for the CPPU, however some instrument setpoints were changed. The instrument setpoints that were changed do not contribute to the response to large transient events. No physical modification or setpoint changes were made to the SRVs. No new systems or features were installed for mitigation of rapid pressurization anticipated operational occurrences for this CPPU. A Scram from high power level results in an unnecessary and undesirable transient cycle on the primary system.

Therefore, additional transient testing involving scram from high power levels is not justifiable. Should any future large transients occur, VYNPS procedures require verification that the actual plant response is in accordance with the predicted response.

Existing plant event data recorders are capable of acquiring the necessary data to confirm the actual versus expected response.

Further, the important nuclear characteristics required for transient analysis are confirmed by the steady state physics testing. Transient mitigation capability is demonstrated by other equipment surveillance tests required by the Technical Specifications. In addition, the limiting transient analyses are included as part of the reload licensing analysis.

MSIV Closure Event Closure of all MSIVs is an Abnormal Operational Transient as described in Chapter 14 of the VYNPS Updated Final Safety Analysis Report (UFSAR). The transient produced by the fast closure (3.0 seconds) of all main steam line isolation valves represents the most severe abnormal operational transient resulting in a nuclear system pressure rise when direct scrams are ignored.

The Code overpressure protection analysis assumes the failure of the direct isolation valve position scram. The MSIV closure transient, assuming the backup flux scram verses the valve position scram, is more significant. This case has been re-evaluated for CPPU with acceptable results.

The CLTR states that: "The same performance criteria will be used as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program."

The original MSIV closure test allowed the scram to be initiated by the MSIV position switches.

As such, if the original MSIV closure test were re-performed, the results would be much less significant than the MSIV closure analysis performed by GE for CPPU.

BVY 03-98 / Attachment 7/ Page 3 The original MSIV closure test was intended to demonstrate the following:

1. Determine reactor transient behavior during and following simultaneous fill closure of all MSIVs.

Criteria:

a) Reactor pressure shall be maintained below 1230 psig b) Maximum reactor pressure should be 35 psi below the frst safety valve setpoint.

(This is margin for safety valve weeping).

2. Functionally check the MSIVs for proper operation and determine MSIVclosure time.

Criteria:

a) Closure time betiveen 3 and 5 seconds.

Item 1: Reactor Transient Behavior For this event, the closure of the MSIVs cause a vessel pressure increase and an increase in reactivity. The negative reactivity of the scram from MSIV position switches should offset the positive reactivity of the pressure increase such that there is a minimal increase in heat flux.

Therefore, the thermal performance during the proposed MSIV closure test is much less limiting than any of the transients routinely re-evaluated.

CPPU will have minimal impact on the components important to achieving the desired thermal performance. Reactor Protection system (RPS) logic is unaffected and with no steam dome pressure increase, overall control rod insertion times will not be significantly affected. MSIV closure speed is controlled by adjustments to the actuator and is considered very reliable as indicated below.

Reactor Pressure Due to the minimal nature of the flux transient, the expected reactor pressure rise, Item 1 above, is largely dependent on SRV setpoint performance. At VYNPS all four SRVs are replaced with re-furbished and pre-tested valves each outage. After the outage, the removed valves are sent out for testing and recalibration for installation in the following outage. Over the past ten years there have been twenty five SRV tests performed. In those twenty five tests only one test found the as-found setting outside the Technical Specification (TS) current allowable tolerance of +/-3%. This valve was found to deviate by 3.4% of its nominal lift setpoint. Note that this is bounded by the VYNPS design analysis for peak vessel pressure which assumes one of the four SRVs does not open at all (one SRV out of service). Given the historical performance of the VYNPS SRVs along with the design margins, performance of an actual MSIV closure test would provide little benefit for demonstrating vessel overpressure protection that is not already accomplished by the component level testing that is routinely performed, in accordance with the VYNPS TSs.

Because rated vessel steam dome pressure is not being increased and SRV setpoints are not being changed, there is no increase in the probability of leakage after a SRV lift. Since SRV leakage performance is considered acceptable at the current conditions, which match CPPU conditions with respect to steam dome pressure and SRV setpoints, SRV leakage performance should continue to be acceptable at CPPU conditions.

An MSIV closure test would provide no significant additional confirmation of Item I performance criteria than the routine component testing performed every cycle, in accordance with the VYNPS TSs.

BVY 03-98 / Attachment 7/ Page 4 Item 2: MSIV Closure Time Since steam flow assists MSIV closure, the focus of Item 2 was to verify that the steam flow from the reactor was not shut off faster than assumed (i.e., 3 seconds).

During maintenance and surveillance, MSIV actuators are evaluated and adjusted as necessary to control closure speed, and VYNPS test performance has been good. To account for minor variations in stroke times, the calibration test procedure for MSIV closure (OP 5303) requires an as left fast closure time of 4.0 +0.2 seconds. The MSIVs were evaluated for CPPU. The evaluation included MSIV closure time and determined that the MSIVs are acceptable for CPPU operation. Industry experience, including VYNPS, has shown that there are no significant generic problems with actuator design. Confidence is very high that steam line closure would not be less than assumed by the analysis.

Other Plant Systems and Components Response The MSIV limit switches that provide the scram signal are highly reliable devices that are suitable for all aspects of this application including environmental requirements. There is no direct effect by any CPPU changes on these switches. There may be an indirect impact caused by slightly higher ambient temperatures, but the increased temperatures will still be below the qualification temperature. These switches are expected to be equally reliable before and after CPPU.

The Reactor Protection System (RPS) and Control Rod Drive (CRD) components that convert the scram signals into CRD motion are not directly affected by any CPPU changes. Minor changes in pressure drops across vessel components may result in very slight changes in control blade insertion rates. These changes have been evaluated and determined to be insignificant. The ability to meet the scram performance requirement is not affected by CPPU.

Technical Specification (TS) requirements for these components will continue to be met.

CPPU Modifications Feedwater System operation will require operation of all three feed pumps at CPPU conditions (unlike CLTP conditions). Operation of the additional Reactor Feed Pump (RFP) will not affect plant response to an MSIV closure transient. All feedwater pumps receive a trip signal prior to level reaching 177 inches. Overfill of the vessel after a trip would only occur if level exceeded approximately 235.5 inches. Since the feedwater pumps, the High Pressure Coolant Injection (HPCI) turbine, and the Reactor Core Isolation Cooling (RCIC) turbine all receive trip signals prior to level reaching 177 inches, a substantial margin exists. VYNPS operating history has demonstrated that this margin greatly exceeds vessel level overshoot during transient events.

Based on this, there is adequate confidence that the vessel level will remain well below the main steam lines under CPPU conditions. The HPCI and RCIC pump trip functions are routinely verified as required by TSs and are considered very reliable.

The modification adding a recirculation pump runback following a RFP trip will not affect the plant response to this transient. The reactor scram signal from the MSIV limit switches will result in control rod insertion prior to any manual or automatic operation of the RFPs. Since control rods will already be inserted, a subsequent runback of the recirculation pumps will not affect the plant response.

BVY 03-98 / Attachment 7 / Page 5 The modification (BVY 03-23 "ARTS/MELLLA") to add an additional unpiped Spring Safety Valve (SSV) will not affect the plant response to this transient. The new third SSV will have the same lift setpoint as the two existing SSVs. This transient does not result in an opening of a SSV, nor is credit taken for SSV actuation.

Generator Load Reject and Turbine Trip Testing "Generator Load Rejection From High Power Without Bypass" (GLRWB) is an Abnormal Operational Transient as described in Chapter 14 of the VYNPS Updated Final Safety Analysis Report (UFSAR).

This transient competes with the turbine trip without bypass as the most limiting overpressurization transient that challenges thermal limits for each cycle. The turbine trip and generator load reject are essentially interchangeable. The only differences are 1) whether the RPS signal originates from the acceleration relay (GLRWB) or from the main turbine stop valves (turbine trip), and 2) whether the control valves close shutting off steam to the turbine or the stop valves close to isolate steam to the turbine. Both tests would verify the same analytical model for plant response. Therefore, the GLRWB is considered bounding or equivalent to the Turbine Trip.

The GLRWB analysis assumes that the transient is initiated by a rapid closure of the turbine control valves. It also assumes that all bypass valves fail to open. The CLTR states that: "The same performance criteria will be used as in the original power ascension tests, unless they have been replaced by updated criteria since the initial test program." The startup test for generator load reject allowed the select rod insert feature to reduce the reactor power level and, in conjunction with bypass valve opening, control the transient such that the reactor does not scram.

