ML20003B730

Discusses Submittal Schedule for Responses to Reactor Sys Branch Open Items on NUREG-0737 Re Reactor Vessel Vent, Thermal Mechanical Rept,Revised Small Break LOCA Model & Plant Specific Calculation Per 10CFR50.46

ML20003B730
Person / Time
Site:	Summer
Issue date:	02/19/1981
From:	Nichols T SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-2.K.2, TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM NUDOCS 8102250400
Download: ML20003B730 (25)

Text

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. s SOUTH CAROLINA ELECTRIC a gas COMPANY POST OF FICE BC A 764 CotuweiA. SOUTH CAROLINA 29218

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February 19, 1981 a u on. ~.s Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation

'U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 4

Reactor Systems Branch - Open Items

Dear Mr. Denton:

The following is a discussion of open items from the Reactor Systems Branch.regarding NUREG-0737 and other miscellaneous items.

-1.

II.B.1 - Reactor Vessel Vents S

FSAR Section 5.5.15 gives the major portion of the resp?nses to this' item. The attached marked up pages (which are included in Amendment 24 to be' issued on March 2, 1981) provides responses

.to'open items discussed with Mr. Chu Liang. FSAR Section 5.5.13 provides a. discussion on venting the pressurizer. The attached marked up table 1.8.1 (Amendment 24) shows reference to that section. The attached generic procedures " Reactor: Vessel Head Vent Operation".from.the Westinghouse Owner's Group provides guidelines to our' operations ~ group to develop plant-procedures for operating the' vessel vent system.- Plant specific guidelines are due-to us by. March 31,.1981.

Plant procedures.will be sent

~

to you for review within the next few months. -This will be well.

before the requirement of January 1, 1982.'

Finally, the system

.is,being installed in the plant.at this; time. This will meet the requirement of' installation by July 1, 1982'.

~2.

II.K.2 Thermal Mechanical' Report M

$ NUREG-0737 item II.K.2-13 allows operating plants to complete

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5 their required ' analyses by January 1,1982, but requires applicants N

y Difor. operating licenses to submit : analyses 'six (6) months prior to qM -

'P E

'ItFissuance of the staff Savety Evaluation Report (SER) for a full dj'h 7p'owerlicense. The' staff SER was-issued.in-February 1981;.six (6) m imonths prior to this.date is September 1980. This date proceeds N

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]thedate-ofissdacce~of.NUREG-0737.

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Mr. Harold R. Denton February 19, 1981 Page Two II.K.2-13 Thermal Mechanical Report (con't)

SCE&G requests that the required submittal date for our analyses be shifted to January 1,1982, the date for OL licensees, for the

.following reasons:

1.

During the period-to January 1, 1982 the Virgil C. Summer Nuclear Station reactor vessel will have undergone an insignificant amount of_ irradiation damage. However, some Westinghouse reactor vessels have experienced more than ten years of operation. If the NRC finds the January 1, 1982 date acceptable for these older vessels, SCE&G concludes that the NRC should find the January 1, 1982 date accepcable for the Virgil C. Summer Nuclear Station 2.

The Westinghouse Owner's Group is embarking on a program to evaluate plants on a generic basis. Plant specific analyses,

. if required, will be based on the results of the generic analyses..SCE&G believes.that a more comprehensive analyses and'the best results can be obtained for the Virgil C. Summer Nuclear-Station vessel by participation in. this generic program.

-Generic results will be presented by January 1,1982.

3.

'II.K.3 Revised Small Break LOCA Model In~our. January 6, 1981 letter a discussion was provided on the revised small break 'LOCA model. _ ' Current plans by the Westinghouse 10wner's Group are to have this revised model submitted to the NRC

-by= January 1, 1982.

4.

II.K.3 Plant Specific C'alculations to Show Compliance with

~

10 CFR Part 50.46 In our January 6,11981 letter.a discussion was provided on the plant

' specific analyses. If.the results of the new model indicate that a plant' specific small break LOCA analyses is required to conform with the-requirements with 10 CFR 50.46, then it will be. submitted by January:1, 1983.-

15..

II.K.3 Automatic Power Operated Relief Valve Operation Current plans Lby the Westinghouse Owner's Group are to have the report-

. issued-by March 15J 1981. This was discussed with Mr. Brian Sheron by

the Westinghoure. Owner's : Group on February 17, 1981.

6.

II.K.3 ECCS Outages

~The attached marked up FSAR Section 13.5.1.-14'is provided to respond.to this item. This material will_be' included.in FSAR Amendment 24.

