ML20196F036

Application for Amend to License NPF-42,revising Tech Spec 3.2.3,Figure 3.2-3,Table 2.2-1 & Corresponding Bases to Reflect Revised RCS Temp Measurement Uncertainties.Fee Paid

ML20196F036
Person / Time
Site:	Wolf Creek
Issue date:	06/24/1988
From:	Bailey J WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20196F040	List: ML20196F036 ML20196F043
References
ET-88-0086, ET-88-86, NUDOCS 8807050027
Download: ML20196F036 (18)

Text

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W@NUCLEAN LF CREEK OPERATING C John A. Bailey Vee President Engmeonng and Technscal Serv 6cv June 24, 1988 ET 88-0036 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington, D. C. 20555

Reference:

WM 87-0254, dated October 2, 1988 from B. D. Withers, to NRC

Subject:

Docket No. 50-482: Cycle 4 Proposed Technical Specification Revisions Gentlemen:

The purpose of this letter is to transmit an application for amendment to Facility Operating License No. NPF-42 for Wolf Creek Generating Station (WCGS)

Unit No. 1. This license amendment request proposes revising Technical Specification 3.2.3, Figure 3.2-3, Table 2.2-1 and corresponding Bases. These changes reflect revised Reagtor Coolant System Temperature Measurement Uncertainties, a modified F part power multiplier and deletion of an inconsistency in applying shialtest exception 3.10.4. The application also revises specifications 3.1.2.5, 3.1.2.6, 3.5.1, 3.5.5 and 3.10.4. These Technical Specifications and corresponding Bases are required for Cycle 4 operation.

In order to facilitate transition to the proposed increased boron concentrations, Wolf Creek Nuclear Operating Corporation (WCNOC) requests I I

NRC approval of tais submittal prior to the shutdown of WCGS for its third refueling outage which is currently scheduled to begin on September 29, 1988. WCNOC presently intends to implement these Technical Specification revisions after entry into MODE 3 but prior to initial entry into MODE 6 c4 for refueling. In any case, the proposed revision to the Wolf Creek M Generating Station Technical Specifications will be fully implemented prior to gQ. STARTUP for Cycle 4 subject to formal Nuclear Regulatory Commission approval.

oo j The Reference proposed revisions to the thermal design flow and F H E" power multiplier for WCGS.N This submittal supercedes the portion of the Reference dealing with the F AH Part power multiplier. 3gl no TT t IdC o

)f 0 0' 03 0 Om Q. PO. Box 411 I E:Angton, KS 66839 / Phone: (316) 364-8831 fdelck 4il6D An Equal OpportMy Employer MEHC/ VET ($

l y s ET 88-0086 l Page 2 of 2 June 24, 1988 In accordance with 10 CFD 50.91, a copy of this application, with attachments is being provided to the designated Kansas state official.

Enclosed is a check (No. 1855) for the $150.00 application fee required by 10 CFR 170.21.

If you have any questions concerning this matter, plea:e contact me or

0. L. Maynard of my staff.

Very truly yours, N

John A. Bailey Vice President Engineering & Technical Services JAB /jad Enclosures cc: G. W. Allen (KOHE),w/a B. L. Bartlett (NRC), w/a D. D. Chamberlain (NRC),w/a R. O. Martin (NRC),w/a P. W. O'Connor (NRC),w/a(2) i

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STATE OF KANSAS )

) SS COUNTY OF COFFEY )

John A. Bailey, o f . lawful age, being first duly sworn upon oath says that he is Vice-President Engineering and Technical Services of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that he has executed that same for and on behalf of said Corporation with full power and aathority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By_ _

John A. Bailey l Vice-President Engineeringanh Technical Services SUBSCRIBED and swarn to before me this [ Y day of <

, 1988.

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t 5 Enclosure to ET 88-0086 Page 1 of 2 June 24, 1988 l

) CGS CYCLE 4 RELOAD AND TECHNICAL SPECIFICATION CHANGES DESCRIPTION 1

l Wolf Creek Generating Station (WCGS) is currently scheduled to conclude its third cycle of operation and commence its Refueling Outage on September 29, 1988. The startup of Cycle 4 is scheduled for early December, 1988. The Cycle 4 reload core is designed to achieve a burnup of 15280 MWD /MTU which is equivalent to 400 EFPD.