Current VYNPS design does not include the select rod insert feature.

The plant was also modified to include a scram from the acceleration relay of the turbine control system. Under current plant design, the original generator load reject test can not be re-performed. If a generator load reject with bypass test were performed, the results would be much less significant than the generator load reject without bypass closure analysis performed for CPPU.

The original generator load reject test was intended to demonstrate the following:

1. Determine and demonstrate reactor response to a generator trip, with particular attention to the rates of changes and peak values ofpoiver level, reactor steam pressure and turbine speed.

Criteria:

a. All test pressure transients must have maximum pressure values below 1230 psig

b. Maximum reactor pressure should be 35 psi below the first safety valve setpoint. (This is margin for safety valve weeping).

c.

The select rod insert feature shall operate and in conjunction with proper bypass valve opening, shall control the transient such that the reactor does not scram.

Due to plant modification discussed above, criterion c. above would no longer be applicable for a generator load reject test. The generator load reject startup test was performed at 93.7% power; however, a reactor scram occurred during testing and invalidated the test. A design change to initiate an immediate scram on generator load reject was implemented and this startup test was subsequently cancelled since it was no longer applicable.

BVY 03-98 / Attachment 7 / Page 6 Item 1 Reactor Response For a generator load reject with bypass event, given current plant design, the fast closure of the Turbine Control Valves (TCVs) cause a trip of the acceleration relay in the turbine control system. The acceleration relay trip initiates a full reactor scram. The bypass valves open, however, since the capacity of the bypass valves at CPPU is 87%, vessel pressure increases. This results in an increase in reactivity. The negative reactivity of the TCV fast closure scram from the acceleration relay should offset the positive reactivity of the pressure increase such that there is a minimal increase in heat flux. Therefore, the thermal performance during a generator load rejection test would be much less limiting than any of the transients routinely re-evaluated.

CPPU will have minimal impact on the components important to achieving the desired thermal performance.

Reactor Protection system (RPS) logic is unaffected and with no steam dome pressure increase, overall control rod insertion times will not be significantly affected. A trip channel and alarm functional test of the turbine control valve fast closure scram is performed every three months in accordance with plant technical specifications.

This trip function is considered very reliable.

Reactor Pressure Due to the minimal nature of the flux transient, the expected reactor pressure rise, Criteria a. and

b. above, are largely dependent on SRV setpoint performance.

Refer to the MSIV closure Reactor Pressure section above for discussion of SRV setpoint performance.

Because rated vessel steam dome pressure is not being increased and SRV setpoints are not being changed, there is no increase in the probability of leakage after a SRV lift. Since SRV leakage performance is considered acceptable at the current conditions, which match CPPU conditions with respect to steam dome pressure and SRV setpoints, SRV leakage performance will continue to be acceptable at CPPU conditions. A generator load rejection test would provide no significant additional confirmation of performance criteria a. and b. than the routine component testing performed every cycle, in accordance with the VYNPS TSs.

Other Plant Systems and Components Response The turbine control system acceleration relay hydraulic fluid pressure switches that provide the scram signal are highly reliable devices that are suitable for all aspects of this application including environmental requirements. There is no direct effect by any CPPU changes on these pressure switches. These switches are expected to be equally reliable before and after CPPU.

The Reactor Protection System (RPS) and Control Rod Drive (CRD) components that convert the scram signals into CRD motion are not directly affected by any CPPU changes. Minor changes in pressure drops across vessel components may result in very slight changes in control blade insertion rates. These changes have been evaluated and determined to be insignificant.

The ability to meet the scram performance requirement is not affected by CPPU. TS requirements for these components will continue to be met.

BVY 03-98 / Attachment 7 / Page 7 CPPU Modifications As previously described, Feedwater System operation will require all three feed pumps at CPPU conditions. Operation of the additional Reactor Feed Pump (RFP) will not affect plant response to this transient. All feedwater pumps receive a trip signal prior to level reaching 177 inches.

Overfill of the vessel after a trip would only occur if level exceeded approximately 235.5 inches.

Since the feedwater pumps, the High Pressure Coolant Injection (HPCI) turbine, and the RCIC turbine all receive trip signals prior to level reaching 177 inches, a substantial margin exists.

VYNPS operating history has demonstrated that this margin greatly exceeds vessel level overshoot during transient events. Based on this, there is adequate confidence that the vessel level will remain well below the main steam lines under CPPU conditions. The HPCI and RCIC pump trip functions are routinely verified as required by TSs and are considered very reliable.

The modification adding a recirculation pump runback following a RFP trip will not affect the plant response to this transient. The reactor scram signal from turbine control valve fast closure will result in control blade insertion prior to any manual or automatic operation of the RFPs.

Since control blades will already be inserted, a subsequent runback of the recirculation pumps will not affect the plant response.

The ARTS/MELLLA modification (BVY 03-23) to add an additional unpiped SSV will not affect the plant response to this transient. The new third SSV will have the same lift setpoint of the two existing SSVs. This transient does not result in an opening of a SSV nor is credit taken for SSV actuation.

HP Turbine modification replaces the steam flow path but will not affect the turbine control system hydraulic pressure switches that provide the turbine control valve fast closure scram signal to the RPS system.

Industry Boiling Water Reactor (BWR) Power Uprate Experience Southern Nuclear Operating Company's (SNC) application for EPU of Hatch Units I and 2 was granted without requirements to perform large transient testing. VYNPS and Hatch are both BWR/4 with Mark 1 containments. Although Hatch was not required to perform large transient testing, Hatch Unit 2 experienced an unplanned event that resulted in a generator load reject from 98% of uprated power in the summer of 1999. As noted in SNOC's LER 1999-005, no anomalies were seen in the plant's response to this event. In addition, Hatch Unit I has experienced one turbine trip and one generator load reject event subsequent to its uprate (i.e., LERs 2000-004 and 2001-002). Again, the behavior of the primary safety systems was as expected. No new plant behaviors were observed that would indicate that the analytical models being used are not capable of modeling plant behavior at EPU conditions.

The KKL power uprate implementation program was performed during the period from 1995 to 2000. Power was raised in steps from its previous operating power level of 3138 MWt (i.e.,

104.2% of OLTP) to 3515 MWt (i.e., 116.7% OLTP). Uprate testing was performed at 3327 MWt (i.e., 110.5% OLTP) in 1998, 3420 MWt (i.e., 113.5% OLTP) in 1999 and 3515 MWt in 2000.

KKL testing for major transients involved turbine trips at 110.5% OLTP and 113.5% OLTP and a generator load rejection test at 104.2% OLTP.

The KKL turbine and generator trip testing

BVY 03-98 / Attachment 7 / Page 8 demonstrated the performance of equipment that was modified in preparation for the higher power levels. Equipment that was not modified performed as before. The reactor vessel pressure was controlled at the same operating point for all of the prated power conditions.

No unexpected performance was observed except in the fine-tuning of the turbine bypass opening that was done as the series of tests progressed. These large transient tests at KKL demonstrated the response of the equipment and the reactor response.

The close matches observed with predicted response provide additional confidence that the uprate licensing analyses consistently reflected the behavior of the plant.

Plant Modeling, Data Collection, and Analyses From the power uprate experience discussed above, it can be concluded that large transients, either planned or unplanned, have not provided any significant new information about transient modeling or actual plant response. Since the VYNPS uprate does not involve reactor pressure changes, this experience is considered applicable.

The safety analyses performed for VYNPS used the NRC-approved ODYN transient modeling code. The NRC accepts this code for GE BWRs with a range of power levels and power densities that bound the requested power uprate for VYNPS. The ODYN code has been benchmarked against BWR test data and has incorporated industry experience gained from previous transient modeling codes.

ODYN uses plant specific inputs and models all the essential physical phenomena for predicting integrated plant response to the analyzed transients. Thus, the ODYN code will accurately and/or conservatively predict the integrated plant response to these transients at CPPU power levels and no new information about transient modeling is expected to be gained from performing these large transient tests.

CONCLUSION VYNPS believes that sufficient justification has been provided to demonstrate that an MSIV closure test, turbine trip test, and generator load rejection test is not necessary or prudent. Also, the risk imposed by intentionally initiating large transient testing should not be incurred unnecessarily.

