ML18141A105

Application for Amend to Licenses DPR-32 & DPR-37,reducing Boric Acid Concentration.Supporting Safety Evaluation Encl

ML18141A105
Person / Time
Site:	Surry
Issue date:	09/13/1983
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Varga S Office of Nuclear Reactor Regulation
Shared Package
ML18141A106	List: ML18141A105 ML18141A107
References
521, NUDOCS 8309190078
Download: ML18141A105 (48)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 W. L. STEWART VICE PRESIDE]'fT September 13, 1983 NUCLEAR OPERATIONS Mr. Harold R. Denton, Director Serial No. 521 Office of Nuclear Reactor Regulation PSE/NAS/cdk:0008N Attn: Mr. Steven A. Varga, Chief Docket Nos. 50-280 Operating Reactors Branch No. 50-281 Division of Licensing License Nos. DPR-32 U. S. Nuclear Regulatory Commission DPR-37 Washington, D. C. 20555 Gentlemen:

AMENDMENT TO OPERATING LICENSES DPR-32 AND DPR-37 SURRY POWER STATION UNITS l AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGE Pursuant to 10CFR 50.90, the Virginia Electric and Power Company requests an Amendment, in the form of changes to the Technical Specifications, to Operating *Licenses DPR~32 and DPR-37 for the Surry Power Station, Units No. 1 and 2, respectively. The proposed changes and the supporting safety evaluation are enclosed.

There are many problems associated with the high boric acid concentration which must be maintained in the boron injection tanks and concentrated boric acid system. During normal operation reactor coolant letdown is concentrated and recycled to the boric acid tanks via the Boron Recovery System. As a result the systems attain high radiation levels. These high radiation levels compound the maintenance problems caused by the high boric acid concentrations. For example, boric acid is a highly corrosive fluid and leakage has led to the degradation of carbon steel components. Leakage has also led to the failure of heat tracing which is required to maintain solution solubility. Failure of heat tracing results in additional maintenance

• problems such as boron plateout and potential line plugging as the solution temperature drops. The increased maintenance causes increased personnel radiation exposures. A reduction in boric acid concentration would reduce maintenance requirements and the associated exposures to plant personnel.

Attachment 1 provides the detailed justification for a proposed reduction in the minimum boron injection tank (BIT) concentration from 11 .5 wt% to O wt

% and a change in the minimum boric acid system concentration from 11 .5 wt%

to 7.0 wt%. A general description of the current design of the Boron Injection Tank and Boric Acid System is given. The proposed physical changes to each system are described and operational and maintenance benefits are discussed.

8309190078. 8309t~i PDR ADOCK 05000280 P PDR

YlllGINIA. ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton e

Analyses have been performed to determine the impact of the proposed changes on the appropriate Surry licensing bases, including a reanalysis of the steamline break accidents discussed in Chapter 14 of the Surry Updated Final Safety Analysis Report (UFSAR). The analysis has been performed by Vepco, using the RETRAN Computer Code and the reactor system transient analysis methodology described in our topical report which was transmitted by letter, dated April 14, 1981 (Serial No. 215). The methodology, assumptions and results of the analysis are discussed in detail in Attachment l. This documentation will be incorporated into the Surry Updated FSAR during the next annual update.

The changes to the Technical Specifications associated with the proposed boron concentration reduction are provided in Attachment 2. The proposed changes and the safety evaluation have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff. It has been determined that this request does not pose a significant hazards consideration as defined in 10CFR 50.92 or an unreviewed safety question as defined in 10CFR 50.59.

We have evaluated this request in accordance with the criteria in 10CFR 170.22. Since this request involves a safety issue which the staff should be able to determine does not involve a significant hazards consideration for Unit l ~nd a duplicate safety issue for Unit 2, a Class III license amendment fee and a Class I license amendment fee are required for Unit l and Unit 2, respectively. Accordingly, a voucher check in the amount of $4,400.00 is enclosed in payment of the required fee.

Inasmuch as the proposed changes will .result in significant operational benefits and reduction in personnel exposure, we solicit your expeditious review and approval by December 15, 1983. We are interested in meeting* with you during the month of September, to discuss the details of this proposed change and the supporting safety analysis.

Very truly yours,,

/;!ft~

W. L. Stewart

Enclosures:

(1) Safety Evaluation for Proposed Changes (2) Proposed Technical Specifications Changes (3) Voucher Check $4,400.

cc: Mr. James P. O'Reilly Mr. J. Don Neighbors Regional Administrator NRC Project Manager - Surry Region II Operating Reactors Branch No.