This reload core will utilize 72 new Westinghouse 17 X 17 fuel assemblies and 60 Wet Annular Burnable Absorber (WABA) asstmblies. The 72 new fuel assemblies will comprise Regions 6A and 68. Region 6A will be composed of 36 fuel assemblies enriched to 3.73 weight percent (wt%)

U-235 and Region 6B will be composed of 36 fuel assemblies enriched to 4.10 wt% U-235. A complete listing of the Cycle 4 reload core regions and their enrichments is presented in Table 1. All of the Region 6 fuel assemblies will be of the same niechan ical , nuclear and thermal hydraulic design as the fuel assemblies used in previous cycles with the exception of modified snag-resistant grid straps and two minor changes to the fuel assembly nozzles to improve their function during fuel handling operations.

As a result of the Cycle 4 reload core design, the following general Technical Specification changes are required:

1. Increase boron concentrations in the refueling water storage tank (RWST) and accumulators (Attachment A).

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2. Revise the E part power multiplier (Attachment B). Also included in Ab$chment B are proposed Technical Specification changes resulting from revised Reactor Coolant temperature  :

measurement uncertainties and an editorial change to correct an l

'nconsistency in applying Special Test Exception 3.10.4. j

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The proposed Technical Specification changes are provided as Attachments A and 3. Each Attachment is composed of three sections.Section I provides a complete Safety Evaluation.Section II provides the No Significant Hazards Consideration determination and Section III provides the marked-up, proposed Technical Specification revisions.

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, Enclosure to ET.88-0086

Page 2 of.2 June-24, 1988 3

TABLE 1 f

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WCGS CYCLE 4 RELOAD CORE 4

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4 Reaion Number of Assemblies Enrichment (wt% V-235) 1 j

1 17 2.10.

5 4 52 3.40 <

5A 20 3.00 1 58 32 3.20.

6A 36 3.73 1

68 36 4.10 h

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1 ATTACIDiENT A Increased Boron Concentration in Refueling Water Storage Tank and Accumulators

t Attachment A Page 1 of 20 June 24, 1988 SECTION I - SAFETY EVALUATION INTRODUCTION AND

SUMMARY

As a result of the evaluation of core design for Cycle 4 and subsequent cycles, changes have been identified in Technical Specifications dealing with boron concentration requirements. Boron concentration must be increased in the Refueling Water Storage Tank (RWST) .and the Safety Injection System (SIS) Accumulators. The specific changes that are required are:

1. Change in the minimum boron concentration in the RWST from 2000 ppm to 2400 ppm. (Technical Specification 3.1.2.5 and Bases).

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2. Change in boron concentration range in the RWST from 2000-2100 ppm to i 2400-2500 ppm (Technical Specification 3.1.2.6 and Bases). j

3. Change in accumulator boron concentration range from 1900-2100 ppm to 2300-2500 ppm. (Technical Specification 3.5.1).  ;

4. Change in RWST boron concentration range from 2000-2100 ppm to 2400-2500 ppm. (Technical Specification 3.5.5 and Bases).

Background

In order to support shutdown requirements for the fuel Cycle 4 core designs, the RWST and Safety Injection System accumulators boron concentrations must be increased above the current Technical Specification values. In addition, two minor editorial changes are being made to Bases 3/4.1.2 Boration Systems.

The maximum expected boration capability volume requirement is being revised from 83,745 gallons to 83,754 gallona of borated water from the RWST. The boration capability volume requirement below 200 F is being revised from 14,076 gallons to 14,071 gallons of borated RWST water. .These revisions are being made to effect the correct volumes assumed in the original analyses and are not the result of the proposed increase in the RWST boron concentration.

A. LOCA ANALYSIS AND RELATED DESIGN CONSIDERATIONS The RWST, accumulatoro and the Safety Injection System (SIS) are subsystems of the Emergency Core Cooling System (ECCS). Upon actuation of the SIS, borated water from the RWST is delivered to the Reactor Coolant System (RCS) in order to provide adequate core cooling as well as provide sufficient negative reactivity following steamline break transients to prevent excessive fuel failures. The accumulators are a passive system and provide borated water to the RCS when the system pressure drops below approximately 600 psig.

For a postulated LOCA, the ECCS is designed to limit the consequences of an accident to meet the acceptance criteria of 10 CFR 50.46. The LOCA analyses takes credit for pumped safety injection from the RWST and passive injection of accumulator water to prevent or mitigate the resulting clad temperature excursion. Also post-LOCA long-term core cooling takes credit for the l .