As such, Entergy does not plan to perform additional large transient testing following the VYNPS CPPU.

Docket No. 50-271 BVY 03-98 Attachment Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 263 Supplement No. 3 Extended Power Uprate - Updated Information Review Guidance Matrix

Matrix 1 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Materials and Chemical Engineering Areas of Review Applicable to Primary

'Secondary SRP.

Focus of SRP Other Template Acceptance Review

.Review Section Usage

Guidance, Safety Evaluation

- 'Review Branch' Branch(es)

Number' Section Number CPPU SARI CPPU LTR BWR PWR Reactor Vessel Material All EPUs EMCB SRXB 5.3.1 GDC-14 RG 1.190 2.1.1 2.1.1 3.2.1 Surveillance Program Draft Rev. 2 GDC-31 April 1996 10 CFR 50, App. H 10 CFR 50.60 Pressure-Temperature Limits and All EPUs EMCB SRXB 5.3.2 GDC-14 RG 1.161 2.1.2 2.1.2 3.2.1 Upper-Shelf Energy Draft Rev. 2 GDC-31 RG 1.190 April 1996 10 CFR 50, App. G RG 1.99 10 CFR 50.60 Pressurized Thermal Shock PWR EPUs EMCB SRXB 5.3.2 GDC-14 RG 1.190 2.1.3 NA for Draft Rev. 2 GDC-31 RG 1.154 BWRs April 1996 10 CFR 50.61 Reactor Internal and Core All EPUs EMCB SRXB 4.5.2 GDC-1 Note 1*

2.1.3 2.1.4 10.7 Support Materials Draft Rev. 3 10 CFR 50.55a April 1996 BYV 03-981 PAGE I MATRIX I OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to Primary Secondary SRP Focus of SRP

'Other Template'.

Acceptance Brach Beoanches NubRP

'ou ofSP Seto ube USR Review

-Review -

Section Usage; Guidance Safety Evaluation Review

,-PPULTBranch, Branch(es)

Number

-Number CppU SAR l,

I,,,

CPPUU TR Reactor Coolant Pressure Boundary Materials All EPUs EMCB EMEB SRXB 5.2.3 Draft Rev. 3 April 1996 GDC-1 10 CFR 50.55a GDC-4 GDC-14 GDC-31 10 CFR 50, App. G RG 1.190 GL 97-01 IN 00-17s1 BL 01-01 BL 02-01 BL 02-02 Note 2*

Note 3*

2.1.4 2.1.5 2.5.3, 3.2.1, 3.2.2 and 10.7 4.5.1 GDC-1 Draft Rev. 3 10 CFR 50.55a April1996 GDC-14 5.2.4 10 CFR 50.55a Draft Rev. 2 April 1996 5.3.1 Draft Rev. 2 April 1996 5.3.3 Draft Rev. 2 April 1996 6.1.1 Draft Rev. 2 April 1996 GDC-1 10 CFR 50.55a GDC-4 GDC-14 GDC-31 10 CFR 50, App. G Leak-Before-Break PWR EPUs EMCB Protective Coating Systems All EPUs EMCB (Paints) - Organic Materials Effect of EPU on All EPUs EMCB Flow-Accelerated Corrosion 2.1.6 NA for BWRs 2.1.5 2.1.7 4.2.6 l4 4

2.1.6 2.1.8 10.7 BYV03-981PAGE2 MATRIX I OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to

.a Primary Secondary SRP

Review, Review Section Branch Branch(es)

Number Focus of SRP Usage

-Other Guidance Template Safety Evaluation Section Number Acceptance Review CPPU SAR I CPPU LTR Steam Generator Tube Inservice PWR EPUs EMCB Inspection Steam Generator Blowdown PWR EPUs EMCB System Chemical and Volume Control PWR EPUs EMCB System (Including Boron Recovery System)

Reactor Water Cleanup System BWR EPUs EMCB 4

+

5.4.2.2 Draft Rev. 2 April 1996 10 CFR 50.55a 2.1.9 NA for BWRs 2.1.10 NA for BWRs 10.4.8 GDC-14 Draft Rev. 3 April 1996 SPLB 9.3.4 SRXB Draft Rev. 3 April 1996 GDC-14 GDC-29 2.1.11 NA for BWRs 3.11 and 10.7 5.4.8 GDC-14 Draft Rev. 3 GDC-60 April 1996 GDC-61 Notes:

1. In addition to the SRP, guidance on neutron irradiation-related threshold for inspection for Irradiation-assisted stress-corrosion cracking for BWRs Is in BWRVIP-26 and for PWRs in BAW-2248 for E>1 MeV and In WCAP-14577 for E>0.1 MeV. For intergranular stress-corrosion cracking and stress-corrosion cracking In BWRs, review criteria and review guidance is contained in BWRVIP reports and associated staff safety evaluations. For thermal and neutron embrittlement of cast austenitic stainless steel, stress-corrosion cracking, and void swelling, applicants will need to provide plant-specific degradation management programs or participate In industry programs to investigate degradation effects and determine appropriate management programs.

2.

For thermal aging of cast austenitic stainless steel, review guidance and criteria Is contained In the May 19, 2000, letter from C. Grimes to D. Walters, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components.'

3.

For ntergranular stress corrosion cracking in BWR piping, review criteria and review guidance Is contained In BWRVIP reports, NUREG-0313, Rev. 2, GL 88-01, and associated safety evaluations.

BYV 03-98/ PAGE 3 MATRIX 1 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 2 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Mechanical and Civil Engineering Areas of Review Applicabie io Primary Secondary SRP Review Review-Sectio'n Branch Branch(es)

Number Focus of SRP Other Usage Guidanci

-.I..Template

Safety, e

Evaluation Section:

' I ' Number BWR PWR Acceptance

' Review CPPU SAR /

CPPU LTR Pipe Rupture Locations and All EPUs EMEB Associated Dynamic Effects Pressure-Retaining All EPUs EMEB Components and Component Supports 3.6.2 GDC-4 Draft Rev. 2 April 1996 3.9.1 GDC-1 Draft Rev. 3 GDC-2 April 1996 GDC-14 GDC-15 2.2.1 2.2.1 10.1 and 10.2 2.2.2 2.2.2 2.5.3,3.1, 3.2.2, 3.4, 3.5, 3.7, and 3.8 3.9.2 Draft Rev. 3 April 1996 GDC-1 GDC-2 GDC-4 GDC-14 GDC-15 IN 95-016 IN 02-026 3.9.3 10 CFR 50.55a IN 96-049 Draft Rev. 2 GDC-1 GL 96-06 April 1996 GDC-2 GDC-4 GDC-14

'GDC-15 5.2.1.1 Draft Rev. 3 April 1996 10 CFR 50.55a GDC-1 RG 1.84 RG 1.147 DG 1.1089 DG 1.1090 DG 1091 BVY 03-98 I PAGE 4 MATRIX 2 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

S S Areas of Review ' '

r Applicable to

Primary, Secondary' SRP.

Review'

- Review-Section

'.Branch Branch(es)

Number

', Focus of SRP Usage-,

. Other Guidance

- Template Safety Evaluation Section Number

-Acceptance Review CPPU SAR /

. CPPU LTR BWR PWR I

Reactor Pressure Vessel Internals and Core Supports All EPUs EMEB 3.9.1 GDC-1 Draft Rev. 3 GDC-2 April 1996 4

4 2.2.3 2.2.3 3.1, 3.3, and 3.4.2 3.9.2 Draft Rev. 3 April 1996 GDC-1 GDC-2 GDC-4 IN 95-016 IN 02-026 3.9.3 10 CFR 50.55a IN 96-049 Draft Rev. 2 GDC-1 GL 96-06 April 1996 GDC-2 GDC-4 3.9.5 Draft Rev. 3 April 1996 10 CFR 50.55a GDC-1 GDC-2 GDC-4 GDC-10 IN 02-026

1.

9 4

4.

4.

Safety-Related Valves and Pumps All EPUs EMEB 3.9.3 Draft Rev. 2 April 1996 GDC-1 10 CFR 50.55a(f)

IN 96-049 GL 96-06 2.2.4 2.2.4 3.1,3.8, 4.1.3,4.4.4, 4.1.6, and 4.2

4.