Division of Licensing Mr. D. J. Burke NRC Resident Inspector Mr. Charles Price Surry Power Station

• Department of Health 109 Governor Street Richmond, Virginia

'cOMMONWEALTH OF VIRGINIA)

)

CITY OF RICHMOND )

The foregoing document was acknowledged before me, in and for the City and Commonweal th aforesaid, today by W. R. Cartwright who is Manager-Nuclear Operations Support, of the Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this Jr day of ~ ~ "'~ 19 S' -3 My Commission expires: -< -

/

}-fD , 19 h~

s Notary Public (SEAL)

S/001

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ATTACHMENT 1 SAFETY EVALUATION FOR REDUCTION IM BOROM CONCENTRATION IH THE BOROM IHJECTIOH TAHK AHD CONCENTRATED BORIC ACID SYSTEM SURRY POWER STATION

- UNITS 1 AND 2

e e PAGE 1 A. INTRODUCTION

1. Objective Surry Power Station has experienced problems associated with the high boric acid concentrations which must be maintained in the Bo~on Injection Tanks and concent~ated Boric Acid System.

The highly concentrated acid is required to be heated to a minimum Technical Specifications limit of 145 Fin order to maintain solution solubility. Numerous failures of the heat tracing circuitry have occurred throughout the life of the plant. These heat tracing failures lead to line plugging and restrictions as the solution temperature decreases and boric acid plateout occurs. These problems necessitate extensive maintenance and operational attention which causes excessive radiation exposure to plant personnel.

Vepco has been evaluating different methods for alleviating these problems. The following sections decribe the design functions of the Boron Injection Tank and the concentrated Boric Acid System and provide the justification for reducing the minimum boron concentration requirements in order to greatly reduce these maintenance problems and thus the associated personnel radiation exposure. The proposed reduction consists of a change in the minimum Boron Injection Tank CBIT) concentration from 11.SX to OX and a change in the minimum Boric Acid System concentration from 11.5~ to 7X. The reduction in BIT concentration can be achieved by taking credit for the

• e PAGE 2 Integral Flow Restrictors in the sa£ety analysis 0£ the main steam line break accident; this accident is discussed in Chapter 14 0£ the Surry Updated Final Sa£ety Analysis Report CUFSAR). The Integral Flow Restrictors were installed during the steam generator repair outage. The reduction in boric acid system concentration can be accomplished by increasing the minimum allowable Boric Acid Tank inventory associated with each unit £rom 4200 gallons to 6000 gallons, thereby preserving the capability £or cold sa£e shutdown at any time in li£e with the most reactive control rod assembly withdrawn £rem the core.

Section A.2 provides a general description 0£ the current design 0£ the Boron Injection Tank and Boric Acid System and describes the proposed physical changes to each system; operational and maintenance benefits 0£ the proposed changes are discussed in Section A.3.

Analyses have been per£ormed to determine the impact of the

'proposed changes on the appropriate Surry licensing bases. A boron concentration reduction in the BIT a£fects only the steamline break transient results. A detailed discussion 0£ the supporting analyses per£ormed £or this transient is provided in section B of this attachment. The proposed boron concentration reduction in the BAT does not impact any of the accident analyses presented in Chapter 14 0£ the UFSAR.

Section C presents an evaluation of the impact of the proposed plant modifications on plant operations and the results 0£ a

e e PAGE 3 review of the FSAR.

A compilation of the required Technical Specification changes to implement the proposed concentration reductions is presented as a separate attachment.

e e PAGE A.2 Changes to Current System Operations BORON IHJECTIOH TANK The Boron Injection Tank CBIT) is a 900 gallon carbon steel tank which is internally clad with stainless steel and is part of the Safety Injection System; it contains boric acid solution at a minimum of 11.5~ by weight boric acid. Redundant tank heaters and line heat tracing are provided to maintain a minimum solution temperature at a Technical Specifications limit of 145 degrees F, thus preventing boron plateout.

Recirculation from the BIT to the Boric Acid Tanks is maintained continuously via a Boric Acid Transfer Pump to ensure the BIT is full of concentrated boric acid at all times and to p~event cold spots and stratification within the tank.

The BIT is isolated from the Reactor Coolant System and the Charging Pumps during normal plant operation by two sets of parallel isolation valves. Figure 1 illustrates the system design as described above.

The purpose of the BIT is to provide injection of highly concentrated boric acid to the Reactor Coolant System to mitigate the reactivity addition resulting from a main steam line break accident. Operation of the BIT, which takes place upon actuation of a Safety Injection Signa1. does not impact any of the accident analysis results presented in Chapter 14 of the Updated Final Safety Analysis Report other than the steamline break.