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Attachment A Page 2 of 20 June 24, 1988 available water in the RWST and accumulator in determining the post-LOCA RCS/ sump boron concentration and the hot leg switchover time to prevent boron precipitation. The effect of an increase in the RWST boron concentration Technical Specification range from 2000-2100 ppm to 2400-2500 ppm, and an increase in the accumulator boron concentration Technical Specification range from 1900-2000 ppm to 2300-2500 ppm is discussed below.

1. Small Break LOCA The small break LOCA analysis is cescribed in USAR section 15.6.5.

The current small break LOCA analyses were performed with the WFLASH Evaluation Model, which assumes the reactor core is brought to a suberitical condition by the trip reactivity of the control rods. There is no assumption requiring the presence of boron in the ECCS water or for negative reactivity provided by the soluble boron. Thus the changes in the RWST and accumulator boron concentrations do not alter the conclusions of the small break LOCA ,

analysis.

2. Large Break LOCA The large break LOCA analysis is described in USAR section 15.6.5.

The current large break LOCA analysis was performed with the NRC Approved 1981 Evaluation Model with BART. The large break LOCA analysis presented in the USAR does not take credit for the negative reactivity introduced by the soluble boron in the ECCS water in determining the reactor power during the early phases of a postulated LOCA. The large break LOCA also does not take credit for the negative reactivity introduced by the control rods. During a large break LOCA, the reactor is brought to a suberitical condition by the presence of voids in the core.

Since credit was not taken for the soluble boron that is present in the core, a change in the RWST or accumulator boron concentrations will have no effect on the current USAR large break LOCA analysis.

3. LOCA Short and Long Term Mass and Energy Releases The containment analyses, described in USAR section 6.2, considers the containment subcompartments, mass and energy releases for postulated LOCAs, and containment heat removal systems. For containment subcompartment analyses, an increase in the RWST and accumulator boron concentrations, would have no effect on the calculated results, since the short duration of the transient (<3 seconds) does not consider any safety injection flow taken from the RWST. The long term mass and energy release calculations do not take credit for the soluble boron present in the safety injection from the RWST supplied to the RCS. This is similar to LOCA analysis assumptions; and, therefore, an increase in ECCS water boron concentration, would have no effect on the calculated long term mass and energy releases.

4. Steam Generator Tubo Rupture The Steam Generator Tube Rupture (SGTR) accident is presented in USAR section 15.6.3. For the SGTR accident, the low pressurizer pressure

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t Attachment A Page 3 of 20 June 24, 1988 safety injection signal is actuated due to the decrease in the reactor coolant inventory shortly after reactor trip, and . borated water from the RWST is delivered (no accumulator actuation occurs) to the RCS. For the USAR SGTR analysis, .the primary to secondary break flow was assumed to be terminated at 30 minutes after the initiation of the SGTR event. However, the operator actions required to. terminate the break flow, including the initial RCS cooldown to provide subcooling margin, were not modeled in the SGTR analysis. Although the RCS cooldown is not modeled, sufficient shutdown margin is assumed to be available initially due to insertion of the control rods following reactor trip, and adequate shutdown margin is assumed to be maintained in the long term by borated safety injection water. If the RWST and accumulator boron concentrations are increased, this results in more negative reactivity insertion after trip in the SGTR accident. Therefore, the higher RWST and accumulator boron concentrations will have no adverse effect on the USAR SGTR analysis.

5. Post-LOCA Long-term Core Cooling Long-term core cooling is discussed in USAR section 15.6.5. The Westinghouse licensing position for satisfying the requirements of 10CFR50.46, paragraph (b), item (5) "Long Term Cooling' is defined in WCAP-8339. A Westinghouse Evaluation Model commitment is that the reactor will remain shutdown by borated ECCS water residing in the RCS/ sump after a LOCA. Since credit for the control rods is not taken for a large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a boron concentration that, when mixed with other water sources, will result in the reactor core remaining suberitical assuming all control rods out.

The effect on the post-LOCA RCS/ sump boron concentration as a result of changing the minimum Technical Specification boron concentration from 2000 to 2400 ppm for the RWST, and from 1900 to 2300 ppm for the accumulators is an increase of about 270 ppm in the RCS/ sump boron concentration. It has been confirmed that this proposed increase will provide enough margin to keep the core suberitical for the post-LOCA long-term core cooling requirement. Future reloads will be verified through the normal refueling safety evaluation process.