Mr. liarold R. Denton February 19, 1981 Page Three Also included in this letter are several minor changes in the boron dilution evaluation in Chapter 15.2.

This material, which will be included in FSAR Amendment 24, is provided in order to expedite its review.

If you have any questions, please let us know.

Very truly yours, A

T. C. Nichols, Jr.

RBC:TCN:rh cc:

V. C. Summer G. H. Fischer T. C. Nichols, Jr.

E. 11. Cr ews, J r.

O. W. Dixon, Jr.

C. A. Price D. A. Nauman W. A. Williams, Jr.

R. B. Clary A. R. Koon A. A. Smith

11. N. Cyrus J. B. Knotts, Jr.

J. L. Skolds B. A. Bursey O. W. Bradham ISEG PRS NPCF/Whitaker File R. Faas A

TABLE 1.8-1 b.J CROSS REFERENCE TMI ACTION PLAN REQUIREMEhTS TO FSAR SECTIONS ACTION PLAN REQUIREMENT FSAR SECTION I.A.1.1 13.1.2.1, 13.1.2.2.2, 13.1.3.1.26, 13.2.1.7.7 I.A.1.2 13.1.2.2, 13.5.1.3.1 I.A.1.3 13.1.2.3, 13.5.1.3 I.A.2.1 13.2.1.1.6, 13.2.3.1, SCE&G 1etters to NRC dated 10/28/80, 10/31/80

}

I.A.2.3 13.2.2, SCE&G 1etter to NRC dated 10/28/80 I.A.3.1 13.2.1.1.6, 13.2.2.6.1, SCE&G 1etter to NRC dated 23 10/28/80 I.B.1.2 13.1.2.1, 13.5.1.3.1, SCE&G 1etter dated 1/2/81 I,. C.1 6.3.3.3.1, 13.5.2, SCE&G 1etters dated 1*/14/80, 12/2/80' and Westinghouse Owners Group Letter 0G-47 dated 12/15/80 I.C.2 13.5.1.3 jj I.C.3 13.5.1.3.1, Tech Spec (6.1.2)

I.C.4 13.5.1.3 I.C.5 13.5.1.13 jj t '

I.C.6 13.5.1.6 I.C.7 13.5.1.3.2 1.C.8 13.5.1.3.3, SCE&G 1etter to NRC dated 11/14/80 I.D.1 1.2.3.1 I.D.2-7.7.3 I.G.I.

14.1.4.4 II.B.1 5.5.15, Question 211.133,5,5.13, scratemeJ.PJk8/SI.

2!

II.B.2 12.1.2.3,~ App. 12A, SCE&G 1etters to b1C dated 8/27/80 and 11/21/80 i :i.

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ZY 1.8-2

[d 4 N"h.

AMENDMENT 53 JellillastT, 1981

TABII 1.8-1 (Continued)

CROSS REFERENCE TMI ACTION PLAN REQUIREMENTS TO FSAR SECTIONS i

'I ACTION Pl.AN REQUIREMENT FSAR SECTION C

II.B.3 9.3.2, App. 12A, Responses to Questions 321.13 and 321.16 II.B.~4 13.2.1.1.6 II.D.1 5.5.13.4 II.D.3 1.2.3.1, 1.7, 5.5.10.2.2.4, 5.6, 7.7.4, SCE&G 1etter to NRC dated 1/13/81 II.E.1.1 SCESG 1etters to NRC dated 8/15/80, 11/5/80 and 12/2/80 23

- (

II.E.1.2 7.3.1.1.1, 7.3.2.2, 7.3.2.3, 7.5.1, 10.4.9.1, 10.4.9.2, 10.4.9.3, 10.4.9.5 II.E.3.1 8.3.1.1.2.a' II.E.4.1 6.2.5.2.1 II.E.4.2 6.2.4.3; 9.4.8.2.2, Items 8.1 and 8.m; 9.4.8.2.3,

.l Items 8.h and 8.8 II.F.1 6.2.5.1.3, 6.2.5.2.3, 6.2.5.3.3, 6.2.5.4.3,

,4 6.2.5.5.3, 6.2.5.5.4, Table 7.5-1, 7.7.3.1.c, l

10.4.2, 11.4, 11.4.2., Figure 11.4-2, 12.2.5, Figure 12.2-2, 12.3.2.2, Response to Question 321.14, SCE&G 1etters to NRC dated 8/28/80 and 12/22/80 II.F.2 1.2.3.1, 5.6, 7.7.5, 7.7.6, SCE6G 1etters to NRC dated 12/4/80, 12/15/80, and, 12/30/80 II.G.1 7.4.1.2.1, 8.3.1.1.3 II.K.1 7.3.1.1, 13.5.1.6, and Technical Specifications