--**. ....... * --~-- - * * * ** * .- *-*--- ** 1 e e PAGE 5 During Safety Injection, the suction of the high head safety injection/charging pumps is diverted from the normal suction at the Volume Control Tank CVCT) to the Refueling Water Storage Tank (RWST). The Safety Injection £low path through the BIT is established by the opening of redundant parallel isolation valves upon a Safety Injection signal. Concurrent with the opening of the BIT isolation valves is the closing of redundant isolation valves in the recirculation line to the BAT. Flow from the safety injection/cha%ging pumps is introduced into the BIT via a sparger internal to the vessel. This sparger is designed to disperse the fluid and create a front, prevent channeling and therby cause slug flow to pass through the tank and into the Reactor Coolant System.

Current plans are to. reduce the required boric acid concentration in the BIT to OX with only a minor physical modification to the plant. This modification would entail (1) cutting the recirculation lines (to the BAT and Volume Control Tank) and welding the ends closed to ensure concentrated boric acid will not leak from the Boric Acid Tank to the BIT and C2) removal of electrical power to the recirculation line isolation valves. Electrical power will also be terminated to the BIT heaters and heat tracing circuits of the recirculation lines.

The BIT would remain in place. All ~ther system components will remain unaffected, including the actuation of the BIT inl~t and outlet isolation valves.

• - ~ .. __ ...

e PAGE 6 Vepco is evaluating the £easibility 0£ physically removing the BIT and associated BIT/BAT recirculation piping at some £uture date.

BORIC ACID SYSTEM The concentrated boric acid system is a part 0£ the Chemical and Volume Control System described in Section 9.1 0£ the Updated Final Sa£ety Analysis Report. The purpose 0£ the system is to provide an inventory of concentrated boric acid £or C1) chemical shim reactivity control, C2) providing makeup to the Reactor Coolant System, Refueling Water Storage Tank, spent fuel pit and refueling cavity as necessary and (3) recirculation 0£ boric acid through the BIT via the boric acid transfer pumps. The system consists of three Boric Acid Tanks, four boric acid transfer pumps, one batch tank, boric acid filters and associated piping, valves, heat tracing, controls and instrumentation. The Boric Acid Tanks are sized to provide sufficient boric acid to bring the reactors to cold shutdown conditions assuming a stuck control rod. A simplified schematic of the system is shown in Figure 2.

The three Boric Acid Tanks (BAT) are 7500 gallon stainless steel tanks ~hich are designed for atmospheric pressure. They serve as the reservoirs for boric acid inventory; the three tanks serve both units. A boric acid solution of 11.5~ to 13~

by weight is maintained at all times. The upper concentration

e e PAGE 7 limit of 13~ is established to ensure concentrations low enough to remain soluble at a 145 degrees F minimum temperature, which is maintained by redundant tank immersion heaters and line heat tracing.

During normal operation, boric acid is supplied to each BAT from the Boric Acid Batching Tank via the Boric Acid Transfer Pumps or from the Boron Recovery System to maintain a minimum level of 4200 gallons dedicated to each unit.

Reduction in the boron concentration requirement for the Boric Acid system to 7~ will require increasing the minimum volume of boric acid stored £or each operating unit to 6000 gallons. This can be accomplished by resetting the existing level instrumentation and alarms £or the new minimum low level. With this increase in volume requirement for each unit, the capability to bring the units to cold shutdown conditions with the most reactive control rod assembly withdrawn from the core is preserved. BAT heater controls and the sy~tem heat tracing controls will be reset to maintain a minimum temperature of 112 degrees F to maintain solution solubility.

In summary, the proposed Boric Acid System changes will require the following plant modifications:

1. Reset Boric Acid Tank level instrumentation and alarms for a minimim volume of 6000 gallons.

2. Reset BAT heater and heat tracing controls to maintain a minimum solution temperature of 112

e e PAGE 8 deg:tees F.

e e PAGE 9 A.3 Benefits The high boric acid concentrations in the boron injection tank and the concentrated boric acid system is causing numerous maintenance problems, resulting in increased radiation exposures to plant personnel. The concentrated boric acid is a corrosive fluid. Leakage from the systems has been linked to (1) corrosion of carbon steel components and supports and C2) failure of heat tracing equipment.

The lowering of the minimum required boric acid concentration in the BIT to zero CO) ppm C1) reduces the potential £or degradation of carbon steel components and supports due to leakage, C2) eliminates the need to maintain BIT heaters and heat tracing on the associated safety injection piping and recirculation lines and (3) eliminates the need for periodic checks of the BIT concentration, thereby reducing radiation exposures £or plant personnel. The reduction in boron concentration in the BIT will also reduce the RCS dilution required £or a retuin to power in the event of an inadvertant Safety Injection. This will reduce the amount of letdown which must he processed by the Boron Recovery and Waste Handling systems.