6. Hot Lea Switchover to Prevent Boron Precipitation A discussion on hot leg switchover time is presented in USAR suctions 6.3.2 and 15.6.5. A hot leg recirculation switchover time analysis has been performed to determine the time following a LOCA that bot leg recirculation should be initiated. This analysis addresses the concern of boron precipitation in the reactor vessel following a LOCA and has been performed to support the increase in Technical Specification RWST and Accumulator maximum boron concentration limit to 2500 ppm.

During a large break LOCA the plant switches to cold leg recirculation after the RWST switchover setpoint has been reached. If the break is in the cold leg there is a concern that the cold leg injection water will fail to establish flow through the core. Safety Injection (SI) entering the broken loop will spill out the break, while the SI entering the intact cold legs will

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Page 4 of 20 June 24, 1988 circulate around the downcomer and out the break. With no flow path established through the core the fluid in the core is stagnant.  ;

Steam is produced in the core from decay heat. In the analysis, .it is conservatively assumed that the boron associated with the steam remains in the vessel. Thus, as steam is boiled off with no circulation present in the core, the boric acid concentration increases in the vessel. The boron concentration 1 in the vessel will increase until the solubility limit of the boric acid m solution is reached at which time boron will begin to precipitate. As the boron precipitates, it may plate out on the fuel rods which would adversely affect their heat transfer characteristics. The purpose of the hot leg recirculation switchover time analysis is to provide a time at which hot leg 1 recirculation can be established such that boron precipitation in the core can  !

be prevented. l The calculation considers the increase in boric acid concentration in the i vessel during the long term cooling phase of a LOCA. The analysis assumes l that following a LOCA the steam boil off from the core does not carry any boron. A constant volume of liquid in the vessel is assumed so that as steam is boiled off and the boron is left behind, the boric acid concentration of i the vessel increases. The time when the boric acid solution reaches the l solubility limit less 4 weight percent is when hot leg recirculation should be initiated. The solubility limit less 4 weight percent at a solution temperature of 212 F has been established as 23.53I. Thus, when the boric acid solution concentration reaches 23.53I, hot leg recirculation should be initiated. Hot leg recirculation provides an injection path into the core which dilutes the boron solution and prevents the further build up of boron.

An analysis has been performed to determine the time following a LOCA l that switchover to hot leg recirculation should be initiated to prevent  ;

boron precipitation in the reactor vessel. This time has been determined  !

to be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The analysis considers the increase in boric acid concentration in the l reactor vessel during the long term cooling phase of a LOCA assuming a conservatively small effective vessel volume. This volume includes only the free volumes of the reactor core and upper plenum below the bottom of the hot leg nozzles. This assumption conservatively neglects the mixing of boric acid solution with directly connected volumes, such as the reactor ,

vessel lower. plenum. The calculation of boric acid concentration in the i reactor vessel considers a cold leg break of the RCS in which steam is generated in the core from decay heat while the boron associated with the boric acid solution is completely separated from the steam and remains in the effective vessel volume.

The results show that the maximum allowable boric acid concentration of 23.53 weight percent established by the NRC (the boric acid solubility limit less 4 weight percent), will not be exceeded in the vessel if hot leg recirculation is initiated 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after LOCA inception.

The operator should reference this switchover time against the reactor trip /SI signal. The typical time interval between the accident inception

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, l Attachment A Page 5 of 20 June 24, 1988 and the reactor trip /SI actuation signal is negligible when compared to j the switchover time. Procedure philosophy assumes that it would be very 1 to differentiate between break sizes and difficult for the operator locations. Therefore, one hot leg switchover time is used to cover the complete break spectrum.

7. Rod Eiection Mass and Energy Releases for Dose Calculations Rod Ejection mass releases are presented in USAR Figure 15.4-27.

Dose releases are discussed in section 15.4.8.3. The increase in the RWST and accumulator boron concentrations will be negligible on the mass releases for the rod ejection accident. Since the SI flow taken from the RWST is modeled under similar assumptions as in the large break and small break LOCA analyses there will be no adverse effect on the mass releases for the rod ejection accident.

8. LOCA Hydraulic Vessel and Loop Forces The LOCA hydraulic forcing functions resulting from a postulated LOCA are considered in the USAR sections 3.9(N).2.5 and 3.6.2.2.1.5. The increase in the RWST and accumulator boron concentrations will have no effect on the LOCA hydraulic forcing functions since the maximum forces are generated within the first few seconds after break initiation. For this reason the ECCS, including the RWST, is not considered in the LOCA hydraulic forces modeling; and, thus, the increase in RWST and accumulator boron concentrations will have no effect on the results of the LOCA hydraulic forcing function calculations.