-II.K.2 SCE6C lettertto NRC dated 1/6/81 2h)[Tl, N

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1.8-3 AMENDMEN1* EB l

M*M destemstV,.1981 d8

TABLE 1.8-1 (Continued)

'a CROSS REFERENCE DfI ACTION PLAN REQUIREMENTS TO FSAR SECTIONS ACTION PLAN REQUIREMENT FSAR SECTION

)

II.K.3 5.5.2.3 ; 7.2.1.1. 2, I tet. 6; Figure 7.7-4; Technical Specifications, SCE&G 1etters to NRC J

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dated 1/6/81(W 1J'9/t1and Response to Question 211.123, j

23 LV 13.5.1.3.4 / 3, 5, I, 8 Y j

III.A.1.1 Radiation Emergency Plan III.A.1.2 1.2.3.1, 6.4, 7.7.3, 12.1.4.2, 12.3.2.2.4, 15.4.1.3 III.A.2 2.3.3.2, Radiation Eciergency Plan III.D 1.1 6.3.2.11.3, Response to Questions 321.12 and 321.15, Technical Specifications III.D.3.3 6.2.5.1.3, 6.2.5.2.3, 6.2.5.3.3, 6.2.5.4.3,

6. 2. 5. 5.4, l'2 '.1.4. 2, 12,3.2.2.4, Response to Question 331.43 III.D.3.4 2.2.1, 2.2.2, 2.2.3, 6.4, 15.4, SCE&G 1etters to NRC j{

dated 11/25/80, 12/15/80 s

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1.8-4 M

AMENDMENT B g

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5.5.15 REACTOR VESSEL HEAD VENT SYSTEM 5.5.15.1.

Design Basis Ihe basic function of the Reactor Vessel Head " Vent System (RVHVS) is to remove noncondensable gases or steam from the reactor vessel head. This system is designed to mitigate a possible condition of inadequate core cooling or impaired material circulation resulting from the accumulation of noncondensable gases in the RCS. The design of the RVHVS is in accordance with the requirements of NUREG-0578 and subsequent definitions and clarifications (References 2 and 3).

5.5.15.2 Design Description and Evaluation I

5.5.15.2.1 General Description The RVHVS is designed to remove noncondensable gases or steam from the reactor coolant system via remote manual operations from the control room. The system discharges to the pressurizer relief tank. The RVHVS is designed to vent a volume of. hydrogen, non-condensible gases, etc., at system design pressure and temperature approximately equivalent to one-half of the reactor coolant system volume in one hour.

AH2 burn from a 100% zircalloy/H O reaction has been addressed and it has been 2

determined that containment integrity would not be breached. Therefore, contained venting.outside containment is not considered necessary.

The flow diagram of the RVHVS is shown in Figure 5.5-13.

The RVHVS consists of two parallel flow paths with redunda'nt isolation valves in each flow path. The venting operation uses only one of the flow paths at any time. The physical

~1ayout of the RCS hot leg piping is such that its entire volume can be vented via the RVHV system.

' The equipment design parameters are listed in Table 5.5-16.

As indicated above, normally the venting from the RVHV is contained by the Pressurizer Relief Tank.

However, venting to containment could occur if the rupture disc ruptures.

In

I that case the location of the PRT is such that excellent gas communication exists within the secondary shield are that any gas that escapes from the PRT will be i

readily mixed with the containment atomosphere with additional mixing being pro-3 moted by the Reactor Building Ventilation System. The active portion of the system

2 -

consists of four two-inch motor operated isolation valves connected to the reactor

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vessel head ~ vent pipe. The isolation valves in series in each' flow path are powered by opposite vital power supplies. The isolation valves are fail as is, active j,

valves. One normally closed isolation valve and one normally open valve are

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located in each flow path. Leakage past the vent valves during normal plant opera--

l tion is detected by the accostic ' leak monitoring syst'em which is described in l.

Section 7.6.9.

All of the isolation valves are qualified to IEEE-323-1974, l

344-1978, and 382-1972 and to the requirements of Regulatory Guide 1.48 as des-l cribed in Appendix 3A.