Reducing the minimum required boric acid concentration £or the concentrated boric acid system will improve heat tracing system performance, which in turn decreases the potential for system line blockage due to boron plateout. This will increase system

e e PAGE 10 zeliability, zeduce maintenance zeguizements and theze£oze pezsonnel zadiation exposuze.

~n summazy, the pzoposed zeductions in BIT and Bozic Acid System concentzations o£fez significant benefits to Vepco in tezms 0£ inczeased opezational zeiiability, zeduced maintenance costs and deczeased pezsonnel zadiation exposuzes.

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e e PAGE 11 B. ACCIDENT ANALYSIS -RUPTURE OF A MAIM STEAM PIPE ASSUMING 0 PPM BOROM IH THE BOROM IHJECTIOH TANK

1. INTRODUCTION AND BACKGROUND A ~upture of a main steam pipe is assumed to include any accident which results-in an uncontrolled steam release from a steam generator. The release can occur due to a break in a pipe line or due to a valve malfunction. The steam release results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the Reactor Coolant System causes a reduction of reactor coolant temperature and pressure. With a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin. If the most reactive control rod assembly is assumed stuck in its fully withdrawn position, there is a possibility that the core will become critical and return to power even with the remaining control rod assemblies inserted.

A* %eturn to power following a main steam pipe rupture is a potential problem mainly because of the high hot channel factors which exist when the most %~active rod is assumed stuck in its fully withdrawn position. Assuming the worst combination of cii:cumstances which could lead to :cesumption of powe:r gene:ratJon following a main steam line b:reak, the core is ultimately shut down by the bo:ric acid in the Safety Injection System.

The analysis of a main steam pipe :rupture is perfo:rmed to

_* ___ ... **-----~ -----*--- _. .._,;.._ **-*-*- .. - .

e e PAGE 12 demonstrate that even with a boron concentration of zero in the Boron Injection Tank CBIT):

a. Assuming a stuck control rod assembly with or without offsite power, and assuming a single failure in the engineered safety features there is no consequential damage to the primary system and the core remains in place and intact.

b. With no stuck control rod assembly, and all equipment operating at design capacity, insignificiant Cor no) cladding rupture occurs.

c. There will be no DNB or clad perforation resulting from any single active failure of the main steam sy~tem. The single active failure is the opening,.

with failure to close, of the largest of any single steam bypass, relief or safety valve.

Although DHB and possible clad perforation following a main steam pipe rupture are not necessarily unacceptable, the following analysis shows that no DNB occurs for any rupture assuming the most reactive control rod assembly stuck in its fully withdrawn position.

The following systems provide the necessary protection against a main steam pipe rupture=

a~ Safety Injection System actuation from any

PAGE 13 of the following*:

(1). Two out of thzee pzessuzizez low pzessuze

_signals.

CZ). Two out of thzee diffezential pzessuze signals between any main steam line and the main steam headez.

(3). High steam flow in two out of thzee main steam lines Cone out of two pez line) in coincidence with eithez low Reactoz Coolant System avezage tempezatuze (two out of thzee) oz low main steam line pzessuze Ctwo out of thzee).

(4). Thzee out of fouz high containment pzessuze signals.

b. The ovezpowez zeactoz tzips Cneutzon flux and AT) and the zeactoz tzip occuzzing upon actuation of the.Safety Injection System.

• The details of the logic used to actuate Sa_fety Injection *aze discussed in Section 7 0£ the FSAR.

e e PAGE 14

c. Redundant isolation of the steam genezato:r feedwatez lines. Sustained high feedwate:r flow would cause additional cooldown, thus, in add~tion to the no:rmal contzol action which closes the main feedwate:r valves, any safety injection signal iapidly closes all feedwate:r contzol valves, tzips the steam generator feedwate:r pumps, and closes the ieedwatez pump discharge valves.

d. T:rip of the fast acting main steam line t:rip valves (designed to close in less than 5 seconds) on:

(1). High steam flow in two out of three main steam lines Cone out of two per line> in coincidence with eithez low Reactoz Coolant System average temperature (two out of three> o:r low steam line pressuze Ctwo out of thzee).

(2). Th:ree out of four high containment pressuz:e signals.

Each main steam line has a fast closing trip valve and a non-return valve. These valves prevent blowdown of mo:re than one steam genezator £oz: any bzeak location even i£ one valve fails to close. Foz: example, for a bzeak upstz:eam of the tzip valve in one line, closuze of either the non-return valve in that line or the tzip valves in the other lines prevent blowdown of the other steam generatozs.