9. Conclusions The increase in the RWST boron concentration from a range of 2000-2100 ppm to a range of 2400-2500 ppm and an increase in the accumulator boron concentration from a range of 1900-2000 ppm to a range of 2300-2500 ppm does not have a negative effect on the LOCA related accidents as previously described. Current margin to the post-LOCA shutdown requirement is increased with continued conformance verified through the normal performed safety evaluation process.

The higher concentrations decrease the allowable time for operator action to initiate hot leg recirculation to 12 hours. Emergency Operating Procedures will be modified to reflect the new hot leg switchover time. The resulting time requirement of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is still more than adequate to assure operator actions can be accomplished.

In conclusion, there is no adverse effect on the USAR LOCA related accidents for the proposed increases of the boron concentrations for the RWST and the accumulators at Wolf Creek Generating Station.

B. NON-LOCA ANALYSIS l 1

1 The only non-LOCA safety analysis in which boron from the RWST or accumulators )

is taken credit for, or assumed to be present, are those in which the SIS is l actuated. These analyses ares l l

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Attachment A I Page 6 of 20 I June 24, 1988 l l

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Inadvertent Operation of the Emergency Core Cooling System During Power Operation (USAR section 15.5.1)

Inadvertent Opening of a Steam Generator Relief or Safety Valve Causing a Depressurization of the Main Steam System (USAR section 15.1.4)

Steam System Piping Failure (USAP section 15.1.5)

Spectrum of Rod Cluster Control Assembly Ejection Accidents (USAR section 15.4.8)

Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures (USAR section 6.2)

The effect of the proposed Technical Specification revisions to the minimum RWST and accumulator boron concentrations on each of these transients is discussed below.

1. Inadvertent Operation of the Emergency Core Cooling System During Power Operation Spurious actuation of the emergency core cooling system while at power is classified as an ANS Condition II event. The accident results in a negative reactivity excursion due to injected boron from the RWST. The decreasing reactor power causes a drop in the core average temperature and coolant shrinkage. If reactor trip on SIS actuation is assumed not to occur.

the reactor will ultimately trip on low pressurizer pressure. DNBR does not drop below the initial value.

The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron concentration. No accumulator actuation is assumed and all other analysis assumptions are verified to remain applicable. If the minimum RWST boron concentration is increased from 2000 ppm to 2400 ppm, the negative reactivity excursion would occur l

at a faster rate causing a more rapid drop in the core average temperature and coolant shrinkage. The reactor will trip on low pressurizer pressure as before, though at an earlier time in the transient. Ao before, the DNBR j will not decreasa below the initial value.

Therefore, in consideration of the proposed changes, the conclusions of the evaluation are that the Condition II safety analysis acceptance criteria continue to be met and the conclusions in the USAR remain valid.

2. Inadvertent Opening of a Steam Generator Safety or Relief Valve An accidental depressurization of the main steam system due to the inadvertent opening of a steam generator safety or relief valve is classified as an ANS Condition II event. The accident results in a I cooldown of the RCS which, in the presence of a negative moderator temperature coefficient, causes a positive reactivity excursion. Borated ,

water from the RWST enters the core following actuation of the SIS on )

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low pressurizer pressure or low steamline pressure. The USAR demonstrates that the negative reactivity provided by the borated water from the RWST i limits the return to power to an acceptable level so that the minimum DNBR remains above the safety analysis limit. As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration in the RCS gradually increases, ultimately causing the core to become suberitical.

The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron concentration. No accumulator actuation is assumed and all other analysis assumptions are verified to remain applicable. If the minimum RWST boron concentration is increased from 2000 ppm to 2400 ppm, more negative reactivity would be available to terminate the return to power sooner and at a reduced peak power level.

Thus, the maximum core heat flux reached will be reduced. Additionally, the core vould become suberitical earlier in the transient. The minimum DNBR would be higher than for the case currently analyzed which assumes the minimum RWST boron concentrar'on of 2000 ppm. Therefore, in consideration of the proposed change the conclusion of the evaluation is 1 that the Condition II safety analys1 acceptance criteria continue to be  !

met and the conclusiens in the USAR remain valid.