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'~'g If one single active failure prevents a venting operation through one l

l-J Y flow path, the redundant path is available for venting.

Similarly, the two isolciion valves in each flow path provide a single failure method of isolating the venting system. With two valves in series, the failure i

j of any one valve or power supply will no*. inadvertently open a vent path.

Thus, the combination of safety grade train assighments and valve l

failure modes will not prevent vessel head venting nor ve'nting isolation 1

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with any single ac tive f ailure.

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The RV11VS has two normally de-energized valves in series in each flow path. Power lockout capability to all four isolation valves is provided by administrative control at the motor control load center.

This l.

arrangement eliminates the possibility of a spuriously opened flow path.

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,The system is operated from the control room.

The position indication from each valve is monitored in the control room by status lights.

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.The RVilVS is connected to the~ head vent pipe as shown on Figure 5.5-13.

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The' system is orificed to limit the blowdown f rom a break downstream of I

either of'the. orifices to within the capacity of one of the centrifugal

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j charging ~ pumps.-

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A break of the RVi!VS line upstream of the orifices,would result in a small LOCA of.not greater than one inc5 diameter.

Such a break is similar to'those analyzed in WCAP-9600-(Reference 4).

Since a bteak in l'

the head vent'line would behave similarly to the hot leg breah c.ase 1

presented in WCAP-9600, the resul ts : presented therein.a're applicabic to p

l a RVilVS line break. This postulated vent'line break,'therefore, results in no calculated core uncovery.-

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i All piping and equipnent frun the vessel 1

! vent up to and including l

the second isolation valve in each flow path are designed'and fabricated in accordance with ASME,Section III, Class 1 requirements. The I

remainder of the piping and equipment is non-nuclear safety, but is scismically supported up to the 12" pressurizer relief line.

The system provides for venting the reactor vessel head by using only safety grade equipment. The RVilVS satisfies applicable requirements and industry standards including ASME Code classification, safety classi-fication single-failure criteria, and environmental qualification.

(

5.5.15.2.2 Supports

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The ' vent system piping is supported to ensure that the resulting loads and stresses on the piping and on the vent connection to the vessel head are acceptable.

The support design for attaching th2 head vent system piping to the reactor vessel head lif ting leg is shown in Figure 5.5-14.

The support i

'is a two-part clamp configuration, called a double bolt riser clamp.

The clamp and. associated bolts, nuts, spacers, and washers are made of stainless steel. A gap exists between the one inch head vent pipe and the support clamp to allow for thermal expansion in the vertical direction.

a The support design for attaching the' head vent system piping to the CRDM Seismic Support Platform is -shown in Figure 5.5-15.

This support is a

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two-part clamp configuration, called a double bolt clamp bracket.

This

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clamp support is.used to rigidly support the piping in the radial direction.

The clamp and associated bolts, nuts, spacers, and washers are made of stainless. steel, with high strength hold down bolts threaded l

into the deck of the CRDM-Seismic Support Platform'.

A gap exists on between-the one inch hea'd vent pipe and the support clamp to allow for

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thermal expansion in the ' axial direction.

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5.5-64b T

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Al* s u;>po r t s and support st ruct ures ceaply with the requirements of the AISC Code, Part II.

G 5.5.15.3 An alytical Consi de rat i ons The analysis 'of the reactor vessel head vent piping is based on the following plant operation conditions d e fi ned in the ASME Code Section i

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III:

1.

Normal Condition:

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Pressure deadweight, and thermal expansion analysis of the vent pipe during a) normal reactor operation with the two inbeard vent iso-lation valves closed and b) post-refueling venting.

i 2.

Upset Condition:

Loads generated by the Operating Basis Earthquake (OBE) response spectra.

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3.

Faulted Coadition:

Loads generated by the Safe Shutdown Earthquake (SSE) and by valve thrust during venting.

In accordance with ASME III, faulted con-ditions are not included in fatique evaluation.

I I

The Class I piping used for the reactor vessel head vent is one inch schedule 160 and, therefore, in accordance with ASME Section III, is

)

analyzed f ollowing the procedures of NC-3600 for Class II piping.

J For all plant operating conditions listed above, the piping stresses are shown to meet the requirements of equations (8), (9), (10) or (11) of ASME III, Section NC-3660, with a design temperature of 6500F and a design pressure of 2485 psig.

U 5.5-64c i

ii

r 5.5.16 REFERENCES 1.

" Reactor Coolant Pucp Integrity in LCCA", WCAP-8163, Septer.ber, 1973.