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e PAGE 15 All Surry steam generators are equipped with integral £low restrictors at the generator outlet. The restrictors have a smaller £low area than the main pipe and serve to reduce the largest e££ective break area which must be considered to 1.4 squa:z::e :feet.

e e PAGE 16

2. METHOD OF ANALYSIS The analysis of the main steam pipe rupture has been per-formed to determine:

a. The core heat flux and Reactor Coolant System tempera-ture and pressure resulting from the cooldown following the steam line break. The analysis was performed with the RETRAM computer code. The calcula-tion describes the plant neutron kinetics, the reactor coolant system including natural circulation, the pressuri2er, steam generators and feedwater system.

The digital program computes pertinent variables in-cluding the bzeak flow rate, core power and point kinetics reactivity and primary coolant temperatures.

h. The thermal and hydraulic behavior of the core following the steam line break. A detailed COBRA core thermal and hydraulic digital computer calcul-ation has been used to determine if DNB occurs for the core conditions computed in (1) above. This calculation solves the continuity, momentum and energy equations of fluid flow in the core and with the W-3 correlation (See reference in Paragraph g below) determines the margin to DMB.

The following assumptions were made:

a. A 1.77~ shutdown reactivity from all but one control

• e PAGE 17 rod assembly at no load conditions. This is the end 0£ life design value including design margins with the most reactive control rod assembly stuck in its fully withdrawn position. The actual shutdown capa-bility is expected to be significiantly greater.

b. A negative moderator coefficient representative of end of life core conditions with all but the most re-active control rod assembly inserted. The variation of the coefficient with temperature and pressure has been included. The reactivity versus temperature corresponding to the negative moderator coefficient used is shown in Figure 1. In computing the power generation following a steam line break, the local reactivity £eedback from the high neutron £lux in the region of the core near the stuck control rod assembly has been included in the overall reactivity balance. The local reactivity feedback is composed of Doppler reactivity from the high fuel temperatures near the stuck control rod assembly and moderator £eedback from the high water temperature near the stuck control rod assembly. For the cases analyzed where steam generation occurs in the high flux regions of the core, the effect of void forma-tion on the reactivity has also been included. The effect of power generation in the core on overall re-activity is shown in Figure 2 The curve

• PAGE 18 assumes end 0£ li£e core conditions with all control rod assemblies in except the most reactive control rod assembly which is assumed stuck in its £ullly withdrawn position (completely removed £zom core).

c. Minimum sa£ety injection capability cozzesponding to the opezation 0£ only one high head safety injection pump. A bozon concentation 0£ 2000 ppm was assumed in the Re£ueling Watez Stozage Tank, fzom which the safety injection pumps take suction. The initial boron concentzation in the Bozon Injection Tank (BIT) and the associated safety injection piping is assumed to be zero. The time-delays zequized to sweep this unbozated watez fzom the piping pzioz to the delivezy of the 2000 ppm bozon have been included in the analysis.

d. A conservatively high steam genezator heat tzansfez coefficient CUA) which is zepzesentative of nucleate boiling in the secondary side of the generatoz thzough out the tzansient. This is conse~vative since no allowance £oz zeduction of the heat transfer UA as the watez level falls into the tube region has been made. The effects of main feedwatez flow Cpzior to feedline isolation) and auxiliary feedwatez flow have been included in the analysis.

e. Hot channel factozs cozzesponding to one stuck contzol

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e PAGE 19 xod assembly -- the contxol xod assembly giving the highest factox at end of life. The hot channel factoxs account fox the void eKisting local to the stuck contxol xod assembly at the pxessuxe that occurs during the return to powex phase following the steam break. This void in conjunction with the large nega-tive* modexatox coefficient paxtially offsets the effect of the stuck control rod assembly. The hot channel factors depend up~n the ~ore temperature, pressure and flow and axe therefore different fox each case studied. The calculations used to obtain the hot channel factors again assume end of life core conditions with all control rod assemblies in except the most reactive control rod assembly.

f. Three combinations of break sizes and initial unit conditions have been considered in determining the core power and Reactor Coolant System transient.

Ca>. Complete sevexence of a main steam pipe, initially at no load conditions with outside powex available. The presence of the integxal flow restxictoxs in the steam genexatoxs will control the steam release xates*fox all *bxeak locations, both inside and outside the containment Cb). Case (a) above with loss of outside power simultaneous with the steam break.

- -** - _, -***- -- *- - ---- -- : ..i.--~*> _;;._ **-

e e PAGE 20 Cc). A break e~uivalent to a steam flow of 2~7 lbs per second at 1100 psia from one steam generator with outside power available. This is larger than or equal to the capacity of any single dump or safety valve.