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3. Steam System Pipinn Failure '

A major rupture of a main steam line is classified as an ANS l Condition IV event. This accident is more severe than the inadvertent I opening of a steam relief or safety valve and results in a rapid cooldown of the RCS which, in the presence of a negative moderator temperature coefficient, causes a positive reactivity excursion. Borated water from the RWST enters the core following actuation of the SIS on low steam line ,

pressure or low pressurizer pressure. The USAR demonstrates that the negative reactivity provided by the borated water from the RWST limits the l return to power to an acceptable level so that the minimum DNBR remains above the safety analysis limit. As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration in the RCS gradually increases, ultimately causing the core to become subcritical.

The proposed increases to the RWST and accumulator boron concentrations change only the safety analysis assumptions for SIS and accumulator boron concentrations, respectively. All other analysis assumptions are verified to remain applicable. Accumulator actuation is assumed for the 1.4 square foot steamline rupture with offsite power available. If the minimum RWST boron concentration is increased from 2000 to 2400 ppm and the minimum accumulator boron concentration is increased from 1900 to 2300 ppm, more negative reactivity would be available to terminate the re*, urn to power sooner and at a reduced peak power level.

Thus, the maximum core heat flux reached will be reduced. Additionally, the core would become suberitical earlier in the transient. The minimum DNBR would be higher than for the case currently analyzed assuming the minimum RWST boron concentration of 2000 ppm. The effect of the accumulator boron concentration increase would be less noticeable since accumulator injection occurs close in time to when the peak power level and maximum core heat flux are reached. Therefore, in consideration of

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Page 8 of 20 June 24, 1988 the proposed changes, it is concluded that the Condition IV safety analysis acceptance criteria continue to be met and the conclusions in the USAR remain valid.

4. Feedwater System Pipe Break A major feedwater line rupture is classified as an ANS Condition IV event. This accident is defined as a break in a feedwater line which is large enough to prevent the addition of sufficient feedwater to maintain shell side fluid inventory in the steam generators. Rapid decrease in secondary pressure following a feedline rupture may cause the low steamline pressure SIS signal to be reached. The negative reactivity insertion due to the addition of borated SIS water is not required to maintain the reactor in a suberitical condition following a feedline breaks although, the cold SIS water serves to reduce the RCS temperatures and pressures. The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron concentration. No accumulator actuation is assumed, and all other analysis assumptions are verified to remain applicable. An increase in the minimum RWST boron concentration from 2000 to 2400 ppm, will increase the negative reactivity insertion rate without significantly affecting the reduction of RCS temperatures and pressures.

Thus, an increase in the minimum RWST boron concentration to 2400 ppm will have no adverse impact on the feedwater line break analysis. Therefore, )

in consideration of the proposed changes, it is concluded that the Condition IV safety analysis acceptance criteria continue to be met and the conclusions in the USAR remain valid, i

5. Spectrum of Rod Cluster Control Assembly Eiection Accidents '

Mechanical failure of a control rod mechanism pressure housing resulting in the ejection of an RCCA and drive shaft is classified as an ANS Condition IV event. This accident results in a rapid positive reactivity insertion and system depressurization together with an adverse core power distribution, possibly leading to localized fuel rod d a ma ge. . Following the ejection of a control rod, the rapid nuclear power excursion causes the RCS to experience a large pressure rise due to the energy released into the coolant. RCS pressure then drops as fluid l inventory is lost through the break (maximum of 2 square inches) in the I control rod housing. As the RCS pressure continues to drop, actuation of '

the SIS on low pressurizer pressure will inject Lorated water from the RWST into the RCS.

The effect of the RWST boron concentration increase for the rod ejection accident from a LOCA standpoint is addressed in the safety evaluation of the proposed changes on the small break LOCA transient (USAR section 15.6.5). No consideration is given to potentially beneficial effects due to SIS or accumulator actuation for the calculations of the acceptance criteria related to fuel core damage and RCS and secondary integrity as defined in USAR section 15.4.8.

Therefore, an increase in the minimum RWST boron concentration from 2000 J

ppm to TSOO ppm does not affect the rod ejection accident analysis

Attachment A Page 9 of 20 June 24, 1988 calculations performed for USAR Section 15.4.8. In consideration of the proposed changes, it is concluded that the Condition IV non-LOCA safety analysis acceptance criteria continue to be met and the conclusions in USAR section 15.4.8 remain valid.