9 3.9.6 Draft Rev. 3 April 1996 GDC-1 GDC-37 GDC-40 GDC-43 GDC-46 GDC-54 10 CFR 50.55a(f)

GL 89-10 GL 95-07 GL 96-05 IN 97-090 IN 96-048s1 IN 96-048 IN 96-003 RIS 00-003 RIS01-015 RG 1.147 RG 1.175 DG 1089 DG 1091 a

a a

a BVY 03-981 PAGES MATRIX 2 OF SECTION 2.1 OF RS-001 (DRAFT)

BVY03-98 1PAGE 5 MATRIX 2 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to. '-. I :

Primary I Review Branch Secondary

'Review Branch(es)

SRP Section

-Number -

. Focus of SRP

, Usage

Other Guidance -

Template Safety Evaluation Section; Number Acceptance Review CPPU SAR /

CPPU LTR BWR l PWR Seismic and Dynamic All EPUs EMEB EEIB 3.10 GDC-1 Qualification of Mechanical and Draft Rev. 3 GDC-2 Electrical Equipment April 1996 GDC-4 GDC-14 GDC-30 10 CFR 100, App. A 10 CFR 50, App. B USI A-46 2.2.5 2.2.5 10.1 and 10.3.3 BVY3-81A0E MTRX OF. SETO 2.1 O_

(DR.T BVY 03-98 /PAGE 6 MATRIX 2 OF SECTION 2.1 OF RS001 (DRAFT)

DECEMBER 2002

MATRIX 3 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Electrical Engineering Areas of Review',' ' '

Applicable to Primary

' Seconda'ry Review Review Branch Branch(es):

Environmental Qualification of All EPUs EEIB Electrical Equipment Offsite Power System All EPUs EEIB AC Onsite Power System All EPUs EEIB DC Onsite Power System All EPUs EEIB SRP Fcus of SRP Section Usage Number 3.11 10 CFR 50.49 Draft Rev. 3 April 1996 8.1 GDC-17 Draft Rev. 3 April 1996 8.2 GDC-17 Draft Rev. 4 April 1996 8.2, App. A GDC-17 Draft Rev. 4 April 1996 8.1 GDC-17 Draft Rev. 3 April 1996 8.3.1 GDC-17 Draft Rev. 3 April 1996 8.1 GDC-17 Draft Rev. 3 10 CFR 50.63 April 1996 8.3.2 GDC-17 Draft Rev. 3 10 CFR 50.63 April 1996

.'Other Guidance

- Template Safety '

Evaluation Section Number Acceptance Review -

CPPU SAR /

CPPU LTR BWR PWR BTP PSB-1 Draft Rev. 3 April 1996 BTP ICSB-11 Draft Rev. 3 April 1996 2.3.1 2.3.1 10.3.1 2.3.2 2.3.2 6.1.1 2.3.3 2.3.3 6.1.2 2.3.4 2.3.4 6.2 BVY03-98/PAGE7 MATRIX 3 OF SECTION 2.1 OF RSO001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to Primary Secondary-SRP Focus of SRP Other Template Safety Acceptance

'Review

-Review Section Usage Guidance Evaluation Section Review Branch'.

Branch(es)

Number Number CPPU SAR CPPU LTR BWR PWR Station Blackout All EPUs EEIB SPLB 8.1 10 CFR 50.63 Note 1*

2.3.5 2.3.5 9.3.2 SRXB Draft Rev. 3 April 1996 8.2, App. B 10 CFR 50.63 Draft Rev. 4 April 1996

1. The review of station blackout includes the effects of the EPU on systems required for core cooling in the station blackout coping analysis (e.g., condensate storage tank inventory, controls and power supplies for relief valves, residual heat removing system, etc.) to ensure that the effects are accounted for In the analysis.

BW 03.98IPAGE8 MATRIX 3 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 4 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Instrumentation and Controls Areas of Review '

Applicable to

Primary, Secondary

' Review Review Branch Branch(es)

Reactor Trip System All EPUs EEIB Engineered Safety Features All EPUs EEIB Systems Safety Shutdown Systems All EPUs EEIB Control Systems All EPUs EEIB Diverse I&C Systems All EPUs, EEIB SRP FocusofSRP Section'

Usage, Number 7.2 10 CFR 50.55(a)(1)

Rev. 4 10 CFR 50.55a(h)

June 1997 GDC-1 GDC-4 7.3 GDC-13 Rev. 4 GDC-19 June 1997 GDC-20 GDC-21 GDC-22 GDC-23 GDC-24 7.4 10 CFR 50.55(a)(1)

Rev. 4 10 CFR 50.55a(h)

June 1997 GDC-1 GDC-4 GDC-13 GDC-19 GDC-24 7.7 10 CFR 50.55(a)(1)

Rev. 4 10 CFR 50.55a(h)

June 1997 GDC-1 GDC-13 7.8 GDC-19 Rev. 4 GDC-24 June 1997 Other Guidance _

Template Safety Evaluation Section Number Acceptance Review CPPU SAR /

CPPU LTR BWR PWR 2.4.1 2.4.1 5.3 2.4.1 2.4.1 5.3 2.4.1 2.4.1 5.3 2.4.1 2.4.1 5.1 and 5.2 2.4.1 2.4.1 5.3 and 9.3.1 BVY 03.981 PAGE 9 MATRIX 4 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review I Applicable to Primary

- Review'

.Branch Secondary SRP-

- Review..

Sectio~n.

branch(es).