All the cases above assume initial hot shutdown conditions with the control rod assemblies inserted (except for one stuck control rod assembly) at time zero. Should the reactor be just critical or operating at power at the time of a main steam line break, the reactor is tripped by the normal overpower protect-ion system when the power level reaches a trip point. Following a trip at power the Reactor Coolant System contains more stored energy than at no load, the average coolant temperature is higher than at no load and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the main steam line break before the no load conditions of Reactor Coolant System temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has been removed, the cooldown.

and reactivity insertions proceed in the same manner as in the

• analysis which assumes a no-load condition at time zero.

However, since the initial steam generator mass is greatest at no load, the magnitude and duration of the Reactor Coolant System cooldown are less for main steam line breaks occurring at power.

e e PAGE 21

g. In determination of the critical flux at which burnout could occur the W-3 correlation was used .. It was considered to be the correlation which most accurately represents the range of parameters produced in the transients analyzed.

h. In computing the steam flow during a steam line break, the Moody critical flow model (Moody, F. J., "Maximum Flow Rate of a Single Component, Two-Phase Mixture," Journal of Heat Transfer, 93, pp 179-187, 1965) was used.

• PAGE 22

3. Results The results presented are a conservative indication of the events which would occur following a main steam line rupture.

The worst case assumes that all of the following occur simultaneously.

(1). Minimum shutdown reactivity margin of 1.77X.

(2). An end-of-life, rodded core moderator temperature coefficient; use of end-of-life conditions maximizes the positive reactivity insertion resulting from cooldown.

(3). The highest worth control rod assembly stuck in its fully withdrawn position.

(4). The single most restrictive failure of the Engineered Safety Features.

A. Core Power and Reactor Coolant System.Transient Figures 3-4 show the Reactor Coolant System transient and core heat flux following a main steam pipe rupture (complete severence of a pipe) at in~tial no load conditions (Case A).

The break assumed is the largest break which can occur anywhere in the system. outside power is-assumed available such that full reactor coolant flow exists. The transient shown assumes the control rod assemblies inserted at time O (with one control rod assembly stuck in its fully withdrawn position> and steam

.- -- ". ~*- .* ~ -. --*.

e e PAGE 23 release from only one steam generator. Should the core be critical at near zero power when the rupture occurs the initiation 0£ safety injection by high di££erential pressure between any steam generator and the main steam header or by high steam £low signals in coincidence with either low Reactor Coolant System temperature or low steam line pressure trips the reactor. Steam release from at least two steam generators is prevented by either the non-return valves or by automatic trip 0£ the fast acting trip valves in the steam lines by the high steam flow signals in coincidence with either low Reactor Coolant System temperature or low main steam line pressure.

Even with the failure 0£ one valve, release is limited to no more than 5 seconds £or two steam generators while the third generator blows down. The main steam line trip valves are designed to be fully closed in less than 5 seconds with no £low through them. With the high £low existing during a main steam line rupture, the valves will close considerably £aster since closure is £low assisted.

As shown in Figure 3, the core attains criticality with the control rod assemblies inserted (with the design shutdown ma%gin, assuming one stuck control rod assembly) at 22 seconds.

Boron solution at 2000 ppm enters the Reactor Coolant System from the Safety injection System, is diluted and mixed with RCS water and reaches the core at 231 seconds. This reflects a delay 0£ 224.5 seconds to purge the boron injection tank and associated piping 0£ 0 ppm water, and 3.5 seconds to transport

e PAGE 2~

dilute boron from the injection point in the cold legs to the core. The calculation also accounts for a 3 second delay to receive and actuate the safety injection signal and 10 seconds to compietely open vaive trains in the safety injection 1ines.

Since the safety injection pump accelerates to full speed in less than the time required to open the valve tzain. and since

- it is expected that the safety injection signal will he generated in less than 3 seconds. the overall delay time is considered conservative.

The calculation assumes the boric acid is mixed with and diluted by the water flowing in the Reactor Coolant System priot to entering the reactor core. The concentration after mixing depends upon the relative flow rates in the R~actor Coolant System and in the Safety Injection System. The variation of mass flow rate in the Reactor Coolant System due to water density changes is included in the calculation as is the variation of flow rate in the Safety Injection System due to changes in the Reactor Coolant System pressure. The Safety Injectioti System flow calculation includes the line losses in the system as well as the pump head curve.

Ko credit has been taken for any boron in the BIT or safety inject1on lines which enters the Reactor Coolant System prior to the 2000 ppm boric acid from the Refueling water Sto%age Tank. The heat flux achieved was 23.7~ of the value at 2~~1 MWt Ca summary of conditions is listed in Table 1>.

e PAGE 25 Figures 5-7 show the responses for the previous break except with a loss 0£ outside power (Case Bl. Reactor Coolant system flow coastdown is assumed to occur simultaneously with the break. The Safety Injection System delay time includes the time required to start safety injection pumps with emergency power from the diesel generators. Only one safety injection pump is assumed. Criticality is attained at 31 seconds and the peak heat flux is 8.1~ 0£ the values at 2441 MWt. A summary of the time sequence £or the above case is given in Table 1.