6. Mass and Enerav Release Analysis for Postulated Secondary Steam System Pipe Ruotures A major rupture of a main steamline results in a rapid cooldown of the RCS which, in the presence of a negative moderator temperature coefficient, causes a positive reactivity excursion. The calculation of steamline b?.eak mass and energy releases for use in determining peak containment pressure and temperature (USAR sectica 6.2) assumes that borated water from the RWST enters the core following actuation of the SIS on low steamline pressure, low pressurizer pressute, or Hi-1 containment pressure. USAR results demonstrate that the negative reactivity provided by the borated water from the RWST limits the return to power to an acceptable level so that the minimum DNBR remains above the safety analysis limit. Additionally, by limiting the return to power, the borated RWST water reduces the total energy that is dissipated via steam relense through the ruptured steam line. As the transient proceeds and more water from the RWST reaches the RCS, the boron concentration in the RCE gradually increases, in time causing the core to become subcritical.

The proposed increase to the RWST boron concentration changes only the safety analysis assumption for SIS boron concentration. No accumulator actuation is assumed and all other analysis assumptions are verified to remain applicable. An increase in the minimum RWST boron concentration from 2000 to 2400 ppm would add more negative reactivity to terminate the return to power sooner and reduce the peak power level. Thus, the maximum core heat flux reached will be reduced and the core would become suberitical earlier in the transient. Over the course o* the transient, the reduced peak power and earlier return to suberiticality would reduce the integral mass and energy releases, as a function of time, relative to those for cases analyzed assuming 2000 ppm boron in the RWST. Therefore, the conclusjons in the USAR remain valid.

7. Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant The boron dilution event in the USAR is analyzed for the six l technical specification MODES, (i.e., dilution during REFUELING, COLD SHUTDOWN, HOT SHUTDOWN, HOT STANDBY, START-UP, and POWER OPERATION). Pursuant {

to NUREG-0800 Standard Review Plan, the analyses are conducted to show that sufficient time exists for either operator intervention or, for some operational modes, for automatic protection system actuation from the time an alarm announces the unplanned dilution event until shutdown margin is lost.

Intervention consists of isolation of the diluent flow (by closing valves BG-LCV-ll2B and C) and initiation of borated water flow from the RWST (by opening valves BN-LCV-ll2D and E). Once operator or automatic protection system intervention occurs, the boron dilution event is considered terminated.

Correspondingly, the analysis methodology is independent of the boron concentration assumed in the RWST, accumulator, and boric acid tank (BAT) as

Attachment A Page 10 of 20 June 24, 1988 well as the minimum borated water volumes assumed in either the RWST or BAT.

Therefore, the conclusions in the USAR remain valid.

8. Conclusions The proposed increase in the RWST minimum boron concentration from 2000 to 2400 ppm and associated changes to the accumulators boron concentration range, have been evaluated for the Wolf Creek plant non-LOCA safety analysis design basis. It is concluded that these changes have no adverse impact on the non-LOCA accident analyses.

C. OTHER SAFETY ISSUES

1. Post-Accident Containment Sorav pH The Containment Spray System in conjunction with the Containment Fan Cooler System and the Emergency Core Cooling System, is capable of removing sufficient heat from the containment atmosphere following the hypothesized LOCA to maintain the containment pressure below the containment design pressure. In addition, the Containment Spray System reduceo the iodine and particulate product inventories in the containment post-LOCA atmosphere. To enhance the iodine absorption capacity of the containment spray, the spray solution is adjusted to an alkaline pH to promote iodine hydrolysis. This is accomplished by adding 282 - 31% concentration by weight NaOH solution in the spray. Approximately 2400 gallons of NaOH solution is presently required to be added to the containment sump to achieve a final post-LOCA containment sump pH of at least 8.5. A minimum pH of 8.5 in the containment sump is necessary to ensure long-term retention of iodine in the l solution. In order to ensure that the required amount of NaOH solution is I added to the sump, spray additive eductor isolation valves in the NaOH l solution supply headers are provided with an interlock from the tank level transmitters to preclude their closure prior to the addition of the required amount of NaOH solution. These valves, once opened by the Containment Spray Actuation Signal (CSAS) can be closed only after a close permissive signal is r given by low level switches installed on the Spray Additive Tank.

As a result of the proposed increase in fuel cycle length, it has been detennined that higher levels of boron concentrations are needed to be maintained in the Refueling Water Storage Tank (RWST) and accumulators.

It has been determined that additional NaOH solution is required to achieve the minimum long-term post-LOCA containment sump pH of 8.5, based on the increased boron concentrations. The setpoint of the subject low level switches must be lowered to ensure that the close permissive for the eductor isolation valves is received only after the required quantity of NaOH solution is added into the containment sump post-LOCA.

By maintaining the same rate of addition and same concentration of NaOH solution, the minimum pH value of the containment spray is affected.