Number, l

~~~~7.0 111 11,June 19g97

. Focus of SRP Other

.Usage.

Guidance Template Safety Acceptance Evaluation Section Review Number CPPU SAR I CPPU LTR BWR PWR General guidance for use of other All EPUs EEIB SRP Sections related to l&C BVY 03-981 PAGE 10 MATRIX 4 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 5 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Plant Systems I

Y Areas of Review Applicable to Primary

Review Branch Secondary SRP Review Section' Branch(es)

Number.-

Focus of SRP Other Usage Guidance

- Template Safety Evaluation Section Number Acceptance

Review, CPPU SAR /

CPPU LTR BWR PWR Flood Protection EPUs that result in significant SPLB Increases in fluid volumes of tanks and vessels Equipment and Floor Drainage EPUs that result In increases in SPLB System fluid volumes or in installation of larger capacity pumps or piping systems Circulating Water System EPUs that result in Increases in SPLB fluid volumes associated with the circulating water system or in installation of larger capacity pumps or piping systems 3.4.1 GDC-2 Rev. 2 July 1981 9.3.3 GDC-2 Rev. 2 GDC-4 July 1981 10.4.5 GDC-4 Rev. 2 July 1981 2.5.1.1.1 2.5.1.1.1 10.1.2 2.5.1.1.2 2.5.1.1.2 8.1*

2.5.1.1.3 2.5.1.1.3 6.4.2' 2.5.1.2.1 2.5.1.2.1 10.1.2**

2.5.1.2.1 2.5.1.2.1 10.1.2t*

Internally Generated Missiles (Outside Containment)

EPUs that result In substantially higher system pressures or changes In existing system configuration SPLB EMCB EMEB 3.5.1.1 Rev. 2 July 1981 GDC-4 Internally Generated Missiles EPUs that result In substantially SPLB EMCB 3.5.1.2 GDC-4 (Inside Containment) higher system pressures or EMEB Rev. 2 changes In existing system July 1981 configuration BVY 03-98/PAGE II

• No IGNIFICANT INCREASE IN FLUID VOLUMES

• • No INCREASE IN SYSTEM PRESSURE MATRIX 5 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

I Areas of Review Applicable to Primary Review Branch Secondary:

Review Branch(es)

+

1-Protection Against Postulated Piping Failures in Fluid Systems Outside Containment EPUs that affect environmental conditions, habitability of the control room, or access to areas important to safe control of postaccident operations SPLB EMCB EMEB Fire Protection Program All EPUs except where the SPLB application demonstrates that previous analysis is bounding PWR Dry Containments, EPUs for PWR plants with dry SPLB Including Subatmospheric containments (including Containments subatmospheric containments) except where the application demonstrates that previous analysis is bounding Ice Condenser Containments EPUs for PWR plants with ice SPLB condenser containments except where the application demonstrates that previous analysis is bounding Pressure-Suppression Type EPUs for BWR plants with SPLB BWR Containments pressure-suppression containments except where the application demonstrates that previous analysis is bounding SRP Focus of SRP Section Usage' Number 3.6.1 GDC-4 Rev. 1 July 1981 9.5.1 10 CFR 50.48 Rev. 3 10 CFR 50, App. R July 1981 GDC-3 GDC-5 6.2.1 GDC-13 Rev. 2 GDC-16 July 1981 GDC-38 GDC-50 6.2.1.1.A GDC-64 Rev. 2 July 1981 6.2.1 GDC-13 Rev. 2 GDC-16 July 1981 GDC-38 GDC-50 6.2.1.1.B GDC-64 Rev. 2 July 1981 6.2.1 GDC-4 Rev. 2 GDC-13 July 1981 GDC-16 GDC-50 6.2.1.1.C GDC-64 Rev. 6 Aug. 1984 Other Guidance 2.5.1.3 2.5.1.3 10.1 and 10.2 I

BWR I Template Safety:

Evaluation Section Number Acceptance Review CPPU SAR /

CPPU LTR PWR Note 1' 2.5.1.4 2.5.1.4 6.7 NA for BWRs

+

NA for BWRs 4.1 through 4.1.2 BVY~~

.398 PAE1 ARX5O ETO FR-O DAT BVY 03-98 /PAGE 12 MATRIX 5 OF SECTDON 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to Areas of Review -

Applicable to - I Primary" Secondary Review Review Branch Branch(es)

Subcompartment Analysis All EPUs except where the SPLB application demonstrates that previous analysis is bounding Mass and Energy Release All EPUs except where the SPLB Analysis for Postulated application demonstrates that Loss-of-Coolant previous analysis is bounding Mass and Energy Release PWR EPUs except where the SPLB Analysis for Postulated application demonstrates that Secondary System Pipe previous analysis is bounding Ruptures Combustible Gas Control In EPUs that impact hydrogen SPLB Containment release assumptions Containment Heat Removal All EPUs except where the SPLB application demonstrates that previous analysis is bounding Secondary Containment EPUs that affect the pressure SPLB Functional Design and temperature response, or draw-down time of the secondary containment SRP Focus of SRP

- Section Usage Number 6.2.1 GDC-4 Rev. 2 GDC-50 July 1981 6.2.1.2 Rev. 2 July 1981 6.2.1 GDC-50 Rev. 2 10 CFR 50, App. K July 1981 6.2.1.3 Rev.1 July 1981 6.2.1 GDC-50 Rev. 2 July 1981 6.2.1.4 Rev. 1 July 1981 6.2.5 10 CFR 50.44 Rev. 2 10 CFR 50.46 July 1981 GDC-5 GDC-41 GDC-42 GDC-43 6.2.2 GDC-38 Rev. 4 Oct. 1985 6.2.3 GDC-4 Rev. 2 GDC-16 July 1981 Other Guidance Template Safety Evaluation Section Number Acceptance Review CPPU SAR /

CPPU LTR 2.5.2.2 2.5.2.2 4.1.2.3 2.5.2.3.1 2.5.2.3.1 4.1.1 through 4.1.2.2 2.5.2.3.2 NA for BWRs 2.5.2.4 2.5.2.4 4.7 DG-1 107 2.5.2.5 2.5.2.5 3.10 4.5 2.5.2.6 BW03-981PAGE 13 MATRIX 5 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

F V

1 I.

Areas of Review Applicable to

- Primary Review Branch Secondary, Review Branch(es):

- SRP-,

Section Number,

Focus of SRP Usage-,-,

Other Template Safety Guidance Evaluation Section I

Numiber Acceptance Review CPPU SAR /

CPPU LTR Minimum Containment Pressure PWR EPUs except where the SPLB SRXB 6.2.1 10 CFR 50.46 Analysis for Emergency Core application demonstrates that Rev. 2 10 CFR 50, App. K Cooling System Performance previous analysis Is bounding July 1981 Capability Studies 6.2.1.5 Rev. 2 July 1981 Pressurizer Relief Tank PWR EPUs that affect SPLB EMEB 5.4.11 GDC-2 pressurizer discharge to the PRT Rev. 2 GDC-4 July 1981 Control Room Habitability System All EPUs except where the SPLB SPSB 6.4 GDC-4 application demonstrates that Draft Rev. 3 GDC-19 previous analysis Is bounding April 1996 ESF Atmosphere Cleanup All EPUs except where the SPLB SPSB 6.5.1 GDC-19 System application demonstrates that Rev. 2 GDC-41 previous analysis Is bounding July 1981 GDC-61 GDC-64 Fission Product Control Systems All EPUs except where the SPLB EMCB 6.5.3 GDC-41 and Structures application demonstrates that Rev. 2 previous analysis is bounding July 1981 Main Condenser Evacuation EPUs for which the main SPLB 10.4.2 GDC-60 System condenser evacuation system is Rev. 2 GDC-64 modified

______July 1981 Turbine Gland Sealing System EPUs for which the turbine gland SPLB 10.4.3 GDC-60 sealing system is modified Rev. 2 GDC-64

~~July 1981 Main Steam Isolation Valve BWR EPU that affect the amount SPLB 6.7 GDC-54 Leakage Control System of valve leakage that Is assumed Rev. 2 and resultant dose July 1981 consequences.

Control Room Area Ventilation All EPUs except where the SPLB SPSB 9.4.1 GDC-4 System application demonstrates that Rev. 2 GDC-19 previous analysis is bounding July 1981 GDC-60 2.5.2.6 NA for BWRs 2.5.2.7 NA for BWRs Note 2-2.5.3.1 Note 3' 2.5.3.1 4.4 2.5.3.2 2.5.3.2 4.5 2.5.3.3 2.5.3.3 4.5 2.5.3.4 2.5.3.4 7.2 (no mod) 2.5.3.5 2.5.3.5 7.1 (no mod) 2.5.3.6 4.6 2.5.4.1 2.5.4.1 4.4 BVY 03-98 /PAGE 14 MATRIX 5 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