2.

Letter from D. B. Vassallo (NRC) to all Applicants f or an Operating I

License, "rolicvup Actions Resul ting Fraa the NFC Staff Reviews Regarding the Three Mile Island Unit 2 Accident", and Enclosure 4:

Insta11atien of Renotely Operated Iligh ' Point Vents in the Reactor Coolant System, September 27, 1979.

3.

Letter frem D. B. Vassallo (NRC) to all Applicants fo. an Operating License, " Discussion of Lessons Learned Short Term Require =ents,", pp. 44-49, Reactor Coolant Sys tem Venting, Novc=ber 9,

'~'

1979.

,(

4.

" Report on Scall Break Accidents t or Wes tinghouse NSSS Sys tec,"

WCAP-9600, June, 1979, (specifically Case F, Section 3.2).

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REUEF TANK NOTES:

1. 3/8 INCH FLOW RCSTBtCTING ORIFICE.'

2. NCRMAL VENT LINE.

3. ~ 2/3 INCH Ft.QW RESTRtCTION ~

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4. STM AP-QM ACQU3 TIC MON 170mNG SENSOR

5. ! P!PE FAdRICATED TO COD 2 Ct. ASS 2 CUT NOT STAWED,'

Figure 5.5-13. Schematic Flow Diagramot the Reactor Verwt Hee Vent System l

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REACTOR VESSEL HEAD VENT OPERATIO!!

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REVISION 0 FEBRUARY, 1981 1

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e REACTOR VESSEL HEAD VENT OPERATION A.

PtnPOSE The objective of thes'e instructions is to specify required operator actions and precautions necessary to remove geses from the reactor vessel head by operation of the Reactor Vessel he.:d Vent.

CAUTION: This venting guideline should not be used as the primary I

means to mitigate an Inadequate Core Cooling event.

Refer to Inadequate Core Cooling Guidelines for appropriate operator actions and precautions.

CAUTION:

This venting guideline assumes that the reactor containment conditions are near normal conditions and that any venting operation is performed prior to throttling safety injection flow 'during a POST-LOCA cooldown and depressurization operation.

B.

SYMPTOMS For plants with a RV level indication l-l.

Reactor vessel level is less than (insert plant specific value which= includes an allowance for normal channel accuracy) percent of span.

For plants with/without a RV level indication 2.. Abnormal reactor coolant system conditions such as large variations in pressurizer level during normal charging of spraying operations-have occurred.-

3.-

If available, reactor vessel he,ad temperatures equal to.or greater -

than saturation temperature.

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4 Plant events have occurred (such as accumulator tank discharge, rapid RCS cooldown, or core uncovery events) that may result in the presence o# a gaseous void in the vessel head.

C.

IMMEDIATE ACTIONS UOrsi D.

SUBSEGUENT ACTIONS i.

l j;

CAUTION:

Do not trip any running or start any non-operating reactor coolant pumps during the performance of the following actions.

NOTE:

If the safety injection system is in operation, then the actions of steps marked by an asterisk will not be applicable.

1.

Terminate any changes to the reactor coolant system that may be in progress and bring the RCS to as close to a steady-state condition as possible.

m.

• 2.

Attempt to recombine any condensible gases by increasing RCS pressure through the use of the pressurizer backup heaters and increased charging flow.

If this step is successful in condensing the gas volume in the vessel head (as indicated by a return to normal readings in those parameters used to determine the presence of the gases) then return to the appropriate operating instruction.

CAUTION:

Increased charging flow with condensible gases in the RCS may result in a decreasing pressurizer level.

If pressurizer level decreases to less than 20% of span, then-attempt to restore level by continuing the charging flow or manually starting safety injection pumps.

If-level cannot be restored, then manually. initiate safety injection and proceed to E0I-0, Immediate Actions and Diagnostics.

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p.y 3.

In preparation for venting, isolate the containment purge and exhaust system and the pressure vacuum relief line and start all available containment air circulation equipment.

4.

Increase the RCS sub-cooling to (insert plant specific value which is 50'F above the value which is the sum of the errors for the temperature measurement system used. and for the pressure measurement system translated into temperature using the

,j saturation tables) by either initiating an RCS pressurization or by dumping steam from the non-faulted steam generators.

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5.

If required, perform the actions of' ppendix B to determine the maximum allowable time period for venting (only for plants which vent directly to containment).

• 6.