Figures 8-9 show the transient following a break equivalent to a steam £low of 247 lbs per sec. at 1100 psia with steam release from one steam generator (Case C). The assumed steam release is larger than or equal to the capacity of any single dump or safety valve.* In this case, safety injection is initiated automatically by low pressurizer pressure at 140 seconds. Operation of one safety injection pump is considered, since this will provide the most conservative results. Criticality is attained at 280 seconds; dilute boron solution reaches the core at 370 seconds and limits the peak heat flux to 4.06~ of the value at 2441 mwt. The cooldown £ox the case shown in Figures 8-9 is more rapid than the case 0£ steam release from all steam generators through one relief, bypass ox safety valve. The transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel

e e PAGE 26 elements ox the energy stored in the_othex steam genexators.

Since the transient occurs over a period of about five minutes, the neglected stoxed

• energy is likely to have a significant effect in slowing the cooldown.

It should he noted that following a main steam line break only one steam generator blows down completely. Thus, two steam generators are still available for dissipation of decay heat after the initial transient is over. In the case of loss of outside power this heat is removed to the atmosphere and the atmospheric safety valves have been sized to cover this condition.

B. Margin to Critical Heat Flux Using the. transients shown in Figures 3 through 9 the Westinghouse W-3 correlations was used in conjunction with the Vepco version of the COBRA core thermal hydxaulics code to determine the margin to DNB. Carefully chosen points from each transient were examined and the results are presented in Table

2. The power and flow conditions are shown together with pressure and core inlet temperatures. It was found that all cases had a minimum DNBR greater than 1.30.

.e e PAGE . 27 TIME SE2UENCE OF EVENTS FOR MAJOR SECONDARY SYSTEM PIPE RUPTURE TABLE 1 ACCIDENT EVENT TIMECSEC.)

Major: Secondary System Pipe Rupture

1. Case a Steam line ruptures 0

~r:essur:i2er: empty 11 Criticality attained 22 Dilute boron reaches core 231

2. Case b Steam line ruptures 0 Pressurizer: empty 12 Criticality attained 31 Dilute boron reaches core 245

4. Case c Steam line ruptures 0 Pressuii2er empty 100 Criticality attained 280 Dilute boron reaches core 370

e e PAGE 28 STEAMBREAK ACCIDENT STATEPOIHTS TABLE 2 Hypothetical Break Credible Break With Power Without Power With PoweJ:

Case A Case B Case C Co:r:e Heat 23.7 8. 1 4. 1 FlUK, ". of 2441 11W'l' RCS Pressure, 959 853 733 psia Loop A Inlet 398 276 460 Temp, °F Loop B Inlet 469 497 482 Temp. °F Core Boron .. 0. 0 0.4 8.2 Concentration, PPM RCS Flow, 100 6.4 100 Reactivity, .007 .003 .026

,,. deltaK/K Time, sec. 201 250 395 DHBR >1.30 >1. 30 >1. 30

Multiplication 6k, %

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%OF 2441 MWt VARIATION OF REACTIVITY WITH POWER AT CONSTANT CORE AVERAGE TEMPERATURE. VALUES INDICATED ,. I' WERE USED IN STEAM PIPE RUPTURE ANALYSIS FOR THE END OF LIFE RODDED CORE WITH ONE CONTROL ROD ASSEMBLY STUCK

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• e PAGE 29 C. OPERATZOHS~FSAR REVZEW

1. Opezations Zapact An evaluation of effects on plant opezations *wa~.aade to detezmine all positive oz negative implications of zeducing hozon concentzation in the Bozic Acid Tank and the Bo%on Znjection Tank. A summazy of those changes is listed below:

a. The zeduction of the Bozic Acid Tank concentxation and the inczease in tank volume was evaluated mnd the folloHing opezations we%e found to be impacted:

(1). Znc%ease in time to boxate undex noxmal and eme%gency ope%ating conditions~

Station Cuzve Book nomogxaphs fox hozon addition will be zevised fox the decxeased minimum concentxation of 7X. All increases in times zequized to bozate ox makeup wexe found to be satisfactozy

£zom a plant opexational standpoint.

(2). Znczeased Bozic Acid Tank volume-

. The inc%ease in BAT volume to 6000 gallons (pxeviously 4200) was evaluated zelevant to ove%flow considezations.