l X

z Areas of Review Applicable to PrimaryI

-Review-Branch

- Secondary Review.

Branch(es)

SRP Section Number Focus of SRP Usage

• Other Template Safety Guidance Evaluation Section Number Acceptance

- Review CPPU SAR /

CPPU LTR Spent Fuel Pool Area Ventilation System All EPUs except where the application demonstrates that previous analysis Is bounding SPLB SPSB 9.4.2 Rev. 2 July 1981 GDC-60 GDC-61 Auxiliary and Radwaste Area All EPUs except where the SPLB Ventilation System application demonstrates that previous analysis is bounding Turbine Area Ventilation System All EPUs except where the SPLB application demonstrates that previous analysis is bounding ESF Ventilation System All EPUs except where the SPLB application demonstrates that previous analysis is bounding Spent Fuel Pool Cooling and All EPUs except where the SPLB Cleanup System application demonstrates that previous analysis Is bounding Station Service Water System All EPUs except where the SPLB application demonstrates that previous analysis is bounding Reactor Auxiliary Cooling Water All EPUs except where the SPLB Systems application demonstrates that previous analysis Is bounding Ultimate Heat Sink All EPUs except where the SPLB application demonstrates that previous analysis Is bounding 9.4.3 GDC-60 Rev. 2 July 1981 9.4.4 GDC-60 Rev. 2 July 1981 9.4.5 GDC-4 Rev. 2 GDC-17 July 1981 GDC-60 2.5.4.2 2.5.4.2 6.6 2.5.4.3 2.5.4.3 6.6 2.5.4.3 2.5.4.3 6.6 2.5.4.4 2.5.4.4 6.6 EMCB 9.1.3 Rev. 1 July 1981 GDC-5 GDC-44 GDC-61 Note 4*

2.5.5.1 2.5.5.1 6.3 9.2.1 GDC-4 GL 89-13 2.5.5.2 2.5.5.2 6.4.1 and Rev. 4 GDC-5 and 6.4.4 June 1985 GDC-44 Suppl. 1 GL 96-06 and Suppl. I 9.2.2 Rev. 3 June 1986 GDC-4 GDC-5 GDC-44 GL 89-13 and Suppl. 1 GL 96-06 and SuDpl 1 2.5.5.3 2.5.5.3 6.4.3 9.2.5 GDC-5 Rev. 2 GDC-44 July 1981 2.5.5.4 2.5.5.4 6.4.5 BVY03-98/PAGE 15 MATRIX 5 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

z

- r S L

Areas of Review '

Applicable to, Secondary SRP

.Review Section, -'-

'Branch(es),

Number. -'

.'Focus of SRP

,' Other Template Safety Usage' Guidance

'Evaluation Section.

Number Acceptance Review -

CPPU SAR /

CPPU LTR I

Auxiliary Feedwater System PWR EPUs except where the application demonstrates that previous analysis is bounding SPLB Main Steam Supply System All EPUs except where the SPLB application demonstrates that previous analysis is bounding Main Condenser All EPUs except where the SPLB application demonstrates that previous analysis is bounding Turbine Bypass System All EPUs except where the SPLB application demonstrates that previous analysis is bounding Condensate and Feedwater All EPUs except where the SPLB System application demonstrates that previous analysis is bounding 10.4.9 GDC-4 Rev. 2 GDC-5 July 1981 GDC-19 GDC-34 GDC-44 10.3 GDC-4 Rev. 3 GDC-5 April 1984 GDC-34 10.4.1 GDC-60 Rev. 2 July 1981 10.4.4 GDC-4 Rev. 2 GDC-34 July 1981 10.4.7 GDC-4 Rev. 3 GDC-5 April 1984 GDC-44 I%

I I 2.5.5.5 NA for BWRs 2.5.6.1 3.5.2 and 7.3 Gaseous Waste Management Systems EPUs that impact the level of fission products in the reactor coolant system, or the amount of gaseous waste SPLB IEHB 11.3 Draft Rev. 3 April 1996 10 CFR 20.1302 GDC-3 GDC-60 GDC-61 10 CFR 50, App. I 2.5.6.2 2.5.6.2 7.2 2.5.6.3 2.5.6.3 7.3 2.5.6.4 2.5.6.4 7.4 2.5.7.1 2.5.7.1 8.2 2.5.7.2 2.5.7.2 8.1 2.5.7.3 2.5.7.3 8.1 Liquid Waste Management EPUs that impact the level of SPLB IEHB 11.2 10 CFR 20.1302 Systems fission products in the reactor Draft GDC-60 coolant system, or the amount of Rev. 3 GDC-61 liquid waste April 1996 10 CFR 50, App. I Solid Waste Management EPUs that impact the level of SPLB IEHB 11.4 10 CFR 20.1302 Systems fission products in the reactor Draft GDC-60 coolant system, or the amount of Rev. 3 GDC-63 solid waste April 1996 GDC-64 10 CFR 71 BVY 03 98/PAGE 16 MATRIX 5 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to

- Primary.

-Review Branch Secondary Review -

Branch(es).

SRP'

-Section-

- Number Focus of SRP

'- Other' Usage Guidance Template Safety Evaluation Section Number Acceptance Review CPPU SAR /

CPPU LTR Emergency Diesel Engine Fuel EPUs that result in higher EDG SPLB Oil Storage and Transfer System electrical demands 9.5.4 GDC-4 Rev. 2 GDC-5 July 1981 GDC-17 2.5.8.1 2.5.8.1 6.1.1 2.5.8.2 2.5.8.2 6.8 Light Load Handling System (Related to Refueling)

EPUs except where the application demonstrates that previous analysis Is bounding SPLB SPSB 9.1.4 Rev. 2 July 1981 GDC-61 GDC-62 Notes:

1.

Supplemental guidance for review of fire protection is provided in Attachment 2 to this matrix.

BVY03-98/PAGE 17 MATRIX 5 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 6 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Reactor Systems X,

Areas of Review Applicable to

. K Primary Secondary SRP Review-Review Section Branch

Branch(es)

Number Focus of SRP Usage Other -

Guidance Template Safety

• Evaluation Section -

- I.Number Acceptance Review CPPU SAR /

. CPPU LTR BWR PWR Fuel System Design All EPUs SRXB Nuclear Design All EPUs SRXB Thermal and Hydraulic Design All EPUs SRXB 4.2 10 CFR 50.46 Note 1*

2.6.1 2.6.1 2.1 Draft Rev. 3 GDC-10 Note 2*

April 1996 GDC-27 GDC-35 4.3 GDC-10 RG 1.190 2.6.2 2.6.2 2.2, 2.3, Draft Rev. 3 GDC-11 GSI 170 and 2.4 April 1996 GDC-12 IN 97-085 GDC-13 GDC-20 GDC-25 GDC-26 GDC-27 GDC-28 4.4 Draft Rev. 2 April 1996 GDC-1 0 GDC-12 Note 3*

2.6.3 2.6.3 2.2, 2.3, and 2.4 4.

Functional Design of Control Rod Drive System All EPUs SRXB SPLB 4.6 Draft Rev. 2 April 1996 GDC-4 GDC-23 GDC-25 GDC-26 GDC-27 GDC-28 GDC-29 10 CFR 50.62(c)(3) 2.6.4.1 2.6.4.1 2.5

.1.

I I

I BVY 03-98 /PAGE 18 MATRIX 6 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to -

Primary Secondary SRP Review.

Review:

Section -

Branch Branch(es)

Number I -

Focus of SRP

- Usage Other Guidance Template Safety Evaluation Section Number Acceptance

- Review CPPU SAR /

CPPU LTR BWR PWR Overpressure Protection during All EPUs SRXB Power Operation Overpressure Protection during PWR EPUs SRXB Low Temperature Operation Reactor Core Isolation Cooling BWR EPUs SRXB System Residual Heat Removal System All EPUs SRXB Emergency Core Cooling System All EPUs SRXB Standby Liquid Control System BWR EPUs SRXB Decrease in Feedwater All EPUs SRXB Temperature, Increase in Feedwater Flow, Increase In Steam Flow, and Inadvertent Opening of a Steam Generator Relief or Safety Valve

.4

.4

4.

4 I

5.2.2 Draft Rev. 3 April 1996 GDC-15 GDC-31 Note 4*

2.6.4.2 2.6.4.2 3.1 5.2.2 GDC-15 Draft Rev. 3 GDC-31 April 1996 5.4.6 GDC-4 Draft Rev. 4 GDC-5 April 1996 GDC-29 GDC-33 GDC-34 GDC-54 10 CFR 50.63 2.6.4.3 NA for BWRs 3.9 3.10 2.6.4.4 5.4.7 Draft Rev. 4 April 1996 GDC-4 GDC-5 GDC-1 9 GDC-34 Note 5*

2.6.4.4 6.3 Draft Rev. 3 April 1996 GDC-4 GDC-27 GDC-35 10 CFR 50.46 10 CFR 50 App. K Note 6*

2.6.5.6.2 2.6.5.6.3 4.2 and 4.3 6.5 EMCB SPLB 9.3.5 GDC-26 Note 12*

2.6.4.5 Draft Rev. 3 GDC-27 April 1996 10 CFR 50.62(c)(4) 15.1.1-4 Draft Rev. 2 April 1996 GDC-10 GDC-15 GDC-20 GDC-26 Note 7*

2.6.5.1 2.6.5.1.1 BVY 03-981 PAGE 19 MATRIX 6 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