Isolate letdown and initiate an RCS makeup by the chemical volume and control system to increase pressurizer level to greater than 50% of span.

• 7.

If not already performed, manually block the low pressure SI initiation if the permissive is anergized.

CAUTION: The venting operation may result in pressure decreasing i

below the SI setpoint. Action should be taken to manually block the autnmatic SI signal when the l

permissive is energized.

• 8.

Increase charging flow to maximum to limit the pressurizer pressure and level decrease during the venting period.

NOTE: Observe the pressurizer level trend during the venting and, from the following conditions, determine the probable status of the reactor coolant system.

i t.

a)

Increasina pressurizer level - Gaseous voMs exist in the RCS other than the reactor vessel head or pre.turizer.

b) Constant pressurizer level - fio significant gaseous voids exist in the reactor coolant system.

c) Decreasina pressurizer level - Gaseous void exists in the reactor vessel head.

9., Open the vent isolation valves in one head vent flow path, f

fiOTE:

If one or both valves fail to open, close both valves and open the isolation valve in the parallel flow path.

10. Close both vent isolation valves when:

.a) Reactor vessel level indication stabilizes, E

b) The time period determined in Step 5 is met, 0_g c) Pressurizer pressure decreases by 200 psi, E

d) Pressurizer level decreases below 20 percent of span E

e) Reactor coolant sub-cooling decreases below (insert plant specific value which is the sum of the errors for the temperature measurement system used, and for the pressure measurement system translated into temperature using the saturation tables).

E f) The reactor' vessel head is refilled as indicated by a decrease in the rate of a depressurization or a change in the rate of the pressurizer level. trend.

I.

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CAUTION:

If during the venting period, a loss of reactor coolant pump operation occurs, continue the venting and allow natural circulation to establish itself.

• ll.

Re-establish normal charging and letdown to maintain the pressurizer water level in the operating range.

.t

• 12.

Evaluate the response of the pressurizer level trend to determine c

if a gas bubble existed in the vessel head.

If a gas bubble

+

existed and the venting was terminated prior to the vessel head being completely refilled, then return to Step 4.

NOTE:

If. multiple venting operations are required and the con-tainment hydrogen concentration is equal to or greater than 3 volume percent, the provisions must be made to remove or reduce the volume of hydrogen from the containment prior to re-opening the reactor vessel head tent.

13. Return-to the appropriate operating instruction following the successful completion' of the venting of the reactor vessel head.

10:

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APPEllDIX "A" RV HEAD VErlT GUIDELIrlE RCS GASEGUS VOID DETECTI0tl AfiD SIZIriG 1.

Achieve a constant pressurizer level and pressure condition.

2.. Place the RCS wide range or pressurizer pressure and the pressurizer level on trend recorders. The scale should be 150 psig pressure and 10% of span for level, v

8 3.

Record the following parameters.

l RCS Pressure PSI

=

PZR Level

=

Charging Rate GPM

=

Seal Injection Flow =

GPM Seal Leakoff Low GPM

=

Time

=

.j-4.

Isolate the RCS letdown fl'ow, turn off all pressurizer heaters, and terminate the' pressurizer spray by placing the spray control in manual and zeroing the demand signal.

1 5.

Allow the RCS charging flow to either increase RCS pressure 100 psi or increase pressurizer level 5% of span.

t 6.

Record the RCS pressure, pressurizer level and time.

RCS Pressure =

PSI PZR Level

=

Time

=

7.

Reinitiate RCS letdown flow and restore normal pressurizer pressure h i % 2es h u N 18"Arl$4.}!d38.trylggg d j ;; g g;gy g gg g.pg gigggd S w m = W h.Sc ; e w.. m,1;enp 4_wsup = Eri-;9 m u ; T b -c w e w i e,_ A cpigf.cI(msry h y p $ s % 4 J +:3 gf;y P,ff ei W $ f @ M $ 7 s M Vs f s 5 :K9fhyC4?%Wyk'-

i:

w.

1. y 9

+n-

-g

8.

Calculate the initial and final prassurizer vapor space volumes.

3 Initial Vapor Volume = (1-PZR Level % X Total Cylindrical PZR Volume FT ),

3 (Upper Spherical Volume FT )

3 FT

=

Final Vaper* Volume

= (Initial Volume) - (a PZR Level X Total Cylindrical Volume) 3 FT

=

.;f 9.

Determine the total charged volume into.the RCS.