The new pH range for the containment spray is 9 to 11 rather than the original pH range of 9.5 to 11. Alsc, since a greater quantity of NaOH solution is needed to achieve the long-term post-LOCA containment sump pH of at least 8.5, it will take less than 28 minutes longer to achieve this

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I l

Attachment A Page 11 of 20 June 24, 1988 pH, based on a worst-case single failure of one of the motor-operated valves in the spray additive line.

The minimum pH value of the containment spray and the longer duration required to achieve the long-term post-LOCA containment sump pH have been evaluated. The new pH of 9 to 11 has no adverse impati on achieving the final containment recirculation sump solution pH of at least 8.5. In addition, the revised pH range will not impact the elemental or particulate iodine absorption coefficients used in the accident analysis; therefore; the . radiological consequences of a LOCA as described in USAR Table 15.6-8 are not affected.

The additional time to add the NaOH solution is not significant because the NaOH solution will continue to be added during sump recirculation and will therefore result in a spray pH of greater than 9, and ensure a final sump solution with a pH of at least 8.5.

2. Environmental Oualification of Eauipment in Containmegi As described in USAR Section 3.11(B)l.2.2, Accident Environments -

Inside Containment, the worst case containment spray pH concentrations, resulting from a single failure, are 4 to 11. Therefore, the revised Containment Spray pH range will have no impact on the pH values utilized for environmental qualification.

As described in WCGS Technical Specification Bases 3/4.6.2.2, the limits on NaOH volume (4340 and 4540 gallons) and concentration of 28-31 percent by weight ensure a pH value of between 8.5 and 11 for the solution recirculated within containment post-LOCA. The subject modifications do not affect these requirements therefore the margin of safety is not reduced.

3. Hydrogen Generation in Containment The capability of the Containment Spray System to perform its safety-related functions as described in USAR Sections 6.2.2.1,is unaffectcd. A lower pH value decreases the rate of hydrogen generation.

Therefore, the rate or the total quantity of post-accident hydrogen generation, as discussed in USAR Section 6.2.5.2.3 is not adversely affected as the upper limit of pH for the Containment Spray remains unchanged.

4. Potential Precipitation of Boric Acid The increase in concentration of boron has been evaluated to assure that solubility limits are not exceeded at the minimum temperature limits for the RWST and SIS Accumulators. The combinations of highest concentrations and minimum allowed temperatures for each tank would not result in precipitation of boric acid crystals.

e Attachment A  ;

Page 12 of 20 June 24, 1988 SECTION II - ADDRESSING THE STANDARDS IN 10 CFR 50.92 This amendment request would revise the boron concentration limits for the RWST and the SIS accumulators. The following sections discuss the proposed changes under the three 50.92 standards:

Standard 1 - Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

, The proposed Technical Specification modifications involve changes in boron concentrations in systems used to mitigate p2 snt accidents and transients. The consec,uences of accidents and transients previously evaluated in the USAR, including LOCA and non-LOC.. events, have been evaluated and/or reanalyzed and are not significantly affected by the proposed changes.

Therefore, the proposed changes would not involve an increase in the probability of occurrence of an accident previously evaluated.

Standard 2 - Create the Possibility of a New or Different Kind of Accident from Any Accident Previously Evaluated The proposed Technical Specification changas would revise boron concentrations in systems used to mitigate plant accidents and transients.

No modifications to plant design or operations are associated with the changes with the exception of the operational requirement to switch to hot leg recirculation 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following a LOCA. However, the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would allow sufficient time for the require / erator action and, therefore, the possibility of boron precipitation . fuel rods following a LOCA is not significantly affected. Therefore, the possibility of a new or different kind of accident would not be created by the proposed changes.

Standard 3 - Involve Significant Reduction in a Margin of Safety The proposed revisions in boron concentrations affect safety analysis inputs and assumptions. The affected safety analyses have been evaluated, and it has been determined that all applicable safety criteria are met with no significant adverse affects on analyses results. Therefore, the margin of sr.fety as defined by the USAR, safety analyses, and the Technical Specification Bases would not ha significantly reduced.

Based on the above discussions and those presented in Attachment 1 , it has been determined that the requested Technical Specification revisions do not involve a significant increase in the probability or consequences of an accident or other adverse condition over previous ovaluations or create the possibility of a new or different kind of accident over previous evaluations; or involve a significant reduction in a margin of safety.

Therefore, the requested changes do not involve a significant hazards consideration.

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