ts Areas of Review Applicable to Primary Secondary' SRP Reviewt.,

Review Section Branch Branch(es)

Number

- Focus of SRP '

Usage

' Other Guidance Template Safety Evaluation Section

' Number Acceptance

'Review CPPU SAR /

CPPU LTR I

'BWR PWR Steam System Piping Failures PWR EPUs SRXB Inside and Outside of Containment Loss of External Load; Turbine All EPUs SRXB Trip, Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)

Loss of Nonemergency AC All EPUs SRXB Power to the Station Auxiliaries 15.1.5 GDC-27 Note 7*

Draft Rev. 3 GDC-28 April 1996 GDC-31 GDC-35 2.6.5.1.2 NA for BWRs 15.2.1-5 Draft Rev. 2 April 1996 GDC-10 GDC-1 5 GDC-26 Note 7*

2.6.5.2.1 2.6.5.2.1 3.1 and 9.1 15.2.6 GDC-10 Note 7*

2.6.5.2.2 2.6.5.2.2 3.1 and 9.1 Draft Rev. 2 GDC-15 April 1996 GDC-26 Loss of Normal Feedwater Flow All EPUs SRXB EEIB 15.2.7 Draft Rev. 2 April 1996 GDC-10 GDC-15 GDC-26 Note 7*

2.6.5.2.3 2.6.5.2.3 9.1

4.

4 4

4 4

4 Feedwater System Pipe Breaks Inside and Outside Containment PWR EPUs SRXB EEIB 15.2.8 Draft Rev. 2 April 1996 GDC-27 GDC-28 GDC-31 GDC-35 Note 7*

-F' 2.6.5.2.4 NA for BWRs Loss of Forced Reactor Coolant All EPUs SRXB Flow Including Trip of Pump Motor and Flow Controller Malfunctions Reactor Coolant Pump Rotor All EPUs SRXB Seizure and Reactor Coolant Pump Shaft Break Uncontrolled Control Rod All EPUs SRXB Assembly Withdrawal from a Subcritical or Low Power Startup Condition 15.3.1-2 Draft Rev. 2 April 1996 GDC-10 GDC-15 GDC-26 Note 7*

2.6.5.3.1 2.6.5.3.1 9.1 15.3.3-4 GDC-27 Note 7*

2.6.5.3.2 2.6.5.3.2 9.1 Draft Rev. 3 GDC-28 April 1996 GDC-31 15.4.1 GDC-10 Note 7*

2.6.5.4.1 2.6.5.4.1 5.3.4 and Draft Rev. 3 GDC-20 9.1 April 1996 GDC-25 BVY 03.981 PAGE 20 MATRIX 6 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

I Areas of Review b

Applicable to Primary Secondary' Review Review Branch Branch(es).

. SRP Section' Number' Focus of SRP

- Usage Other-Guidance Template Safety Evaluation Section, Number Acceptance Review CPPU SAR/

CPPU LTR

'BWR PWR Uncontrolled Control Rod All EPUs SRXB Assembly Withdrawal at Power Control Rod Misoperatlon PWR EPUs SRXB (System Malfunction or Operator Error)

Startup of an Inactive Loop or All EPUs SRXB Recirculation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate Chemical and Volume Control PWR EPUs SRXB System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant Spectrum of Rod Ejection PWR EPUs SRXB Accidents Spectrum of Rod Drop Accidents BWR EPUs SRXB Inadvertent Operation of ECCS All EPUs SRXB and Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory Inadvertent Opening of a PWR All EPUs SRXB Pressurizer Pressure Relief Valve or a BWR Pressure Relief Valve

-l I

4 15.4.2 Draft Rev. 3 April 1996 GDC-10 GDC-20 GDC-25 Note 7' 2.6.5.4.2 2.6.5.4.2 5.3.5 and 9.1 15.4.3 GDC-10 Note 7*

Draft Rev. 3 GDC-20 April 1996 GDC-25 15.4.4-5 GDC-10 Note 7*

Draft Rev. 2 GDC-15 April 1996 GDC-20 GDC-26 GDC-28 15.4.6 GDC-10 Note 7*

Draft Rev. 2 GDC-15 April 1996 GDC-26 15.4.8 GDC-28 Note 7*

Draft Rev. 3 April 1996 2.6.5.4.3 2.6.5.4.3 NA for BWRs 2.6.5.4.4 9.1 2.6.5.4.5 NA for*

BWRs 15.4.9 Draft Rev. 3 April 1996 GDC-28 Note 7*

NA for BWRs 9.2 9.1 15.5.1-2 Draft Rev. 2 April 1996 GDC-10 GDC-15 GDC-26 Note 7*

Note 8*

2.6.5.5 2.6.5.5 15.6.1 GDC-10 Note 7*

2.6.5.6.1 2.6.5.6.1 9.1 Draft Rev. 2 GDC-15 April 1996 GDC-26 BVY 03-981 PAGE 21 MATRIX 6 OF SECliON 2.1 OF RS.001 (DRAFT)

BWO 3-98/1PAGE 21 MATRIX 6 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review Applicable to Areas of Review Applica, ble to' -

Primary Secondary SRP -'

Review Review '

Sectlon

' Branch' ' Branch(es)

Number, Focus of SRP Usage Other Guidance I Template Safety

- Evaluation Section Number Acceptance

' Review CPPU SAR /

CPPU LTR Inadvertent Opening of a PWR All EPUs SRXB Pressurizer Pressure Relief Valve or a BWR Pressure Relief Valve Steam Generator Tube Rupture PWR EPUs SRXB Loss-of Coolant Accidents All EPUs SRXB Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary Anticipated Transient Without All EPUs SRXB Scram New Fuel Storage EPU applications that request SRXB approval for new fuel.

Spent Fuel Storage EPU applications that request SRXB approval for new fuel.

4.

I 15.6.1 Draft Rev. 2 April 1996 GDC-10 GDC-15 GDC-26 Note 7*

2.6.5.6.1 2.6.5.6.1 9.1 15.6.3 Note 7*

Note 7*

Draft Rev. 3 Note 9-Note 9*

April 1996 2.6.5.6.2 NA for BWRs 15.6.5 Draft Rev. 3 April 1996 GDC-35 10 CFR 50.46 Note 7*

Note 10' 2.6.5.6.2 2.6.5.6.3 4.3 and 9.2 Note 7*

Note 11*

Note 12-2.6.5.7 2.6.5.7 9.3 9.1.1 Draft Rev. 3 April 1996 GDC-62 9.1.2 GDC-4 Draft Rev. 4 GDC-62 April 1996 2.6.6.1 2.6.6.1 1.2.3 and.

2.1*

2.6.6.2 2.6.6.2 1.2.3 and 2.1*

Notes:

1.

When mixed cores (i.e., fuels of different designs) are used, the review covers the licensee's evaluation of the effects of mixed cores on design-basis accident and transient analyses.

2.

The current acceptance criteria for fuel damage for reactivity insertion accidents (RIAs) requires revision per Research Information Letter No. 174, Interim Assessment of Criteria for Analyzing Reactivity Accidents at High Bumup." The Office of Nuclear Regulatory Research Is conducting confirmatory research on RlAs and the Office of Nuclear Reactor Regulation is discussing the issue of fuel damage criteria with the nuclear power industry as part of the industry's proposal to Increase fuel bumup limits In the future. In the interim, current methods for assessing fuel damage In RIAs are considered acceptable based on the NRC staffs understanding of actual fuel performance, as shown in three-dimensional kinetic calculations which indicate acceptably low fuel cladding enthalpy.

BVY 03-98/ PAGE 22

• No NEW FUEL FOR CPPU MATRIX 6 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 7 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Source Terms and Radiological Consequences Analyses Areas of Review K

-. i.

Applicable to

' Primary Review Branch Secondary Review Branch(es).

.... I SRP Section

- Number Focus of SRP Other Template Safety Usage Guidance Evaluation Section Number-BWR PWR Acceptance

'Review CPPU SAR /

CPPU LTR-Source Terms for Input into All EPUs SPSB Radwaste Management Systems Analyses 11.1 10 CFR 20 Draft Rev. 3 10 CFR 50, App. I April 1996 GDC-60 2.7.1 2.7.1 8.4 Radiological Consequence Analyses Using Alternative Source Terms EPUs that utilize alternative source term SPSB EEIB EMCB EMEB IEHB SPLB SRXB 15.0.1 Rev. 0 July 2000 10 CFR 50.67 GDC-19 10 CFR 50.49 10 CFR 51 10 CFR 50, App. E NUREG-0737 2.7.2 2.7.2 9.2 Radiological Consequences of Main Steamline Failures Outside Containment for a PWR PWR EPUs that do not utilize alternative source term whose main steamline break analyses result in fuel failure SPSB SRXB 15.1.5, App. A Draft Rev. 3 April 1996 10 CFR 100 Notes 4, 5, 6, 7,27*

6.4 Draft Rev. 3 April 1996 GDC-19 Notes 1, 2, 3, 28, 29*

Radiological Consequences of EPUs that do not utilize SPSB SRXB 15.3.3-4 10 CFR 100 Notes 5, 8, Reactor Coolant Pump Rotor alternative source term whose Draft Rev. 3 9, 27*

Seizure and Reactor Coolant reactor coolant pump rotor April 1996 Pump Shaft Break seizure or reactor coolant pump shaft break results in fuel failure 6.4 GDC-19 Notes 1, 2, Draft Rev. 3 3, 28, 29*