Charged Volume = (Charging + Seal Injection - Seal Leakoff GPM) X 1

(Time) X (7.45 g )

GPM 3

FT

=

10. Determine the expected pressurizer level change.

3 Expected a level = (Charging Volume FT )-X (Time) X (Total P olume fid)'

,4

=

111.

If the actual pressurizer level change is less than the expected

. level change then a gaseous void exists in the reactor coolant system.

Perform the following step to determine the volume of the RCS void.

12. The initial and final RCS-gaseous void volumes can be calculated I

from'the following equations.

Initial RCS Void = (Initial Vacor Volume) - (Final Vacor Volume) -(Charced Volu-(), Initial Pressure)

Final Pressure 3

FT

=

Final RCS Void = (Initial P.CS Void) X (Initial Pressure)

(Final Pressure) 3 FT

=

i i

i 1

.I I

A

t APPEriDIX "B" RV HEAD VEliT GUIDELIllE VEllTIriG TIME PERIOD 1.

Convert the containment free-volume to containment volume at standard temperature and pressure conditicns.

Cont.' Volume (STP) = (Cont. Volume FT ) X (Cont.

ure")X(Co p.")

3 q,

3 FT

=

!i

• Temperature in degrees Rankine (*F + 460)

)

• • If containment pressure has increased above 14.7 psia then use 14.7 psig as pressure for conservatism.

2.

Determine the containment hydrogen concentration in volume percent units.

fiOTE: The containment hydrogen concentration will be insigificant if there has been no leakage from the RCS to the containment.

3.

Calculate the maximrm hydrogen volume that can be vented to the containment which will result in a contcinment hydrogen cencentration of-less than or ec,ual to 3 volume percent.

l, Maximum H2 Volume = (3.0%-Cont. Ho Concentra on %) X (Cont. Volume [STP])

to be Vented

[

4 From Curve #1 (P.CS Pressure vs. H Flow Rate) determine the allowable 2

venting perio'. which will limit tne containment hydrogen concentration to 3 volume percent.

Venting Period = J x. H9 Vented (From Step 3)

H2 Flow Rate Mins.

=

(

lll a.

" * * ' 13.5.1.10 Control of Specimi Processes During Operations Procedures nese procedures assure that special processes are accomplished under 3

controlled conditions in accordance with applicable codes, standards, specifications, criteria and other special requirements using qualified personnel and procedures.

13.5.1.11 Nonconformance Control / Deficiency Reporting Procedures l3 Rese procedures provide for control of items, services or activities Wich do not conform to requirements.

Rese procedures include inst ru ctions for identification, documentation, segregation, notifica-tion of affected organizations and method of disposition of such items, services or activities.

~

13.5. 1. 12 Test Control Procedures Rese procedures assure that testing required to demonstrate that an 3

iten will perform satisf actorily in service is accomplished properly.

{

Test procedures incorporate or reference the requirements and accep-tance limits con tained in applicable design documents.

These test procedures may include preoperational tests, initial operational phase tests, surveillance tests. and tests during design, fabrication and construction activities associated with plant maintenance and modification.

f 13.5.1.13 Feedback of Operating Experience In accordance with NUREG 0737, item I.C.5, a program will be established C-23 for evaluating operating plant experience and providing the results of the l

cvaluations, as necessary, to pertinent plant personnel. This program will primarily be performed as a function of the Shift Technical Advisor Group. The services of " Industry Groups" such as INPO will be utilized to the extent possible in the performance of this function.

q l *3s S'o ir / 4 1 op l 13.5.2 CONTROL ROOM OPERATING PROCEDURES' Control room operating procedures are those procedures that are per-forced by the licensed Control Room Operator or under his direction i

2. '/ I 13.5-7 AMENDMENT EB JANUARY, 1981

't

Insert on Page 13.5-7 13.5.1.14 ECCS Outages In accordance with NUREG 0737, item II k 3.17, a program has been established using existing plant procedu.es for data collection for determining ECCS outage times.

A plant pro'edure for removal and restoration of station equipment provides metsures for data collection.

The ECCS data taken by this procedure will be ruviewed by appropriate plant personnel to determine if improvements to av..'1'bility of ECCS is needed.

9

=

1.

Dtlution During kefueling An uncontrolled boron dilution accident based on a failure in the 6

primary water makeup system cannot occur during refueling. This accident is prevented by administrative controls which isolate the RCS from the potential source of unborated water.