Zt was aetezmined that sufficient tank capacity is available to zeplenish the tank in standazd batch volumes without

. -,:---. *- -*----*-*-*- __ ,._r._ *.._ ,--~--. ***-*

PAGE 30 ovexflowing the tank.

(3). Setpoint. chemistxy. and opexating pxoceduxe changes have been identified and will he zevised in accozdance with

&ppzoved pzccedures.

h. The zeduction of the minimum boxic acid concentzation in the Bozon %njection Tank (BIT) to 0~ bozon concentxation was evaluated and found to effect the following:

(1). Recizculation of the BIT will no longex be xequixed to maintain BIT concentxa-tion.

<2>. Pexiodic sampling of B%T concentzation will no longex be xequized.

2. UFSAR REVIEW The Updated Final Safety Analysis Repozt for SU%%Y has been zeviewed fox zequized changes as a xesult of the bozon concentzation reduction. Sevezal axeas will zequixe zevision due to these changes. Upon appzoval of this submittal. these changes will be submitted with the nozmal yeazly UFSAR update.

Foz example. a change xequired to Section 9.1.1.2-Chemical and Volume Contzol- is the time requixed to shut _the reactor down (i.e .* to hot shutdown> with no zods inserted. This condition

..:.: - _: _: -* _::: *- -' ~ _;*:* __; .::... ... _-_

e PAGE 31 has been analyzed to show that the increase in time zequired to shut down (from approximately 15 minutes to appzoximately 25 ainutes> is acceptable.

An evaluation of the time zequized to hozate to cold shutdown conditions with the zeduced boric acid coneent~ation has elso been performed. The results of the evaluation show that. for conservative worst-case conditions, the boric acid volume req*uired to borate to cold shutdown conditions will be less than 6000 gallons, and the associated boration times will be less than 100 minutes. This is well within the requirements of the most restrictive action statement in the Technical Specifications (i.e ** he in cold shutdown within 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />s>.

The evaluation was based on a review of design data for several recent Surry core reloads. Significant conservatisms in the analysis include the followings

a. The initial condition assumes the peak xenon concent~ation which would occur following a zeactor trip from full power; the final condition is assumed to correspond to no xenon.

b. The total shutdown margin available £ollowing zeactor trip is assumed to correspond to the Technical Specifications limit; for most reload cores the available margin is significantly higher.

c. The moderator temperature defect is based on end of life core conditions, where it will be largest.

• ------*- .. ~_.*~._, __ -- *--

. -:-~ ---*-*-----:

-4. - . *-*-*---*--***- . -~;..

PAGE 32 The calculated zeacitvity zequizements weze inczeased by an unceztainty factoz of 20X to cover calculational uncertainties and cycle-to-cycle vaziations in the zequirements.

e. A volumetric bozon mixing model is used which consezvatively neglects the fact that the hozic acid is being intxoduced at coldez tempezatuzes and therefoze highez densities than exist in the Reactoz Coolant System.

An evaluation of the maximum zeactivity insertion due to boron dilution zesulting fzom inadveztent dischaxge of the BIT to the Reactor Coolant System has been pezfozmed. The results show that inadvertent cxiticality cannot result fxom such a discharge at either zefueling oz cold shutdown conditions. At hot shutdown or at-power conditions. the maximum reactivity insertion rates realized fzom such a discharge are well within the zange considezed in the zod withdrawal analyses of Chapter 14 of the UFSAR.

e -- PAGE 33 D. COHCLUS:rOHS The zeduction of minimum bo:ic acid concen~zation zequ1zements fzom 11.5X to OY. in the Bozon :rnjection Tank and fzoa 11.5X to 7.0X in the Concentzated Bozic Acid System at Suzzy Powe:

Station offe:s signifieant benefits to Vi:ginia Elect~ic &nd Powez Company. These benefits include inczeased opezational zeliability, zeduced maintenance costs and deczeased pezsonnel zadiation exposuze.

Additionally. a detailed opezational zeview was conducted; i t has been concluded that the plant can continue to be opezated in a safe and efficient manner following the change.

Analyses of the Hain Steam Pipe Ruptuze have been pezfozmed to demonstzate that the conclusions zeached in Chaptez 14 of the Updated Final Safety Analysis zepozt will not be impacted by the change. As such. the change will not intzoduce any un:eviewed safety questions as defined in 10 CFR 50.59. Also.

the change_ does not involve a significant hazazds considezation. The:e*is a zelaxation in the limiting condition foz opezation; howeveE. a commensuzate level of safety is maintained by the pzesence of the steam genezatoz integzal flow zest.rictozs, thus,* allowing ze:ico ppm concent:ation of bozon in the bozon injection tank.