April 1996 2.7.2 NA for BWRs 2.7.3 NA for BWRs l

BVY 03-98 PAGE 23 MATRIX 7 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

t Y

7 F

7 7

Areas of Review Applicable to Primary-Review Branch Secondary,-

Review

. Branch(es).-

SRP Section -

Number Focus of SRP Usage

- Other Guidance

....'i.

Template Safety Evaluation Section Number Acceptance Review CPPU SARI

. CPPU LTR LI BWR PWR Radiological Consequences of a PWR EPUs that do not utilize SPSB SRXB 15.4.8, App. A 10 CFR 100 Notes 4, Control Rod Ejection Accident alternative source term whose Draft Rev. 2 21,22, 27*

rod ejection accident results in April 1996 fuel failure or melting 6.4 GDC-19 Notes 1, 2, Draft Rev. 3 3, 28, 29*

April 1996 NA for BWRs Radiological Consequences of Control Rod Drop Accident BWR EPUs that do not utilize alternative source term whose control rod drop accident results In fuel failure or melting SPSB SRXB 15.4.9, App. A Draft Rev. 3 April 1996 10 CFR 100 Notes 9, 10, 27*

4 4.

9.2 9.2 6.4 Draft Rev. 3 April 1996 GDC-19 Notes 1, 2, 3, 28, 29*

.4 1

Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment EPUs that do not utilize alternative source term whose failure of small lines carrying primary coolant outside containment result in fuel failure SPSB 15.6.2 GDC-55 Draft Rev. 3 10 CFR 100 April 1996 2.7.3 2.7.5 6.4 Draft Rev. 3 April 1996 GDC-19 Notes 1, 2, 3, 28,29*

Radiological Consequences of Steam Generator Tube Failure PWR EPUs that do not utilize alternative source term whose steam generator tube failure results in fuel failure SPSB SRXB 15.6.3 Draft Rev. 3 April 1996 10 CFR 100 Notes 4, 13,14,15, 27*

9 4.

6.4 Draft Rev. 3 April 1996 GDC-19 Notes 1, 2, 3,28, 29*

NA for BWRs 9.2 4

I

4.

.4 9

4.

Radiological Consequences of Main Steamline Failure Outside Containment for a BWR BWR EPUs that do not utilize alternative source term whose main steam line failure outside containment results In fuel failure SPSB SRXB 15.6.4 Draft Rev. 3 April 1996 10 CFR 100 Note 27*

4 4.

6.4 Draft Rev. 3 April 1996 GDC-19 Notes 1, 2, 3, 28, 29*

I J.

I ______________

I J.

I BVY 03.981 PAGE 24 MATRIX 7 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

Areas of Review-Applicable to

~

Primary

- Secondary SRP Section.

- Focus of SRP Other Template Safety Acceptance Review Review.

Number Usage Guidance Evaluation Section Review Branch Branch(es)

Number CPPU SARI CPPU LTR B W R PW R Radiological Consequences of a EPUs that do not utilize SPSB SPLB 15.6.5, App. A 10 CFR 100 Notes 4, 2.7.5 2.7.7 9.2 Design Basis Loss-Of-Coolant-alternative source term Draft Rev. 2 23, 24,25, Accident Including Containment April 1996 26,27*

Leakage Contribution 6.4 GDC-19 Notes 1, 2, Draft Rev. 3 3, 28, 29-April 1996 Radiological Consequences of a EPUs that do not utilize SPSB SPLB 15.6.5, App. B 10 CFR 100 Notes 11, 2.7.5 2:7.7 9.2 Design Basis Loss-Of-Coolant-alternative source term Draft Rev. 2 27*

Accident: Leakage from ESF April 1996 Components Outside Containment 6.4 GDC-19 Notes 1,2, Draft Rev. 3 3,28, 29*

April 1996 Radiological Consequences of a BWR EPUs that do not utilize SPSB 15.6.5, App. D 10 CFR 100 Notes 9, 2.7.5 9.2 Design Basis Loss-Of-Coolant-alternative source term Draft Rev. 2 12, 27*

Accident: Leakage from Main April 1996 Steam Isolation Valves 6.4 GDC-19 Notes 1, 2, Draft Rev. 3 3, 28, 29*

Radiological Consequences of EPUs that do not utilize SPSB SPLB 15.7.4 10 CFR 100 Notes 4, 5, 2.7.6 2.7.8 9.2 Fuel Handling Accidents alternative source term Draft Rev. 2 GDC-61 18,19, 20, April 1996 27*

6.4 GDC-19 Notes 1,2, Draft Rev. 3 3,28, 29*

April 1996 Radiological Consequences of EPUs that do not utilize SPSB EMEB 15.7.5 10 CFR 100 Notes, 5, 2.7.7 2.7.9 NA (VYNPS Spent Fuel Cask Drop alternative source term SPLB Draft Rev. 3 GDC-61 16,17, 8, utilizes Accidents April 1996 18,27' AST) 6.4 GDC-19 Notes 1,2, Draft Rev. 3 3,28, 29' April 1996 BVY 03-981 PAGE 25 MATRIX 7 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 8 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Health Physics Areas of Review a,

Applicable to a

.,..S Primary.

Review

- Branch Secondary SRP Review -

Section:

Branch(es)

Number I.

Focus of SI

Usage RP Other Termpial Guidance Evaluatic Nur

___ ~

BWR te Safety in Section nber Acceptance Review -

CPPU SAR I CPPU LTR I

-I.

Radiation Sources All EPUs IEHB Radiation Protection Design All EPUs IEHB Features Operational Radiation Protection All EPUs IEHB Program 12.2 10 CFR 20 Draft Rev. 3 April 1996 12.3-4 10 CFR 20 Draft Rev. 3 GDC-19 April 1996

PWR 2.8.1 2.8.1 8.3 and 8.4 2.8.1 2.8.1 8.5 12.5 Draft Rev. 3 April 1996 10 CFR 20 Note

• 2.8.1 2.8.1

8.5 Notes

1. Regulatory Guide 8.14 was withdrawn on February 9,2001, and should not be used.

BVY03-981PAGE26 MATRIX 8 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 9 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Human Performance Areas of Review

^

^

Applicable to, I...: ' ........

Primary Secondary.

Review Review-_

Branch,

^Branch(es)

Reactor Operator Training All EPUs IEHB Training for Non-Licensed Plant All EPUs IEHB Staff Operating and Emergency All EPUs IEHB Operating Procedures Human Factors Engineering All EPUs IEHB

'SRP Focus of SRP Section Usage Number 13.2.1 Specific review Draft Rev. 2 questions are Dec. 2002 provided in the template safety evaluations.

13.2.2 Specific review Draft Rev. 2 questions are Dec. 2002 provided in the template safety evaluations.

13.5.2.1 Specific review Draft Rev. 1 questions are Dec. 2002 provided in the template safety evaluations.

18.0 Specific review Draft Rev.

questions are Dec. 2002 provided in the template safety evaluations.

Other '

Template Safety Guidance Evaluation Section Number-

, ^ BWR PWR Acceptance

- Review CPPU SAR /

CPPU LTR 2.9 2.9 10.6 2.9 2.9 10.6 2.9 2.9 10.9 2.9 2.9 10.6 SPLB SRXB BVY 03-98 / PAGE 27 MATRIX 9 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 10 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Power Ascension and Testing Plan Areas of Review......... -. .

M Applicable to :

Primary I Revlew:I.

Branch

.Secondary:

Review Branch(es)..

SRP Section

NumberE Focus of SRP

-

Other Template Safety.

Usage Guidance Evaluation Section Number BWR 1:PWR Acceptance Review.

CPPU SAR /

CPPU LTR Power Ascension and Testing All EPUs IEHB EEIB 14.2.1 Entire Section EMCB Draft Rev. 0 EMEB Dec. 2002 SPLB SPSB SRXB 2.10 2.10 10.4 BVY03-98/PAGE28 MATRIX 10 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002

MATRIX 1 1 SCOPE AND ASSOCIATED TECHNICAL REVIEW GUIDANCE Risk Evaluation

4.

Areas of Review.

Applicable to.

v Primary Secndary SRP Focus of SRP Other Review..-

Review.,

Section Usage Guidance Branch..

Branch(es)

- Number Template Safety Evaluation Section'

- Number -

Acceptance Review CPPU SAR /

CPPU LTR:

BWR

- PWR Risk Evaluation All EPUs SPSB Note 1*

2.11 2.11 10.5 RG 1.174 RIS 2001-02 Notes:

1. The staff's review is based on Attachment 2 to this matrix. Attachment 2 invokes SRP Chapter 19, Appendix D, If special circumstances are identified during the review.

BVY 03-98 PAGE 29 MATRIX 11 OF SECTION 2.1 OF RS-001 (DRAFT)

DECEMBER 2002