6430, Valves "^"  %, Fr - ' % 8454, 8441, and 8439 will be locked 5

closed during refueling operations. These. valves will block the flow paths which could allow unborated makeup water to reach the 6

(4 Acomo To TV E RCS. Any makeup which is required during refueling will bei N d REA(.TCA co0LANT 5VtTEM OY U.NLOCKING THE5E VALVES ANO INI,TIATING THE

= r

,,'4-A tram tta r-I-

' y water s*srage tsak. imf tire hm REGVLRED BLE.A1DED MAKF_UP \\NATER. I-MW. AFTER THE AEGUIRED VdLUNil ONLENDED N\\AKEup FLOW HA5 EEEM ADDEo,THESE VALVE.5 WILL AGAI Aj 8' LOCKE.D CLOSE D. AN ALTERN ATE. SCORCE OF BC The most limiting al ternate source of uncontrolled boron dilution m

would be the inadvertent opening of a valve in the boron thermal IT8 regeneration system (BTRS).

For this case, highly borated RCS water Ah x m is depleted of boron as it passes through the BTRS and is returned D o P

via the volume control tank. The following conditions are assumed 6 gg for an uncontrolled boron dilution during refueling.

f vs C

Technical Specifications require the reactor to be borated to at 3d least 2,000 ppm and shutdown by at least 5.0 percent Ak/k at 23 " b P

refueling. The maximum boron concentration to lose all shutdown n1 T1 o) margin is very conservatively estimated to be 1,500 ppm.

-C

.E @

O Dilution flow is assumed to be the maximum capacity of the BTRS (120 7

gpm) with 0 ppm water returning to the RCS. This is assumed 6

k although normally this system is nct operated at refueling condi-5 A

tions, tA O

Mixing of the reactor coolant is accomplished by the operating of y

one residual heat removal pump.

Y A minimum water volume (3000 ft3) in the RCS is used.

This is a 23 7-conservative estimate of the minimum volume of the RCS for residual I

heat removal system operation..

h

[

zu 15.2-22

~ AMENDMENT-S m

M"o Jastmftst, 1981 li t

2.

Dilution During Cold Shutdown Technical Specifications require the reactor to be shutdown by at least 1.0 percent ak/k ddring cold shutdown. The minimum boron

~

concentration required to meet this shutdown margin is conser-vatively estimated to be 1672 ppm.

If the reactor is in cold shut-down and on the residual heat removal system with RCS piping filled and vented, the following conditions are assumed for an uncontrolled boron dilution. Dilution flow is assumed to be a maximum of 150 gpm, which is the capability of one primary water makeup pump to deliver unborated water to the RCS. Mixing of the reactor coolant g

is accomplished by the operation of one residual heat removal pump.

\\

j A minimum volume of 4705 f t3 in the reac' tor coolant system is used. This corresponds to the active volu=e of the reactor coolant system minus the pressurizer volume, while on the residual heat removal system.

'23 If the reactor is in cold shutdown and the RCS water level is drained down to reactor vessel mid nozzle while on RHR, an inadver-tent dilution is prevented by administrative controls which isolate the RCS fro = the potential source of unborated water. Valves !tev-M30

% R"'d 8454, 8441, and 8439 will be locked closed during operations in these conditions. These valves block all flow paths that could allow unborated makeup w'ater to reach the RCS.

Any make-ado ii up which is. required will bebe EO TD THE REAC. TOR, CCOLANT 595 TEM BV at= w:grz:r =r qp li d

+,-a= tisu 1ct3T izy UNLoG4N6 THE6E V.ALVE5 AND INITIArtM6 TME. AEckulREO BLENDED T#e tme tunsa m.

cv twee MAKEUP WATEWFt.oDO.nenAFT R. THE REGLVIRED VOLUN\\E cp BLEN DED MAREVP WATER Ft.oW RA6 'EEEN ADDE 0 THE6E VALVES WILL AS,AIN EE 3

3.- Dilution During llot Standby 1.OCKEo ct.osEo. AM ALTER NATE 60VRCE op(

BoAATED WATER THAT MAV EE V6ED 15 FRotvk require gEpVELIM6 WATT 6R STORAGE TAMW TO nu THE.

Technical Spectfteattons e reactor to be shutdown oy at MMW6 least 1.77% Ak/k during hot standby. The minimum boron concentra-PtfMfD tion required to meet this shutdown margin is very conservatively M DOM.

estimated to be 1556 gpm.

The following conditions are assumed for a' continuous boron dilution during hot standby:

ay 15.2-22a

.AMENDMEfff B Pod JdusWeeK, 1981 gi